RNP-RA/05-0001, License Amendment Request Regarding Implementation of the Alternative Source Term

License Amendment Request Regarding Implementation of the Alternative Source Term

ML050250200
Person / Time
Site:	Robinson
Issue date:	01/21/2005
From:	Lucas J Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/05-0001
Download: ML050250200 (19)

Text

10 CFR 50.67 E Progress Energy 10 CFR 50.90 Serial: RNP-RA/05-0001 JAN 2 1 2009 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 LICENSE AMENDMENT REQUEST REGARDING IMPLEMENTATION OF THE ALTERNATIVE SOURCE TERM Ladies and Gentlemen:

In accordance with 10 CFR 50.67(b)(1) and 10 CFR 50.90, Progress Energy Carolinas (PEC), Inc., also known as Carolina Power and Light Company, requests a change to the licensing basis for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The change would implement the Alternative Source Term (AST) for the analysis of the radiological consequences of a design basis Loss of Coolant Accident (LOCA). There are no changes proposed to the Operating License or Technical Specifications.

In letters dated May 10, 2002, March 12, 2003, April 10, 2003, March 5, 2004, and July 22, 2004, PEC had previously submitted a license amendment request for full implementation of the AST (TAC No. MB5105). In a letter dated September 24, 2004, the NRC approved the implementation of the AST at HBRSEP, Unit No. 2, with the exception of the submitted LOCA analysis.

Attachment II provides a discussion of the LOCA dose analysis assumptions that were revised from those provided in the previous submittals, including a revision to the aerosol removal coefficient modeling. The proposed modeling for aerosol removal uses a spray removal coefficient based on Standard Review Plan 6.5.2 methodology, plus an additional factor based on diffusiophoresis. Diffusiophoresis is not discussed in Regulatory Guide 1.183, however, this removal mechanism has been accepted in the application of the AST at two other sites. Technical justification for this revised assumption, as well as the other revised assumptions, is also provided in Attachment II.

Attachment III provides a listing of the LOCA dose analysis assumptions and the dose analysis results. Attachment IV provides a No Significant Hazards Consideration and an Environmental Impact Consideration for the proposed amendment.

Attachment I provides an Affirmation pursuant to 10 CFR 50.30(b).

Progress Energy Carolinas, Inc.

Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

United States Nuclear Regulatory Commission Serial: RNP-RA/05-0001 Page 2 of 2 In accordance with 10 CFR 50.91(b), the State of South Carolina is being provided a copy of this letter.

PEC requests approval of this change by July 29, 2005. This requested date is based on the core reload safety analyses required for the next refueling outage, which is currently scheduled to begin on September 17, 2005.

If you have any questions concerning this matter, please contact Mr. C. T. Baucom at (843) 857-1253.

Sincerely, Manager - Support Services - Nuclear Attachments:

I. Affirmation II. Revised LOCA Analysis Assumptions - Technical Justification III. LOCA Analysis Assumptions and Results IV. No Significant Hazards Consideration and Environmental Impact Consideration RAC/rac c: Mr. T. P. O'Kelley, Director, Bureau of Radiological Health (SC)

Mr. H. J. Porter, Director, Division of Radioactive Waste Management (SC)

Dr. W. D. Travers, NRC, Region II Mr. C. P. Patel, NRC, NRR NRC Resident Inspectors, HBRSEP Attorney General (SC)

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/05-0001 Page 1 of I AFFIRMATION The information contained in letter RNP-RA/05-0001 is true and correct to the best of my information, knowledge, and belief; and the sources of my information are officers, employees, contractors, and agents of Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company. I declare under penalty of perjury that the foregoing is true and correct.

Executed On: /1 '/ 05 William G. Noll Director - Site Operations HBRSEP, Unit No. 2

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/05-0001 Page 1 of 5 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REVISED LOCA ANALYSIS ASSUMPTIONS - TECHNICAL JUSTIFICATION The following provides a discussion and technical justification for the Loss of Coolant Accident (LOCA) analysis assumptions that have been revised since the most recent submittal provided in the March 5, 2004 letter.

