Text

Reply to Information Requested at Pubic Meeting Between Rg&E and NRC Staff Held on August 19, 2003
	ML033430141
Person / Time
Site:	Ginna
Issue date:	12/01/2003
From:	Mecredy R; Rochester Gas & Electric Corp
To:	Clark R; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML033430141 (100)
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Always at Your Service Robert C. Mecredy Vice President Nuclear Operations December 1, 2003 Mr. Robert L. Clark Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission Attn:

Document Control Desk Washington, D.C. 20555-0001

Subject:

References:

Reply to Information Requested at Public Meeting between RG&E and NRC Staff Held on August 19, 2003.

R.E. Ginna Nuclear Power Plant Docket No. 50-244

1. Letter from Robert C. Mecredy (RG&E) to Robert L. Clark (NRC) dated May 21, 2003, License Amendment Request Regarding Revision of Ginna Technical Specification Sections 1.1, 3.3.6, 3.4.16, 3.6.6, 3.7.9, 5.5.10, 5.5.16, and 5.6.7 Resulting From Modification of the Control Room Emergency Air Treatment System and Change in Dose Calculation Methodology to Alternate Source Term.

2. Letter from Robert C. Mecredy (RG&E) to Robert L. Clark (NRC) dated September 30, 2003, Summary of Public Meeting Between RG&E and NRC Staff Held on August 19, 2003.

Dear Mr. Clark:

On August 19, 2003 representatives of Rochester Gas and Electric Corporation (RG&E) met with members of the NRC Staff at your offices in White Flint. The purpose of the meeting was to brief the Staff and provide an overview of our Control Room Emergency Air Treatment System (CREATS) License Amendment Request (Ref. 1). Subsequent to the above meeting, RG&E summarized that meeting in a letter (Ref. 2), and committed to re-perform certain analysis and provide additional information. The attachments to this letter verify that those commitments are satisfied and should be docketed as an addendum to Reference 1.

I declare under penalty of perjury under the laws of the United States of America that I am authorized by Rochester Gas and Electric Corporation to submit this documentation and that the foregoing is true and correct.

If you have questions regarding the content of this correspondence please contact Mr. Mike Ruby at (585) 771-3572.

Executed on December 1, 2003 Sincerely obert C. Mecredy I oac94 An equal opportunity employer 89 East Avenue I Rochester, NY 14649 tel (585) 546-2700 www.rge.com AK'1CP

(

I An Energy East Company

Attachments:

1.

Verification of Commitment Resolution

2.

Summary of Radiological Analysis, Revision 1

3.

Particulate Removed by Sprays

4.

MSLB/Locked Rotor Two Hour Steam Release

5.

Clarification of SGTR Steam Releases

6.

PWR Natural Deposition Coefficients

7.

UFSAR Excerpt Cc:

Mr. Robert L. Clark (Mail Stop 0-8-C2)

Project Directorate I Division of Licensing Project Management Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Mr. Peter R. Smith, Acting President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire State Plaza, 10t Floor Albany, NY 12223 Verification of Commitment Resolution

Attachment I Verification of Commitment Resolution

1. NRC would prefer that RG&E use 2% versus 1% iodine partitioning for beyond 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months of Emergency Core Cooling System (ECCS) leakage.

Initial Response:

Agree. RG&E will revise the appropriate dose calculations.

Resolution:

The Loss of Coolant Accident (LOCA) dose analysis was revised to reflect the above and to incorporate the new X/Q values for the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ).

The Summary of Radiological Analysis (Attachment 2) was revised to reflect the new analysis.

2. Table 5.2 of the Summary of Radiological Analysis in the LAR submittal (page 38) shows that RG&E assumed certain efficiencies of the Containment (CNMT) post accident charcoal filters, when RG&E is planning to remove these filters.

Initial Response:

This is incorrect and will be revised.

Resolution: Summary of Radiological Analysis (Attachment 2) has been revised to correct the indicated omission.

3. The NRC needs the calculation for spray removal coefficient for particulate as described in the Summary of Radiological Analysis attachment to the LAR.

Initial Response:

RG&E will provide the calculation.

Resolution: See Attachment 3.

4. For Main Steam Line Break (MSLB) and Locked Rotor accidents, what is the steam release assumed for the intact SG at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ?

Initial Response:

RG&E agreed to provide the requested information.

Resolution:

See Attachment 4.

5. Please clarify LAR Attachment 1, Table 8.2 and what the mass flows actually mean, including the partitioning factors for Steam Generator Tube Ruptures (SGTR)

Initial Response:

RG&E will clarify the information in a future submittal.

Resolution: The Summary of Radiological Analysis (Attachment 2) was revised to clarify table values. See Attachment 5 for further explanation.

6. In LAR Attachment 1, Table 10.1, what is the basis for the natural deposition coefficient value?

Initial Response:

RG&E agreed to provide this information.

Resolution: See Attachment 6.

7. On LAR Attachment 1 page 77, the second paragraph under 12.2, what is meant by "Section 7.1.6".

Initial Response:

This is apparently a typographical error which RG&E agreed to correct, and provide the necessary information in a future submittal.

Resolution: The Summary of Radiological Analysis (Attachment 2) was revised to correct the error.

8. Why did RG&E assume a 2-hour dispersion for Gas Decay Tank (GDT) rupture, especially since the previous analysis done for Ginna, related to TS Amendment 78, assumed a puff release?

Initial Response:

RG&E agreed to revise the analysis and do a puff release with no CR isolation, consistent with the analysis performed for Amendment 78.

Resolution:

The GDT dose analysis was revised to reflect the above comment and to incorporate the new atmospheric dispersion (X/Q) values for the EAB and LPZ. The control room doses were also calculated out to 30 days for consistency. The Summary of Radiological Analysis (Attachment 2) was revised to reflect the new analysis.

9. LAR Attachment 1 page 80, bottom of page has typo (renumbers).

Initial Response:

RG&E agreed and will correct the typo.

Resolution: The Summary of Radiological Analysis (Attachment 2) was revised to correct numbering error.

10. What is the technical basis for lowering the required NaOH tank levels?

Initial Response:

RG&E agreed to provide the basis.

Resolution: The basis for reducing the required NaOH tank level is described in the Ginna UFSAR, Section 6.1.2.1.4. See Attachment 7.

2/2.

Summary of Radiological Analysis

Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 R. E. Ginna Station Docket Number 50-244 Summary of Radiological Analyses Alternative Source Term and Control Room Emergency Ventilation System Submittal Revision 1 November 2003

TABLE OF CONTENTS 1.0 Summary of Radiological Analysis....................................

3 2.0 Atmospheric Dispersion (XIQ)....................................

4 3.0 Iodine Spiking....................................

29 4.0 General Discussion....................................

32 5.0 Loss-of-Coolant-Accident....................................

33 6.0 Fuel Handling Accident....................................

42 7.0 Main Steam Line Break....................................

49 8.0 Steam Generator Tube Rupture (SGTR)..................................

55 9.0 Locked Rotor Accident....................................

60 10.0 Rod Ejection Accident....................................

65 11.0 Tornado Missile in Spent Fuel Pool....................................

71 12.0 Waste Gas Decay Tank Rupture....................................

77 13.0 References....................................

80 Summary of Radiological Analyses, Revision 1, 11/03 Page 2 of 81

1.0 Summary of Radiological Analysis Each of the below accidents was analyzed for dose consequences using the Alternative Source Term Methodology per Regulatory Guide 1.183. All dose results are expressed in terms of TEDE for comparison with the appropriate limits. The accident consequences were calculated for both the Control Room Operator and the public at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ). The following table summarizes the results of the analysis.

TABLE 1.1 ALTERNATE SOURCE TERM DOSE ANALYSIS

SUMMARY

Accident EAB Max. 2-hour LPZ Control Room Limit Dose Limit Dose Limit Dose LOCA 25.0 3.70 25.0 0.92 5.0 3.51 FHA-CNMT 6.3 1.1 6.3 0.07 5.0 1.2 FHA-AUX 6.3 0.31 6.3 0.02 5.0 0.09 MSLB' 2.5 1.05 2.5 0.15 5.0 0.64 MSLB2 25.0 0.15 25.0 0.03 5.0 0.18 SGTR' 2.5 0.22 2.5 0.02 5.0 0.14 SGTR2 25.0 0.71 25.0 0.05 5.0 0.88 Locked Rotor 2.5 2.754 2.5 0.554 5.0 3.72 Rod Ejection 6.3 1.47 6.3 0.24 5.0 1.04 SFP-TMA 6.3 0.073 5.0 0.06 GDT Rupture 0.5 0.172 0.5 0.013 5.0 0.067

' Accident Initiated Iodine Spike 2 Pre-Accident Iodine Spike 3 EAB X/Q = 1.74E-6 calculated as discussed in Section 2.7 4 EAB and LPZ X/Q calculated by PAVAN, see Section 2.8 Summary of Radiological Analyses, Revision 1, 11/03 Page 3 of 81

2.0 Atmospheric Dispersion (X/Q)

The atmospheric dispersion factors currently described within the UFSAR were reviewed as part of the control room ventilation system upgrade. As a result of this review, the atmospheric dispersion factors for the control room intake were recalculated as described below. The atmospheric dispersion factors for the EAB and LPZ are described in Section 2.8.

The atmospheric dispersion factors for each pathway from on-site source to control room intake were recalculated using the ARCON96 code (Reference 1) combined with the draft 2 Reg. Guide DG-1 111 methodology (Reference 2).

The meteorological data collected by a Regulatory Guide 1.23 system for the years 1992, 1993, and 1994 was used in the calculations. This data is considered to be typical of any time period. This data was readily available and used in prior submittals. The data covered 26,304 hours0.00352 days 0.0844 hours 5.026455e-4 weeks 1.15672e-4 months , of which 512 hours0.00593 days 0.142 hours 8.465608e-4 weeks 1.94816e-4 months were missing or invalid. This represents approximately 2% which is within the ARCON96's default setting of 10%.

The wind speed statistics for a typical year (1992) are:

Average wind speed:

Maximum:

Total hours (including invalid):

Invalid hours:

4.16 m/sec 24.5 m/sec 8572 212 The stability distribution for the same year (1992) was:

Stability Class A

B C

D E

F G

Duration (hr) 739 466 351 3132 2389 835 660 Summary of Radiological Analyses, Revision 1, 11/03 Page 4 of 81

2.1 Containment Leakage The containment shell is modeled as a diffuse vertical area source. This source is used in the dose calculations for LBLOCA, and the containment leakage portion of a control rod ejection accident. The source width is the containment O.D. and the source height is the distance from ground to the top of the containment dome. This is consistent with Reference 2, Figure 1. The diffuse source model is used because leakage is assumed to be distributed over the containment surface and all penetrations, not isolated to a specific point.

The source to receptor distance uses the shortest horizontal distance from the containment surface to the intake and assumes the source and receptor are at the same height. This results in the shortest source to receptor distance as illustrated by points C and B on Figure 2.1.B. The ARCON96 input parameters and resulting X/Qs are presented on Table 2.1 and Figures 2.1A and 2.1B.

TABLE 2.1 CONTAINMENT LEAKAGE INPUT AND RESULTS Distance to receptor, m 32 Intake height, m 13.8 Direction to source, degrees 247 Release type ground level, diffuse vertical area Release height, m 13.8 Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m GoA 5.7 Ozo 5.9 Lower measurement height, m 10 Upper measurement height, m 100 Elevation difference, m 0

Summary of Radiological Analyses, Revision 1, 11/03 Page 5 of 81

TABLE 2.1 CONTAINMENT LEAKAGE INPUT AND RESULTS Resulting X/Q, sec/rn3 0-2 hr 1.57 E-03 2-8 hr 1.12 E-03 8-24 hr 4.47 E-04 1-4 days 3.69 E-04 4-30 days

3. 10 E-04I Summary of Radiological Analyses, Revision 1, 11/03 Page 6 of 8 l

FIGURE 2.1A - CONTAINMENT AND CONTROL BUILDING ELEVATIONS CONTAINMENT NOT TO SCALE 313 CONTROL BUILDING AIR INTAKE Summary of Radiological Analyses, Revision 1, 11/03 Page 7 of 81

FIGURE 2.1B - CONTAINMENT LEAKAGE PLAN VIEW Column Column 22 1 ft TURBINE BLDG Column 72f B

r CONTROL BLDG AIR INTAKE (B)

INTERMEDIATELow Roo BLDG t1, 7, Summary of Radiological Analyses, Revision 1, 11/03 Page 8 of 81

2.2 Containment Equipment Hatch (Roll-Up Door)

This vertical area source is used for the fuel handling accident in Containment. In this case, all Containment leakage is assumed to come from the equipment hatch, a large penetration located in the south-east sector of the Containment perimeter. During refueling, the hatch is removed, and the open penetration is covered by a roll-up door.

The source dimensions are based on face area of the roll-up door. Radioactivity is postulated to leak through the open hatch, to the environment, through the perimeter seals of the roll-up door. The calculation uses:

1.

The shortest horizontal distance between the door perimeter and the Control Room Intake

2.

A diffuse vertical source is assumed. The dimensions being that of the roll-up door (23'6" wide, 22' high).

3.

The assumed release height is equal to the distance from grade to the top of the roll-up door. This results in the shortest source - receptor distance.

4.

The cross-section area of Containment was assumed for the building wake effect (1071 in2 ). A sensitivity run was made where the wake area was doubled. There was an insignificant change in X/Q. The ARCON96 input parameters and resulting X /Q are presented on Table 2.2 and Figure 2.2.