1. Time period of no containment spray during switchover to recirculation The previously assumed 23 minute "no spray" time has been replaced by a 10 minute "no spray" time. The revised timeline is:

• 3 minutes (0.05 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) spray automatically initiates

• 77 minutes (1.2833 hours0.0328 days <br />0.787 hours <br />0.00468 weeks <br />0.00108 months <br />) spray stopped by operators for switchover

• 87 minutes (1.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />) spray re-initiated in recirculation mode

• 167 minutes (2.7833 hours0.0907 days <br />2.176 hours <br />0.013 weeks <br />0.00298 months <br />) spray stopped due to stable containment conditions This change is based on an earlier start time for the operator actions to perform the switchover from injection mode to recirculation mode. The allowable time for the operators to perform the manual switchover tasks was not affected. The previous spray timing inputs conservatively assumed that operators would take no action until the Low-Low Refueling Water Storage Tank (RWST) Level Alarm required termination of sprays. This has been determined to be an overly conservative assumption, since based on existing procedures, the operators take preparatory actions beginning at the Low RWST Level Alarm. The revised 10 minute "no spray" assumption is consistent with current procedures.

2. Particulate (aerosol) removal model The March 5, 2004 submittal replaced the Standard Review Plan (SRP) 6.5.2 spray removal model with a 50 percentile Power's model, as incorporated into the RADTRAD code. The proposed reanalysis does not use the Power's model, but employs the 'user defined removal coefficient' option of the code. The user defined coefficients in this analysis are based on the spray removal methodology of SRP 6.5.2, plus an adder based on diffusiophoresis. The development of these coefficients is discussed in the following sections.

3. Spray particulate removal coefficient - SRP 6.5.2 method The spray particulate removal coefficient was developed by implementing the Standard Review Plan 6.5.2, Section JII.4.c(4), formula that is a function of spray flow, fall height, sprayed volume of the containment, and a dimensionless efficiency factor (E/D) that is either 10 or 1, depending on whether a decontamination factor (DF) of 50 has been achieved yet or not.

Removal coefficient (particulate) = 3*h*F*(E/D) 2*V

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/05-0001 Page 2 of 5 Where: E/D = Dimensionless efficiency; =10 (initial) or 1 (after DF=50)

F = Spray flow (ft3 /hr) h = Fall height (meters)

V = Sprayed volume (ft3 )

Spray flow of 932 gpm (7474.94 ft3 /hr) was calculated for the "A" train and 968 gpm (7763.67 ft /hr) for the "B" train. The fall height for "A" train from the "average" spray ring to the containment flood level is 165.5 ft (50.444 meters). For "B" train, the fall height is 154.1 feet (46.970 meters). The sprayed volume for "A" train is 82.9% of containment, or 1,623,618.1 ft 3 .

For "B" train, the sprayed volume is 81.5%, or 1,596,198.7 ft3 . Using these numbers, the two train dependent spray removal coefficients are:

"A" Train: 3*h*F*(E/D) = 3

• 50.444

• 7474.94

• 10 2*V 2

• 1623618.1

= 3.484 per hour "B" Train: 3*h*F*(E/D) = 3

• 46.97

• 7763.67

• 10 2 *V 2

• 1596198.7

= 3.427 per hour

4. Particulate removal coefficient - diffusiophoresis The above spray removal coefficients will be augmented with a particulate removal term based on diffusiophoresis. Diffusiophoresis is basically the removal of particulates by: a) the condensation of containment steam produced by the large break LOCA onto airborne particulates, b) the entrainment and removal of particulates via steam condensation onto passive heat sinks; or c) condensation and removal of particulates by active, forced heat removal mechanisms, such as fan coolers, or the condensation of steam caused by the cooling action of the sprays. This is a separate mechanism from the spray removal term derived above. Since these mechanisms are separate, the removal coefficients for each can be combined into an effective total particulate removal coefficient, and treated in RADTRAD as a constant, user input removal term.