TABLE 2.2 CONTAINMENT EQUIPMENT HATCH INPUT AND RESULTS Distance to receptor, m 29 Intake height, m 13.8 Direction to source, degrees 227 Release type ground level, diffuse vertical area Release height, m 6.7 Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m Go0 1.2 GZ0 1.1 Lower measurement height, m 10 Summary of Radiological Analyses, Revision 1, 11/03 Page 9 of 81

TABLE 2.2 CONTAINMENT EQUIPMENT HATCH INPUT AND RESULTS Upper measurement height, m 100 Elevation difference, m 0

Resulting X/Q, sec/rn3 0-2 hr 5.64 E-03 2-8 hr 4.69 E-03 8-24 hr 1.66 E-03 1-4 days 1.58 E-03 4-30 days 1.31 E-03 Summary of Radiological Analyses, Revision 1, 1 1/03 Page 10 of 81

FIGURE 2.2 - ROLL-UP DOOR PLAN VIEW '

INTERMEDIATE

\\/

I ICONTROL BLDG

\\

L BLDG AIR

~ CONTAINMENTN e

is paltry INTAKE Roll-Up Door Summary of Radiological Analyses, Revision 1, 11/03 Page I 1 of SI

2.3 Atmospheric Relief Valves (ARVs)

This source is used for releases from the steam generators. The pathway is based on the ARV discharge that is closest to the Control Room Intake. This will result in larger X/Qs. The discharge of the ARV is modeled as a ground-level point source rather than an elevated vent since Reference 2 advises against using the vent release model, pending further NRC evaluation. The assumed release height is equal to the distance from grade to the vent point. The cross-section area of Containment was assumed for the wake area (1 071 m2). A sensitivity run was made where the wake area was doubled. There wvas an insignificant change in X/Q. The ARCON 96 input parameters and resulting X/Q are presented on Table 2.3 and Figure 2.3.

TABLE 2.3 ATMOSPHERIC RELIEF VALVES INPUT AND RESULTS Distance to receptor, m 40 Intake height, m 13.8 Direction to source, degrees 273 Release type ground level, point Release height, m 22 Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m G)0O 00 UzO Resulting X/Q, sec/m 3 0-2 hr 3.66E-03 2-8 hr 2.49E-03 8-24 hr 1.07E-03 1-4 days 7.86E-04 4-30 days 7.177E-04 Summary of Radiological Analyses, Revision 1, 11/03 Page 12 of SI

FIGURE 2.3 - ARVT GROUP B PLAN VIEWNI INTERMEDIATE BLDG CONTROL BLDG AIR INTAKE Summary of Radiological Analyses, Revision 1, 11/03 Page 13 of 81

2.4 Plant Vent This source is used for releases from a fuel handling accident in the spent fuel pool. Thle vent is modeled as a horizontal area source, rather than a vent source, based on guidance in Reference 2, which advises against using the vent release model pending further NRC evaluation. The assumption of an area source is considerably more conservative than the vent source assumption and only slightly less conservative than a point source. For the horizontal area source, the horizontal diffusion coefficient is based on the vent diameter (55") and the vertical coefficient is set to zero. The assumed release height is equal to the distance from grade to the vent point. The wake area of 1071 m2 is again assumed.

Doubling this value had an insignificant affect on X/Q. The ARCON96 input parameters and resulting X/Q are presented on Table 2.4 and Figure 2.4.

TABLE 2.4 PLANT VENT INPUT AND RESULTS Distance to receptor, m 53 Intake height, m 13.8 Direction to source, degrees 272 Release type ground level, diffuse horizontal area Release height, m 36 Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m Cryo 0.23 CFZO 0

Resulting X/Q, sec/M3 0-2 hr 1.79E-03 2-8 hr 1.15E-03 8-24 hr 4.95E-04 1-4 days 3.71 E-04 4-30 days 3.29E-04 Summary of Radiological Analyses, Revision 1, 11/03 Page 14 of SI

FIGURE 2.4 - PLANT VENT PLAN VIEW INTERMEDIATE BLDG Summary of Radiological Analyses, Revision 1, 11/03 Page 15 of 81

2.5 Auxiliary Building Leakage This source is used when ECCS leakage is considered. The subgrade floors of this building contain ECCS equipment that is postulated to leak. The source is modeled as a vertical area source assumed to be the building wall closest to the Control Room Intake.

The building north wall is modeled which approximates the cross-sectional area perpendicular to the line of site from the building surface to the control room intake. The assumed release height is the distance from grade to the top of the Auxiliary Building.

The wake area equivalent to the Containment cross-section area is again assumed. The shortest source - receptor distance is calculated from the corner of the Auxiliary Building to the Control Room Intake. The ARCON96 input parameters and resulting X/Q are presented on Table 2.5 and Figure 2.5.

TABLE 2.5 AUXILIARY BUILDING LEAKAGE INPUT AND RESULTS Distance to receptor, m 30 Intake height, m 13.8 Direction to source, degrees 183 Release type ground level, diffuse vertical area Release height, m 13 Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m GyO 3.9 RioZ2.1 Resulting XIQ, sec/m 3 0-2 hr 3.89E-03 2-8 hr 2.99E-03 8-24 hr 9.63E-04 1-4 days 8.98E-04 4-30 days 8.23E-04 Summary of Radiological Analyses, Revision 1, 11/03 Page 16 of 81

FIGURE 2.5 - AUXILIARY BUILDING LEAKAGE PLAN VIEW

/

\\

I 1

1 N

~~~CONTROL INTERMEDIATE `

\\'ElTRL BLDGEBLDG AIR CONTAINMENT (1

r INTAKE Summary of Radiological Analyses, Revision 1, 11/03 Page 17 of SI

2.6 Main Steam Header Turbine Building This source is used to model activity released from a main steamline break. The rupture site is assumed to be inside the Turbine Building on the mezzanine level. (See Section 7.1 for additional details.) Since the released steam is assumed to blow-out windows and metal siding of the Turbine Building, no confinement of the plume is assumed. The source is modeled as a ground level point source. The distance to the receptor is that from the header midpoint to the Control Room Intake. The release height is the distance from grade to the top of the header. The wake area equivalent to the Containment cross-section area is again assumed. The ARCON96 input parameters and resulting X/Q are presented on Table 2.6 and Figure 2.6.

TABLE 2.6 MAIN STEAM HEADER TURBINE BUILDING INPUTS AND RESULTS Distance to receptor, m 48 Intake height, m 13.8 Direction to source, degrees 278 Release type ground level, point source Release height, m 4

Building area, m2 1071 Sector width constant 4.3 Surface roughness 0.2 Initial diffusion coefficients, m 0ZO 0

Resulting X/Q, sec/m3 0-2 hr 2.57E-03 2-8 hr 1.92E-03 8-24 hr 8.08E-04 1-4 days 5.77E-04 4-30 days 5.50E-04 Summary of Radiological Analyses, Revision 1, 11/03 Page IS of 81

FIGURE 2.6 - STEAM LINE PLAN VIEW CONTROL BLDG AIR INTAKE Summary of Radiological Analyses, Revision 1, 11/03 Page 19 of 81

2.7 Tornado Missile The tornado missile accident assumes that a utility pole, propelled by the wind, penetrates the Auxiliary Building roof and impacts fuel stored in the spent fuel pool. The specific location of the impact cannot be predicted. Thus, the shortest source-receptor distance is conservatively assumed. The source is modeled as a ground level point source. The release height is the distance from grade to the spent fuel pool surface. The wake area is again assumed equivalent to the cross-section area of Containment.

The calculation of atmospheric dispersion for tornado conditions is a unique task that cannot be performed with ARCON96. The primary reason that ARCON96 can't be used is the lack of meteorological data for a tornado. Further, if data were available, the duration of a tornado is too short for ARCON96 to provide a meaningful average. An ARCON96 calculation typically averages I to 5 years of hourly meteorological data. A tornado would provide two data points at most.

While the use of the ARCON96 code is not practicable for tornado conditions, the use of the dispersion models executed by ARCON96 may be used to conservatively estimate dispersion. The CONHAB module of the HABIT code calculates a single, direction and time-independent dispersion value using the basis dispersion models developed for ARCON96.

CONHAB is used to calculate dispersion factors for tornado wind speeds and to also show the sensitivity of the model to stability class. The input and results for these cases is summarized in Table 2.7. The basis for the selection of wind speed and stability class is as follows:

wind speed The range is 24.5 to 60 meters/sec. 24.5 meters/sec is the maximum recorded hourly wind speed during normal atmospheric conditions. 132 miles/hour (about 60 meters/sec) is the wind speed for the design basis tornado.

Pasquill stability class Stability class is a user input to the CONHAB dispersion model. There are 7 stability classes, A through G. "A" represents extremely unstable conditions, and "G" represents stable conditions. Unstable conditions enhance dispersion. Stable conditions diminish dispersion. However, the dispersion model predicts a diminishing effect with increasing wind speed. Test cases are run to show this effect and also to show that there are no discontinuities or instabilities in the model due to increasing wind speed. The cases show converging X/Qs with increasing wind speed (Figure 2.7). For the cases listed on Table 2.7, Pasquill F Summary of Radiological Analyses, Revision 1, 11/03 Page 20of 81

provides somc conservatism over Pasquill A, even at wind speeds up to 60 meters!sec. Therefore, the tornado X/Q will be based on Pasquill F and a wind speed of 24.5 meters/sec., since this combination results in a larger X/Q.

TABLE 2.7 CONIlAB TORNADO MISSILE INPUT AND RESULTS Parameter CBIA I CBIF I CB2A CB2F1l CB3F Distance to receptor, m 67 503 Intake height, mn 13.8 Release type ground level, point sourcc Release height, m 2.1 Building area, m2 1071 Wind Speed, metcrs/sec 24.5 24.5 60 60 24.5 Stability Class A

F A

F F

Resulting X/Q, sec/m3 2.85E-5 4.36E-5 2.77E-6 3.03E-6 1.74E-6 Cases CBIA through CB2F are for the Control Room air intake.

Case CB3F is for the 503 meter EAB.

Summary of Radiological Analyses, Revision 1, 11/03 Page 21 of SI

FIGURE 2.7 - SPENT FUEL POOL PLAN VIEWN TURBINE BLDG INTERMEDLATE I

BLDG

-\\

I 1

CONTROL BLDG AIR INTAKE

1. E -0A E1~

23 25 30 35 40 45 50 55 Wind Spood (mlsoc) 60 Summary of Radiological Analyses, Revision 1, 11/03 Page 22 of 81

2.8 EAB and LPZ Atmospheric Dispersion Factors All the EAB and LPZ dose calculations used the current X/Qs presented in the Ginna UFSAR (Reference 3, Section 2.3.4.2.1) except for the tornado missile and locked rotor.

Due to the uniqueness of the tornado missile dose calculation and to maintain consistency between the control room calculation and the EAB calculation, the EAB calculation was done using the same methodology as used for the control room calculation. See Section 2.7 for a description of the methodology.

The ultra conservative assumptions used in the locked rotor dose calculation result in assuming 100% fuel failure. This assumption and the current UFSAR EAB X/Q results in an unrealistically high EAB dose. To obtain a more realistic dose, the EAB and LPZ X/Qs were recalculated using the PAVAN computer code. The recalculated values have only been used to calculate the locked rotor doses. The intent is to use these new values in any future calculations.

2.8.1 Current UFSAR Atmospheric Dilution Factors (Reference 3, Section 2.3.4.2.1)

Site boundary (0-2 hr) 4.8E-04 Low population zone 0-8 hr 3.0E-05 8-24 hr 2.1 E-05 1-4 days 8.6E-06 4-30 days 2.5E-06 Summary of Radiological Analyses, Revision 1, 11/03 Page 23 of 81

2.8.2 Recalculated Atmospheric Dispersion Factors (PAVAN code)

The same meteorological data used to calculate the Control Room X/Q was used to recalculate tile Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dispersion factors. The dispersion factors were calculated using KR PAVAN which is a PC-based version of the NRC's PAVAN code (Reference 9). Dispersion factors were calculated using the direction-dependent method and the direction-independent method.

The direction-dependent method determined the 0.5 percent X/Q value for each of the 16 wind speed directions. The direction-independent method determined the 5 percent X/Q value for the overall site. Thle direction-dependent dispersion values were limiting, for both EAB and LPZ boundaries. The input assumptions are listed on Table 2.8.

The EAB dispersion factor calculated by the ENVLOP routine of PAVAN is a conservative bounding 0.5 percent X/Q value. Since the EAB dose can be limiting for certain accidents, such as locked rotor, a more accurate X/Q value is desired. Therefore, the output X/Q vs. percentile for the limiting sector is analyzed in a spreadsheet to obtain a more accurate value.

The results of the spreadsheet analysis are shown in Figures 2.8.1 and 2.8.2. A logarithmic trend line is fit to the data and the resulting 0.5% X/Q value (0-2 hours) is determined. The 0.5 percent code and spreadsheet values for the limiting sector (SW) are:

code value 3.32E-4 equation value using all data points 3.368E-4 equation value using only low percentage data 2.978E-4 Visual inspection of the data and trend line show good agreement. This is also confirmed by the R2 values (0.9414 and 0.9388) which are close to 1.0. The EAB dispersion factor will be 2.98E-4.

The LPZ dispersion factor output was also evaluated using a spreadsheet. The equation value, using all data, was higher than tile value generated by KR PAVAN and tile equation value using only the low probability data was lower than the value generated by KR PAVAN. Since a high degree of accuracy is not needed for the LPZ values and the KR PAVAN value is reasonable, the code values will be used for the LPZ.

The X/Q values (sec/m3) are:

Boundary 0-2 hr 0-8 hr 8-24 hr 24-96 hr 96-720 hr EAB 2.98E-4 LPZ 2.29E-5 I.62E-5 7.59E-6 2.57E-6 Summary of Radiological Analyses, Revision 1, 11/03 Page 24 of 81

TABLE 2.8 KR PAVAN INPUTS Meteorological Data Years 1992, 1993, 1994 EAB distances, 16 wind speed directions (m)

S 450 SSW 450 SW 503 WSW 915 W