The diffusiophoresis removal coefficient is time dependent based upon containment conditions.

However, for this analysis, a single conservative value was chosen that bounds all calculated values during the time period when a spray and diffusiophoresis removal factor are applied (approximately 2.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />, at which time spray flow is terminated). The resulting value is an adder of 0.5 to the removal coefficient. Conservatively, this adder is only applied to the sprayed region of the containment and only during the period that sprays are operating. Hence, the total user defined removal coefficients are 3.984 per hour for the "A" train and 3.927 per hour for the "B" train.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/05-0001 Page 3 of 5 Diffusiophoresis is not discussed in Regulatory Guide 1.183, however, the application of diffusiophoresis as an acceptable removal mechanism in addition to sprays in Alternative Source Term (AST) LOCA analyses has been approved by the NRC for at least two other sites. The applicable references are "Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendment RE: Selective Implementation of Alternative Source Term and Control Room Habitability Technical Specification Changes (TAC Nos. MB5303 and MB5304), September 10, 2003," and "Fort Calhoun Station, Unit No. 1 - Issuance of Amendment (TAC No MB1221),

December 5, 2001." For these other two sites, time dependent removal coefficients that considered both spray effectiveness and diffusiophoresis were developed using the SWNAUA computer code. The approach proposed for the H. B. Robinson Steam Electric Plant (HBRSEP),

Unit No. 2, dose analysis uses these precedents to justify a simplified application of the diffusiophoresis removal mechanism. The analysis will not use the SWNAUA code to determine a time dependent aerosol removal. Instead, the analysis documents that a conservative, bounding constant diffusiophoresis adder may be used in conjunction with the SRP spray removal coefficient.

The HBRSEP, Unit No. 2, containment response to the large break LOCA was evaluated in detail by Westinghouse in WCAP-15304, "Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2, LOCA Containment Integrity Analysis, September 1999." The basic output of that WCAP analysis consists of tables of mass and energy release, and containment atmospheric conditions. Event results such as containment relative humidity, containment atmospheric total pressure and steam temperature, sump water temperature, break flow (each end of double-ended guillotine break), and break flow energy/enthalpy were extracted directly from the WCAP results tables. The WCAP event progression was also determined and compared to the AST LOCA dose consequence event progression to determine the suitability of the WCAP results for use in dose applications.

Steam condensation rate is an input into the calculation of the diffusiophoresis factor. The containment analysis computer runs summarized in the WCAP do not contain steam condensation rate as a direct output. The available WCAP outputs and steam tables were used in a spreadsheet to determine steam condensation rates. The spreadsheet used the break flow mass and energy inputs from the WCAP to determine a steam addition rate that was assumed to be balanced by steam condensation rates during times when containment pressure is not changing rapidly. Steam condensation by active heat removal mechanisms (such as containment fan coolers, and the cooling effects of the containment sprays) were modeled, in addition to the derived condensation that occurs on heat sink masses within containment. The results are plotted in the figure below. The spreadsheet technique was qualified for this application by applying it to the similar information described in the submittals from the other sites. The spreadsheet produced condensation rate results very similar to, and conservative, relative to the condensation rates reported by the other sites.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/05-0001 Page 4 of 5 RNP Steam Condensation for Dose Analysis (600 sec. No Spray) 100000.0 40 10000.00 30

-A--- Total

~ 100.00

_____Condensation O 100020 (Spray, Fan

- Cooler, and Wall) co -a Active Steam O.~ 100.00- - 0.010 0.. -10Removal Fan Cooler)(Spray,

.00, 0f-  ! _- _ 0 -.- Total Containment 10.00 -.- PRESSURE 1.00 -1 0 1 10 100 1000 10000 100000 1000000 Time (Seconds)

Diffusiophoresis factors were calculated based on a method provided in "Direct Measurement of Diffusiophoretic Deposition of Particles at Elevated Temperatures," Buntz, H., and Schock, W.,

Aerosol: Science, Technology, and IndustrialApplications ofAirborne Particles,Proceedingsof the First InternationalAerosol Conference, held September 17-21, 1984, Minneapolis, Minnesota. This method converts the plant specific containment steam/vapor fractions into a ratio of the aerosol deposition velocity to the steam deposition velocity, and uses that ratio, along with the time dependent steam density and steam condensation rate in the containment, to determine time dependent removal coefficients.