945 WNW 701 NW 1000 NNW 1000 N

1000 NNE 1000 NE 1000 ENE 1000 E

747 ESE 640 SE 503 SSE 450 LPZ distances (m) 4827 Wind speed considered to be calm (rn/sec)

<0.5 Activity releases ground level Height of wind speed measurement (m) 10 Calm hours input separately from joint frequency distribution Summary of Radiological Analyses, Revision 1, 11/03 Page 25 of 81

Building - wake (m2 )

1071 Wind speed categories 14 Terrain adjustment factors default Summary of Radiological Analyses, Revision 1, 11/03 Paue 26 of 81

Figure 2.8.1 - Spreadsheet Analysis of Low Probability EAB X/Q Data RGE Case I a SW Sector EAB X/Q Curve Fit Based on Low Probability Data 1.60E-03 1.40E-03 1.20E-03 1.OOE-03 8.00E-04 6.00E-04

.U U)

CY (12

-2.39615E-04Ln(x R2 = 9.41381

+ 1.31676E-04 E-01 I

I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

4.00E-04 2.00E-04 O.OOE+00 I

0 0.1 0.2 0.3 0.4 0.5 Percent 0.6 The 0.5 percent EAB valuc:

y:= (-2.39615 4)1n(x) + 1.31676-10 4 y = 2.97764 x 1O-4 The resulting X/Q is rounded to 2.98E-4 sec/ni3.

Summary of Radiological Analyses, Revision 1, 11/03 Page 27 of 81

Figure 2.8.2 - Spreadsheet Analysis of All EAB X/Q Data RGECase la SW Sector EAB XIQ Curve Fit Based on All Data 1.60E-03 1.40E-03 1.20E-03 1.OOE-03 E

CY 8.00E-04 6.00-E04 4.00-E04 2.002-04 O.OOE+00 0

0.5 1

1.5 2

2.5 3

3.5 4

4.5 5

Percent The 0.5 percent EAB value:

y:=- I.92363 1 O-4 I n(x) + 2.03492-1'0 y = 3.36828 x IO' The resulting X/Q is rounded to 3.37E-4 sec/m3 Summary of Radiological Analyses, Revision 1, 11/03 Pa-c 28 of 8

3.0 Iodine Spiking For events where no fuel failure is postulated, iodine spiking is assumed. Two cases of iodine spiking are considered.

1.

Accident Initiated Spike

7.

Pre-Accident Spike 3.1 Accident Initiated Spike The primary system transient causes an iodine spike in the primary system. The appearance rate is based on an equilibrium concentration of 1.0 ftCi/gm Dose Equivalent 1-131. The rate of increase and duration ofthe spike is event dependent. The following inputs are used in the calculation of the appearance rate.

Summary of Radiological Analyses, Revision 1, 11/03 Page 29 of SI

TABLE 3.1 ACCIDENT INITIATED SPIKE INPUTS AND RESULTS Reactor coolant system volume, ft3 rcs 5506 pzr (nominal minus 5% uncertainty) 436 Letdown purification flow rate, gpm 60 + 10%

Reactor coolant iodine concentrations @I,uCi/gram 1131 0.786 of DE I-131,,uCi/gram 1132 4.54 E-3 1133 0.192 1134 1.55 E-4 1135 0.018 Mixed-bed demineralizer DF 100 Identified primary coolant leak rate, gpm 10 Unidentified primary coolant leak rate, gpm I

Primary-to-secondary leak rate, gpd per SG 150 Letdown conditions Pressure, psia 15 Temperature, °F 127 Reactor coolant conditions Pressure, psia 2250 Temperature, °F 559 Appearance rates, Ci/hr 1-131 1.39E+I 1-132 2.49E-1 1-133 4.1OE+0 1-134 1.79E-2 1-135 5.39E-I Summary of Radiological Analyses, Revision 1, 11/03 Page 30 of 81

3.2 Pre-Accident Spike - This assumes a transient has occurred prior to the event and has raised the primary coolant iodine concentration to the maximum full power value. This analysis assumes a value of 60 [tCi/gm DE 1-131. The resulting concentrations and inventories are:

1-131 4.71 E+I MCi/gm 5.88 E+3Ci 1-132 2.72 E-I 3.39 E+I 1-133 1.15 E+I 1.43 E+3 1-134 9.32 E-3 1.16 E+O 1-135 1.07 E+O 1.33 E+2 Summary of Radiological Analyses, Revision 1, 11/03 Page 31 of SI

4.0 General Discussion 4.1 The control room dose calculations use the same X/Q for both prc-isolated outside air and unfiltered inleakage. Pre-isolated outside air is all from the control room intake. Ginna does not have dual air intakes. Unfiltered inleakage may come from doors, penetrations into tile control room envelope, air recirculation/filtration equipment, etc. The source to leakage location for all possible leak points is through other structures first (resulting in torturous paths) or longer source-to-receptor distances. Thus, the leakage point specific X/Q would be greater than that for the control room intake. The control room intake X/Q is assumed to be bounding for all control room dose calculations.

4.2 The nuclide data base used for all calculations is from ORIGEN2 (Reference 12). The nuclides are for a plant-specific representative 18 Month Fuel Cycle at end of life. The (lose significant nuclide concentrations have been slightly increased to produce bounding doses.

4.3 All dose calculations assume the FGRI I and FGRI2 dose conversion factors (References 10 and 11).

4.4 No credit is taken for elemental or methyl iodine removal inside containment by charcoal filters. This is indicated by assuming 0% efficiency as an input parameter. Credit is taken for particulate removal. Particulate removal is done by the inside containment HFPA filters.

4.5 Filter Loading - The RADTRAD code was used to calculate the inside containment H EPA filter particulate loading. The calculation was done for the conditions associated with a LBLOCA. The calculation assumed the filters operate for the duration of the calculation (720 hr.) which essentially removed all particulates from containment. The filter loading was approximately I oz/f12 which is judged to be well within the holding capability of the filters.

Summary of Radiological Analyses, Revision 1, 11/03 Page 32 of 81

5.0 Loss-of-Coolant-Accident 5.1 Analysis The analysis uses the alternate source term (AST) as defined in Reg. Guide 1.183 (Reference 5). The AST assumptions are listed oln Table 5.1 and are consistent with Reg.

Guide 1.183. The analysis is performed with the HABIT code version 1.1 (Reference 6) and the nuclide data base discussed in Section 4.2. The LBLOCA analysis consists of two parts: 1) Containment Leakage and 2) ECCS continuous leakage outside Containment. The resulting doses arc summarized on Table 5.4 The airborne fraction (flashing fraction) used in the analysis is picce-wise time dependent and bounds the values based on sump (ECCS leakage) temperature from a Ginna-specific calculation. The values used in the analysis are illustrated oln Figure 5.1.

The flashing fraction is estimated as follows:

FF

-1exit H,

I14z 1-li Where:

FF = flashing fraction HxI-t = enthialpy of tile relieved fluid (sump conditions)

HI= enthalpy of liquid at 15 psia, saturated H, = enthalpy of vapor at 15 psia, 212'F.

Sump water temperature varies from 260'F at I hr. into the LOCA to 1 700F at 50 hr.'

Sump pH is maintained greater than 7.0 on recirculation.

To determine the airborne fraction, a number of points were selected along the flashing curve, and then the curve was converted into a conservative step function. The value Of each step is approximately 0.01 above the calculated flashed fraction. Even though the curve predicts that the flashed fraction goes to 0 at about 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months , the minimum airborne fraction is maintained at 0.01 out to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (only 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months shown in Fig 5.1). This is done to account for some droplet atomization. However, due to NRC comments during the review of the analysis, the flashing fraction was increased to 0.02 for T >18 hours.

Although these values are not as conservative as the fixed value of 10% suggested in the SRP, they are consistent with the intent of the SRP which is to use a conservative approximation.

Summary of Radiological Analyses, Revision 1, 11/03 Page 33 of SI

5.2 Assumptions A Large Break Loss of Coolant Accident (LBLOCA) occurs inside Containment.

One train of emergency power is assumed to fail. This results in only one train of Containment Recirculation Fan Coolers (CRFCs) operating and one train of Containment Spray.

At 52 minutes Containment Spray is stopped and sump recirculation is started and continues for the duration of the calculation.

At 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the CRFCs are arbitrarily stopped, terminating particulate removal by filtration.

The Control Room parameters are listed on Table 5.3.

The Control Room is assumed isolated at 60 seconds and CREATS is lip and operating at 70 seconds. Isolation from the radiation monitors and/or safety injection would occur well before the 60 seconds assumed in the analysis.

A passive ECCS failure of 50 gpm as identified in the Ginna UFSAR is not assumed in this analysis. However, the ECCS leakage has been increased to 4 gphl.

The analysis uses the source term parameters in Table 5.1 and the Containment Icakagc parameters on Table 5.2.

Control room parameters are shown in Table 5.3 and 5.4.

ECCS Leakage - The analysis assumes a continuous leakage of 4 gph.

5.3 Results The results arc provided in Table 5.5.

Summary of Radiological Analyses, Revision 1, 1 1/03 Pa-c 34 of 81

FIGURE 5.1 - AIRBORNE FRACTION l ~0 0

l_

_l_

l_ _

006 C

4 1

]~~~~~_

002 I

0t.

I o-n-I 0 04 0.01~ ~ ~~