As a further conservatism, the diffusiophoresis factor employed in the dose analysis is assumed to be a constant rate, rather than the time dependent, generally decreasing rate produced by the estimating technique. The value selected for this constant adder (0.5 per hour) is conservatively smaller than any rate estimated for the duration of spray operation. The range of time dependent specific rates calculated begins at about 2 per hour (25 seconds), goes up to more than 4 per hour (about 60 seconds), then decreases to about 1.0 per hour at 1200 seconds, and ends up at about 0.55 per hour at 39,000 seconds.

5. Removal factor reduction times Changing the particulate removal models and parameters results in revised times for reaching aerosol removal decontamination factors at which the removal rate must be decreased. As noted above, a single effective aerosol removal coefficient that combined spray effectiveness with diffusiophoresis was used in the dose analysis. In accordance with the guidance in SRP 6.5.2 and Regulatory Guide 1.183, Appendix A, Position 3.3, this single, constant coefficient is decreased by a factor of 10 when an overall aerosol decontamination factor of 50 is reached. As

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/05-0001 Page 5 of 5 a further conservatism, diffusiophoresis is only credited when sprays are running, and is only applied in the sprayed region of containment. The total (spray plus diffusiophoresis) aerosol removal coefficient is used to determine when the DF=50 point is reached, and this process is applied in a manner consistent with the guidance in the RG 1.183, Appendix A, Position 3.3, for determining DF=50 timing. The revised time for reaching a DF of 50 is 2.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> for "A" train and 2.62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br /> for "B" train.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 1 of 7 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 LOCA ANALYSIS ASSUMPTIONS AND RESULTS Table 1 - Core Activity Available for Release Isotopes Curies Isotopes Curies Isotopes Curies Co-58 5.99E+05 Ru-103 9.87E+07 Cs-136 3.5213+06 Co-60 4.5813+05 Ru-105 6.83E+07 Cs-137 8.87E+06 Kr-85 7.3013+05 Ru-106 3.7313+07 Ba-139 1.15E+08 Kr-85m 1.51E+07 Rh-105 6.33E+07 Ba-140 1.13E+08 Kr-87 3.03E+07 Sb-127 5.38E+06 La-140 1.1713+08 Kr-88 4.2013+07 Sb-129 2.03E+07 La-141 1.0213+08 Rb-86 1.16E+05 Te-127 5.31 E+06 La-142 9.8313+07 Sr-89 5.90E+07 Te-127m 8.87E1+05 Ce-141 1.04E+08 Sr-90 6.16E+06 Te-129 1.90E+07 Ce-143 9.55E+07 Sr-91 7.39E+07 Te-129m 3.84E1+06 Ce-144 8.1813+07 Sr-92 7.88E+07 Te-131m 1.23E+07 Pr-143 9.34E+07 Y-90 6.6213+06 Te-132 8.9113+07 Nd-147 4.17E+07 Y-91 7.69E+07 I-131 6.20E+07 Np-239 1.25E+09 Y-92 7.93E+07 1-132 9.0213+07 Pu-238 2.8 1E+06 Y-93 6.0713+07 I-133 1.28E+08 Pu-239 2.4413+04 Zr-95 1.05E+08 I-134 1.41E+08 Pu-240 3.55E+04 Zr-97 1.00E+08 1-135 1.21E+08 Pu-241 9.89E+06 Nb-95 1.06E+08 Xe-133 1.28E+08 Am-241 1.1813+04 Mo-99 1.1613+08 Xe-135 3.68E+07 Cm-242 3.23E+06 Tc-99m 1.03E+08 Cs-134 1.25E+07 Cm-244 3.8813+05

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 2 of 7 Table 2 - Fraction of Fission Product Inventory Released During the Gap and Early In-Vessel Phases Gap Early Release In-Vessel Group Isotopes Phase Phase Total Noble Gases Xe, Kr 0.05 0.95 1.00 Halogens I, Br 0.05 0.35 0.40 Alkali Metals Cs, Rb 0.05 0.25 0.30 Tellurium Metals Te, Sb, Se 0.00 0.05 0.05 Ba, Sr Ba, Sr . 0.00 0.02 0.02 Noble Metals Ru, Rh, Pd, Mo, Tc, Co 0.00 0.0025 0.0025 Cerium Group Ce, Pu, Np 0.00 0.0005 0.0005 Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, 0.00 0.0002 0.0002 LanthnidesCm, Am

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 3 of 7 Table 3 - Control Room Ventilation System Parameters Parameter Value Control Room Free Air Volume 20,124 ft3 Normal Ventilation Flow Rates Filtered Makeup Flow Rate 0 cfm Filtered Recirculation Flow Rate 0 cfm Unfiltered Makeup Flow Rate 400 cfm Unfiltered Inleakage (Total) 170 cfm initial LOCA 100 cfm @1 hr Unfiltered Recirculation Flow Rate N/A Pressurization Mode Flow Rates Filtered Makeup Air Flow Rate 400 cfm Filtered Recirculation Flow Rate 2600 cfm Unfiltered Inleakage (Total) 170 cfm initial LOCA 100 cfm@1 hr Hagan Room Unfiltered Air Inleakage 70 cfm (Terminates after I hr)

Unfiltered Recirculation Flow Rate N/A Filter Efficiencies Elemental 95%

Organic 95%

Particulate 99%

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 4 of 7 Table 4 - TSC/EOF Ventilation System Parameters Parameter Value TSC/EOF Free Air Volume 262,640 ft3 Normal Ventilation Flow Rates Unfiltered Makeup Flow Rate 3420 cfm Unfiltered Inleakage (Total) 500 cfm Pressurization Mode Flow Rates Filtered Makeup Air Flow Rate 3420 cfm Filtered Recirculation Flow Rate N/A Unfiltered Inleakage (Total) 500 cfm Filter Efficiencies Elemental 99%

Organic 99%

Particulate 99%

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 5 of 7 Table 5 - LOCA Analysis Inputs/Assumptions Input/Assumption Value Power Level 2346 MWt (includes power uncertainty)

Maximum Fuel Assembly/Batch Burnup 60,000 MWD/MTU Fuel Enrichment Bounding enrichment of 4.95 w/o (up to 5.0 w/o bounded by sensitivity study)

Core Fission Product Inventory Table 1 Containment Leakage Rate 0 to 24 hrs 0.1% (by weight)/day After 24 hrs 0.05% (by weight)/day LOCA Release Phase Timing Gap Release 0 to 30 min Early In-Vessel 30 min to 1.8 hrs Core Inventory Release Fractions (Gap Release and Table 2 Early In-Vessel Phases)

Engineered Safety Feature (ESF) Systems Leakage 4 gph (two times the Technical Rate (from 21 min to 30 days) Requirements Manual 3.23 limit)

ESF Leakage Flash Fraction 5.3%

Containment Ground Release Location Containment nearest point to Control Room ventilation intake and TSC/EOF ventilation intake ESF Ground Release Auxiliary Building Residual Heat Removal Heat Exchanger Room nearest point to Control Room ventilation intake and TSC/EOF ventilation intake Sump Water Volume 21 - 40 min 35,850 ft3 40 - 51.5 min 40,889 ft3 51.5 min - 720 hrs 43,939 ft3

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 6 of 7 Table 5 (continued)