~~~~~~~~~~~~~~~~~2 Time, hours Note: Due to NRC comments during the review process, the assumed flashing fraction was increased to 0.02 for T > 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Summary of Radiological Analyses, Revision 1, 11/03 Page 35 of 81

TABLE 5.1 ALTERNATE SOURCE TERM (REFERENCE 5)

Core Inventor' Fraction Released Into Containment Nuclide Group Gap Release Phase Early In-Vessel l'hase Total' lHalogens 0.05 0.35 0.4 Noble Gases 0.05 0.95 1.0 Alkali Metals 0.05 0.25 0.3 Tellurium 0

0.05 0.05 Ba, Sr 0

0.02 0.02 Noble Metals 0

0.0025 0.0025 Cerium 0

0.0005 0.0005 Lanthanides 0

0.0002 0.0002 TABLE 5.1 ALTERNATE SOURCE TERM (REFERENCE 5)

Timing of LOCA Core Inventory Release Phases Release Phase Onset Duration Gap Release 30 sec 0.5 hr5 Early In-Vessel 0.5 hr 1.3 hr TABLE 5.1 ALTERNATE SOURCE TERM (REFERENCE 5)

Nuclide Groups I lalogens Noble Gases Kr, Xe 4Fractions apply to both containment and ECCS leakage 5The duration of the gap release, specified in Reference 5, is 0.5 hr. The specified start of the gap release is 30 seconds and the end of thle release is 0.5 hr. Thus, the duration of the gap release is modeled as 0.5 hr - 30 sec = 0.492 hr, rather than 0.5 hr.

Sunmmary of Radiological Analyses, Revision 1, 11/03 Page 36 of Sl

TABI,E 5.1 AlTERNATE SOURCE TERM1 (REFERENCE 5)

Nuclide Groups Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc. Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np TABI,E 5.1 AlTERNATE SOURCE TERM (REFERENCE 5)

Nuclide Composition, fraction Forn In Containment Atmosphere In ECC Solution Iodinc elemental 0.0485 0.97 methyl 0.0015 0.03 particulate 0.95 0

All other nuclides particulate 1.0 1.0 Summary of Radiological Analyses, Revision 1, 11/03 Page 37 of 81

TABLE 5.2 CONTAINMENT/ECCS LEAKAGE PARAMTETERS Parameter Value Reactor power, M~wt(including 2% uncertainty) 1550 Containment net free volume, ft3 L.0E6 Containment sprayed fraction 0.78 Containment leak rate, %!day 0-24 hours 0.2

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.1 Containment fan cooler flow and operation number of operating units (per train) 2 flow rate per unit, cfm 30,000 total filtered flow rate, cfm 1IEPA (2 units) 60,000 initiation delay, sec.

50 termination of iodine removal, hours 4

Containment fan cooler iodine removal efficiency, %

Elemental 902 Methyl 502 Particulate 95 Containment injection spray flow rate, gpm (per train) 1300 initiation delay, see 80 termination (end of spray injection), min 52 Iodine and particulate removal by spray, hr-I elemental 20 particulate 3.5' Containment sump volume, ft3 264,700 2Case without iodine removal by carbon assumes 0% efficiency.

'Represents the 1 0th percentile value from the Powers model (Reference 7).

Summary of Radiological Analyses, Revision 1, 11/03 Page 38 of 81

TABLE 5.2 CONTAINMENT/ECCS LEAKAGE PARAMETERS Parameter Valuc ECCS leakage Continuous leakage rate, gal/lhr 4

Start time, hr I

Termination time, hr 720 Airbornc fraction 0-3 hr 0.07 3-8 hr 0.04 8-14 hr 0.03 14-18 hr 0.02

>18 hr 0.02 Atmospheric dispersion X/Q, sec/m3 EAB 0-2 hr 2.98E-4 LPZ 0-8 hr 2.29E-5 8-24 hr 1.62E-5 24-96 hr 7.59E-6 96-720 hr 2.57E-6 Breathing rates, m3/sec EAB & LPZ 0-8 hr 3.47E-4 8-24 hr 1.75 E-4 24-720 hr 2.32E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 39 of 81

TABLE 5.3 CONTROL ROOM PARAMETERS Parameter Value Habitable volime, ft3 36,211 Normal Operating Modc make-up air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal cfficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

Unfiltered in-leakage, cfni 300 Breathing rate, m3/sec 3.47E-4 Occupancy factors 0-24 hr I

24-96 hir 0.6 96-720 hr 0.4 Atmospheric dispersion X/Q sec/m3 Containment ECCS Leakage Leakage 0-2 hr 1.57E-3 3.89E-3 2-8 hr 1.12E-3 2.99E-3 8-24 hr 4.47E-4 9.63E-4 24-96 hr 3.69E-4 8.98E-4 96-720 hr 3.10E-4 8.23E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 40 of 81

Table 5.4 Flow Rate and Iodine Removal Schedule Inleakage Recirculation cfm iodine removal cfm iodine removal Time, hours efficiency, %

efficiency, %

0-0.01672 2200 0/0/0 0

0!0!0 30.0167-0.0194 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98 TABLE 5.5 LBLOCA DOSE

SUMMARY

, REM1 TEDE EAB LPZ Control Room Max. 2-hour 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 720 hour Containment Leakage 3.403 0.741 2.063 ECCS Leakage 0.295 0.178 1.447 Total 3.70 0.919 3.51 Acceptance Criteria 25 25 5

I Elemental/Methyl/Particulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 41 of SI

6.0 Fuel Handling Accident 6.1 Analysis This calculation determines the offsite and Control Room doses (TEDE) for a fuel handling accident (FHA). The analysis uses the alternate source term and accompanying TEDE methodology and conservative control room X/Q values that are calculated with the ARCON96 code. Two cases will be evaluated:

FHA inside Containment

FHA in thle Spent Fuel Pool The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT tiuclide data base as discussed in Section 4.2 are used. Since the release from the FHA is assumed to end after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the dose calculations arc terminated after all contributions arc accounted for; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for LPZ and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the Control Room. The resulting doses are presented on Table 6.4.

6.2 Assumptions Both cases assume the fuel rods in one fuel assembly fail.

Activity from the damaged fuel rods is assumed to be instantaneously released to the pool water.