Control Room Ventilation System Table 3 Switch from Normal to Pressurization Mode Automatic on Safety Injection in 35 sec TSC/EOF Ventilation Table 4 Switch to Emergency Air Filtration Mode Manual switchover in 2 hrs Containment Natural Aerosol Deposition - applied 0.1/hr continuously to unsprayed region and to sprayed region when spray not operating Containment Spray Region Volume A Train 1,623,618.1 ft3 B Train 1,596,198.7 ft3 Containment Unsprayed Region Volume A Train 334,907.9 ft3 B Train 362,327.3 ft3 Flow Rate Between Sprayed and Unsprayed Volumes 65,000 cfm starting at 76 sec Spray Removal Rates Elemental Iodine 20/hr Time to Reach DF of 200 - A Train 2.01 hrs Time to Reach DF of 200 - B Train 2.03 hrs Particulate Iodine (includes diffusiophoresis)

A Train 3.984/hr (0.3984 at 2.58 hrs)

B Train 3.927/hr (0.3927 at 2.62 hrs)

Spray Operation Initiation Time - RWST Suction 3 min Termination Time - RWST Suction 77 min Initiation Time - Containment Sump Suction 87 min Termination Time - Containment Sump Suction 167 min Atmospheric Dispersion Factors Table 6 Breathing Rates Offsite RG 1.183, Section 4.1.3 Control Room and TSCJEOF RG 1.183, Section 4.2.6 Control Room and TSC/EOF Occupancy Factors RG 1.183, Section 4.2.6

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/05-0001 Page 7 of 7 Table 6 - Ground Level Onsite & Offsite X/Q's (sec/m 3 )1')

T Exclusion Low CV RHR HX CV RHR Time Area Population Nearest Room to Point Room to Peid Boundary 2) Zone Point CR CR PoSC CRomt 0 - 2 hrs 1.77E-03 8.92E-05 4.15E-03 7.13E-03 1.64E-04 1.38E-04 2 - 8 hrs 1.77E-03 3.50E-05 2.74E-03 5.49E-03 1.43E-04 1.23E-04 8 - 24 hrs 1.77E-03 2.19E-05 1.17E-03 2.29E-03 6.49E-05 5.52E-05 1 -4 days 1.77E-03 7.95E-06 8.18E-04 1.71E-03 4.411E-05 3.78E-05 4 - 30 days 1.77E-03 1.85E-06 6.74E-04 1.37E-03 3.50E-05 3.01E-05 NOTES:

(1) These values have not been corrected for Control Room (CR) occupancy.

(2) Conservatively, the highest X/Q value was used for all time intervals to determine the worst two hour Exclusion Area Boundary (EAB) dose.

Table 7 - LOCA Dose Consequences(5)

EAB(') LPZ(2) CR TSC/EOF Item (REM (REM (REM (REM TEDE) TEDE) TEDE) TEDE)

Containment 21.8 1.31 2.45 1.82 R elease__ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _

ESF Leakage 1.1 0.23 1.94 0.18 Containment 0.03 0.03 Shine__ _ _ _ _ _ __ _ _ _ _ _

Plume Shine - 0.03 0.02 Total Dose 22.9 1.54 4.45 2.05 Regulatory 25(3) 25(3) 5(4) 5(4)

Limit NOTES:

(1) Worst two hour integrated dose.

(2) 30-day integrated dose.

(3) RG 1.183, Table 6, "Accident Dose Criteria."

(4) 10 CFR 50.67 and 10 CFR 50, Appendix A, Criterion 19.

(5) Based on the limiting B Train of containment spray.

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/05-0001 Page 1 of 4 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 NO SIGNIFICANT HAZARDS CONSIDERATION AND ENVIRONMENTAL IMPACT CONSIDERATION NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company, is proposing changes to the licensing basis of Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. These changes revise the licensing basis for HBRSEP, Unit No. 2, to implement the Alternative Source Term (AST) described in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," for the evaluation of the Loss of Coolant Accident (LOCA). There are no Operating License or Technical Specifications changes required or proposed to implement this licensing basis change.