There is a minimum of 23 feet of water above the fuel.

The activity release rate is independent of the actual ventilation fnow rate. Thle activity release rate is adjusted to ensure all radioactive material that escapes from the reactor cavity or spent fuel pool is released to the environment over a two hour period.

Thle activity from a FHA in Containment is assumed to be released from Containment to the environment via the perimeter seals of the Equipment Hatch roll-up door. No filtration or absorption of iodine is assumed.

The activity from a FHA in the spent fuel pool is assumed to be released from the pool area to the environment via the plant vent. The dose conversion factors from FGRI 1 and 12 arc used (References 10 and 11).

Note that the charcoal filter system for the spent fuel pool area is not ESF or safcty-rclated and the charcoal filter system would be unavailable if a coincident loss ofoffsite power were to occur. The Technical Specifications require use of the system during irradiated fuel movement within the Auxiliary Building to minimize doses. Therefore, the system is credited in the dose analysis.

Summary of Radiological Analyses, Revision 1, 11/03 Page 42 of 81

Tile FHA dose analysis assumptions are listed on Table 6.1. The Control Room assumptions are listed on Table 6.2.

The Control Room is assumed to be isolated within 60 seconds via the radiation monitors.

A comparison of the nuclide concentration in the Control Room intake for the FHA to the radiation monitor response showed a Control Room isolation signal would occur before tile 60 seconds assumed in the calculation.

Fission product inventories and activities released from the SFP are shown in Table 6.3.

Sumlmlary of Radiological Analyses, Revision 1, 11/03 Page 43 of 81

TABLE 6.1 FHA DOSE ANALYSIS ASSUMPTIONS Parameter Value Reactor power, Mwt(including 2% uncertainty) 1550 Power Peaking Factor 1.75 Number of damaged fuel assemblies 1

Fission product inventory in damaged assemblies after decay Values shown in Table 6.3 Time after reactor shutdown, hr 100 Fuel rod gap fractions 1-131 0.08 other halogens 0.05 Kr-85 0.1 other noble gases 0.05 alkali metals 0.12 Iodine species above water elemental iodine 0.57 organic iodide 0.43 Pool DF elemental iodine 500 organic iodide 1

particulate 00 Overall Pool DF 200 Containment net free volume, ft3 1E6 Exhaust flow rate, cfm 7.68E4 Duration of activity release, hr 2

Iodine removal efficiency Containment (all iodine forms) 0 Fuel Pool elemental iodine 0.9 organic iodide 0.7 Summary of Radiological Analyses, Revision 1, 11/03 Page 44 of 81

TABLE 6.1 FHA DOSE ANALYSIS ASSUMPTIONS Parameter Value Atmospheric dispersion, X/Q, sec/M3 EAB 0-2 hr 4.8 E-4 LPZ 0-8 hr 3.0 E-5 Breathing rate, m3/sec EAB &LPZ 0-8 hr 3.47 E-4 TABLE 6.2 CONTROL ROOM PARAMETERS Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 Flow rate, cfm 6000-10%

Unfiltered in-leakage, cfm 300 Breathing rate, m3/sec 3.47 E-4 Occupancy factor 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion, X/Q, sec/M3 FHA Containment FHA Spent Fuel Pool 0-2 hr 5.64 E-3 1.79 E-3 2-8 hr 4.69 E-3 1.15 E-3 8-24 hr 1.66 E-3 4.95 E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 45 of 81

Flow Rate and Iodine Removal Schedule Inleakage Recirculation Time, hours cfm iodine removal cfm iodine removal efficiency, %

efficiency, %'

0 - 0.01672 2200 0/0/0 0

0/0/0 30.0167 - 0.0194 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98

'Elemental/Methyl/Particulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 46 of 81

TABLE 6.3 FISSION PRODUCT INVENTORY AND ACTIVITY RELEASED FROM POOL Total Core Activity -

Activity 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Core Gap Released

decay, Damage Fraction Peaking Overall from Pool, Nuclide Ci(A)

Fraction (F)

(G)

Factor (P)

Pool DF Ci (A)

I-131 2.98E+07 0.008264 0.08 1.75 200 1.76E+02 I-132 2.52E+07 0.008264 0.05 1.75 200 9.29E+01 I-133 3.12E+06 0.008264 0.05 1.75 200 1.15E+01 I-134 0.OOE+00 0.008264 0.05 1.75 200 0.OOE+00 I-135 2.23E+03 0.008264 0.05 1.75 200 8.22E-03 Kr-85m 2.15E+00 0.008264 0.05 1.75 1 1.55E-03 Kr-85 4.98E+05 0.008264 0.1 1.75 1 7.20E+02 Kr-87 4.58E-17 0.008264 0.05 1.75 1 3.31 E-20 Kr-88 7.48E-04 0.008264 0.05 1.75 1 5.41E-07 Xe-131m 4.42E+05 0.008264 0.05 1.75 1 3.20E+02 Xe-133m 1.1OE+060 0.008264 0.05 1.75 1 7.95E+02 Xe-133 5.71E+07 0.008264 0.05 1.75 1 4.13E+04 Xe-135m 3.57E+02 0.008264 0.05 1.75 1 2.58E-01 Xe-135 1.09E+05 0.008264 0.05 1.75 1 7.88E+01 Core damage fraction is 1/121 = 0.008264.

121.

The total number of fuel assemblies in the core is The activity released from the pool (A) is calculated as'follows: Note 2% added to iodine in Section 4.2.

Ac*F*G*P A =

DF Summary of Radiological Analyses, Revision 1, 11/03 Page 47 of 81

TABLE 6.4 FHA, DOSE, REM, TEDE EAB Max -2 hr LPZ, 2 hr Control Room, 24 hr FHA - inside Containment via roll-up door 1.12 0.07 1.18 FHA - Spent Fuel Pool 0.31 0.019 0.089 Acceptance Criteria 6.3 6.3 5

Summary of Radiological Analyses, Revision 1, 11/03 Page 48 of 81

7.0 Main Steam Line Break 7.1 Analysis This calculation determines the offsite and Control Room doses (TEDE) for the Main Steam Line Break (MSLB) outside the Containment. The analysis uses the alternate source term and the accompanying TEDE methodology and conservative control room X/Q values that are calculated with the ARCON96 code. The MSLB analysis includes the following cases:

MSLB with accident initiated iodine spike

MSLB with pre-accident iodine spike The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT nuclide data base as discussed in Section 4.2 are used. No fuel failures are postulated for the MSLB.

7.2 Assumptions The purpose of this analysis is to calculate the steam releases from the faulted and intact Steam Generator (SG) during a steam line break to the atmosphere. Therefore, breaks inside Containment are non-applicable.

Because of an augmented inspection program, breaks between the Containment penetrations and inside the Intermediate Building are limited to connection pipes only with the largest pipe being 6" (UFSAR Section 3.6.2.4.5.2). Larger pipe breaks can only be postulated downstream of the Intermediate Building, i.e., inside the Turbine Building.

Therefore, the break is assumed to occur in the 36" header inside the Turbine Building.

This is the largest pipe break that can occur outside Containment. The break area is limited to 1.4 ft2 because of a flow restrictor in the SG outlet nozzle.

The scenario consists of a header break. The single failure is assumed to be a failure of the main steam isolation valve on the faulted SG. Initially the break is fed by both SGs.