The NRC has previously approved the AST dose analyses for the Fuel Handling Accident in a letter dated October 4, 2002, and for the Main Steam Line Break, Reactor Coolant Pump Shaft Seizure (Locked Rotor), Single Rod Control Cluster Assembly (RCCA)

Withdrawal, and Steam Generator Tube Rupture (SGTR) in a letter dated September 24, 2004. As part of the full implementation of the AST, the total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

Revision of the LOCA analysis to the Alternative Source Term methodology does not affect the design or operation of HBRSEP, Unit No. 2. Rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences of the postulated accident. The implementation of the Alternative Source Term has been evaluated in revisions to the LOCA dose analysis at HBRSEP, Unit No. 2. Based on the results of this analysis, it has been demonstrated that the dose consequences are within the regulatory guidance provided by the NRC. This guidance is presented in 10 CFR 50.67 and Regulatory Guide 1.183.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/05-0001 Page 2 of 4

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated.

The proposed change does not affect plant structures, systems, or components.

The proposed change is to an evaluation methodology and does not initiate design basis accidents.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed change is associated with the implementation of a new licensing basis for HBRSEP, Unit No. 2. The new licensing basis implements an Alternative Source Term in accordance with 10 CFR 50.67 and the associated Regulatory Guide 1.183. The results of the revised limiting design basis analysis are subject to revised acceptance criteria. The analysis has been performed using conservative methodologies in accordance with regulatory guidance or other methodologies approved by the NRC in prior plant-specific license amendments.

The dose consequences are within the acceptance criteria found in the regulatory guidance associated with Alternative Source Terms.

The proposed change continues to ensure that doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Specifically, the margin of safety for the radiological consequences of these accidents is considered to be that provided by meeting the applicable regulatory limits.

Therefore, this change does not involve a significant reduction in a margin of safety.

Based on the above discussion, Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company, has determined that the requested change does not involve a significant hazards consideration.

ENVIRONMENTAL IMPACT CONSIDERATION 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions for categorical exclusion for performing an environmental assessment. A proposed change for an operating license for a facility does not require an environmental

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/05-0001 Page 3 of 4 assessment if operation of the facility in accordance with the proposed change would not (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; (3) result in an increase in individual or cumulative occupational radiation exposure.

Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company, has reviewed this request and determined that the proposed change meets 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 is required in connection with the issuance of the amendment. The basis for this determination follows:

Progress Energy Carolinas, Inc., also known as Carolina Power and Light Company, is proposing a change to the licensing basis of Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. This change revises the licensing basis for HBRSEP, Unit No. 2, to implement the Alternative Source Term (AST) described in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," for the evaluation of the Loss of Coolant Accident (LOCA). There are no Operating License or Technical Specifications changes required or proposed to implement this licensing basis change.

The NRC has previously approved the AST dose analyses for the Fuel Handling Accident in a letter dated October 4, 2002, and for the Main Steam Line Break, Reactor Coolant Pump Shaft Seizure (Locked Rotor), Single Rod Control Cluster Assembly (RCCA)

Withdrawal, and Steam Generator Tube Rupture (SGTR) in a letter dated September 24, 2004. As part of the full implementation of the AST, the total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

Basis The proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not involve a significant hazards consideration.

2. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated and does not result in the possibility of a new or different kind of accident. Therefore, the proposed change does not result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite.

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/05-0001 Page 4 of 4

3. The Alternative Source Term does not affect the design or operation of the facility. Rather, once the occurrence of an accident has been postulated, the Alternative Source Term is an input to evaluate the consequences. The implementation of the Alternative Source Term has been evaluated in revisions to the analysis of the LOCA at HBRSEP, Unit No. 2. Based on the results of this analysis, it has been demonstrated that the dose consequences of the LOCA event are within the regulatory guidance provided by the NRC for use with the Alternative Source Term. Therefore, the proposed change does not result in a significant increase in either individual or cumulative occupational radiation exposures.