Following steam line isolation, the break is fed only by the faulted SG. At approximately 10 minutes the faulted SG is isolated by operator action. The intact SG is then used for cooldown, where steam is released to the atmosphere through the intact SG Atmospheric Relief Valve (assumed to be 8 hr.) until the releases are stopped.

A primary to secondary leakage of one gpm to each SG is assumed for the duration of the event (8 hr). The faulted SG is assumed to be dry at 10 minutes. and remain dry for the event. The intact SG is isolated from the break within the first minute and auxiliary feedwater maintains SG level for the duration of the event.

Summary of Radiological Analyses, Revision 1, 11/03 Page 49 of 81

All of the initial iodine inventory in the faulted SG is assumed released to the environment by 10 minutes. The iodine from the primary-to-secondary leakage into the faulted SG is released directly to the environment with no credit for retention. The initial iodine inventory in the intact SG is mixed with the primary-to-secondary leakage into the SG and released to the environment assuming an iodine partition of 100. The steam release from the intact SG is based on a LOFTRAN simulation of the MSLB followed by an energy balance to simulate the cooldown to RHR conditions. All noble gas activity carried over to the SGs is assumed to be immediately released to the environment.

Initially the Control Room HVAC is operating normally with a nominal 2000 cfm of makeup air. Isolation is assumed to occur at 60 sec and CREATS is operating at 70 sec assuming a nominal 6000 cfm recirculation flow. Since isolation is caused by a safety injection signal, the Control Room would be isolated well before the 60 sec. assumed in the analysis. Following isolation, 300 cfm of unfiltered inleakage is assumed for the duration of the calculation.

The releases from the steam break are assumed to stop at 8 hr. The Control Room calculation is continued until 24 hr to ensure all dose contributions are accounted for.

Accident - Initiated Iodine Spike: A spike factor of 500 with a duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is assumed. The initial appearance rates are listed on Table 3.1.

Pre-Accident Iodine Spike: The iodine concentrations are based on 60,Ci/gm DE I-131 and listed in Section 3.2.

Additional assumptions are listed in Table 7.1.

The Control Room parameters are listed on Table 7.2 and 7.3.

7.3 Results The results for the MSLB are shown in Table 7.4.

Summary of Radiological Analyses, Revision 1, 11/03 Page 50 of 81

TABLE 7.1 MSLB DOSE ANALYSIS ASSUMPTIONS Parameter Value Reactor power, Mwt(including 2% uncertainty) 1550 Initial reactor coolant activity, pre-accident iodine spike iodine /2Ci/gm of D.E. I-131 60 noble gas fuel defect level, %

1.0 Initial reactor coolant activity, accident initiated iodine spike iodine jsCi/gm of D.E. I-13 1 1.0 noble gas fuel defect level, %

1.0 Concurrent iodine spike factor 500 Duration of concurrent iodine spike, hours 8

Initial secondary coolant iodine activity iodine /2Ci/gm of D.E. I-13 1 0.1 Concentration Ci I-131 4.57 E+O I-132 2.64 E-2 I-133 1.12 E+O I-134 9.04 E-4 1-135 1.03 E-1 Primary-to-secondary leakage (post accident) to SGs leak rate (cold conditions) per SG, gpm 1

duration of leakage, hours 8

Mass of primary coolant, gm 1.247 E+8 Initial mass of secondary coolant, gm faulted SG 5.817 E+7 intact SG 5.817 E+7 Summary of Radiological Analyses, Revision 1, 11/03 Page 51 of 81

TABLE 7.1 MSLB DOSE ANALYSIS ASSUMPTIONS Y.

Parameter Value Steam Releases faulted SG 0 - 610 sec 128,237 lb 610sec-8hr 0 lb intact SG 0 - 610 sec 37,780 lb 610 sec - 8 hr 755,097 lb Primary to Secondary Leakage 1 gpm per SG Steam generator iodine partition coefficients (mass-based)

Activity release from faulted SG elemental 1

methyl 1

Activity release from intact SG elemental 100 methyl 1

Noble gas, all SG 1

Iodine fractions assumed in the reactor coolant and SG water elemental iodine 0.97 organic iodide 0.03 Atmospheric dispersion X/Q sec/M3 EAB 0-2 hr 4.8E-4 LPZ 0-8 hr 3.01E-5 Breathing rate m3/sec EAB & LPZ 0-8 hr 3.47 E-4 8-24 hr 1.75 E-4 24-720 hr 2.32 E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 52 of 81

TABLE 7.2 CONTROL ROOM PARAMETERS Parameter Value Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

Unfiltered in-leakage, cfm 300 Breathing rate, m3/sec 3.47 E-4 Occupancy factor 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion, X/Q, sec/M3 0-2 hr 2.57 E-3 2-8 hr 1.92 E-3 8-24 hr 8.08 E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 53 of 81

Table 7.3 Flow Rate and Iodine Removal Schedule Inleakage Recirculation Time, hours cfm iodine removal cfm iodine removal efficiency, %

efficiency, %

0 - 0.01672 2200 0/0/0 0

0/0/0 30.0167 - 0.0194 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98 TABLE 7.4 RESULTS FOR MAIN STEAM LINE BREAK EAB Max - 2 hr LPZ, 8 hr Control Room 24 hr TEDE TEDE TEDE Accident Initiated Iodine Spike 1.05 0.15 0.64 Acceptance Criteria 2.5 2.5 5

-Pre-Accident Iodine Spike 0.15 0.03 0.18 Acceptance Criteria 25 25 5

'Elemental/Methyl/particulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 54 of 81

8.0 Steam Generator Tube Rupture (SGTR) 8.1 Analysis This calculation determines the offsite and Control Room doses for the SGTR accident.

The analysis uses alternate source term and accompanying TEDE methodology and conservative Control Room X/Q values, that are calculated with the ARCON96 code. The SGTR analysis includes the following cases:

SGTR with accident-initiated spike

SGTR with pre-accident iodine spike The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT nuclide data base discussed in Section 4.2 are used.

8.2 Assumptions Input parameters are listed in Table 8.1 below.

The break flow and steam release data for the ruptured SG, and steam release data for the intact SG is taken from the analysis discussed in Section 15.6 of Reference 3 and listed on Table 8.2.

The Control Room parameters are listed on Table 8.3.

Control Room isolation is assumed at 6 minutes which bounds the safety injection signal generation time for the Reference 3, Section 15.6 SGTR. The ARV is the source point for the Control Room X/Q.

Accident-Initiated Iodine Spike:

The initial appearance rates are listed on Table 3.1. The input parameters are listed on Table 8.1 and the results are presented on Table 8.5. The dose calculations are terminated after all dose contributions are accounted for.

Pre-Accident Iodine Spike:

The iodine concentrations are based on 60,pCi/gm DE I-131 and listed in Section 3.2. The input parameters are listed on Table 8.1 and results are presented on Table 8.5. The dose calculations are terminated after all dose contributions are accounted for.

Summary of Radiological Analyses, Revision 1, 11/03 Page 55 of 81

TABLE 8.1 SGTR DOSE ANALYSIS ASSUMPTIONS Parameter Value Reactor power, MwT(including 2% uncertainty) 1550 Initial reactor coolant activity, pre-accident iodine spike iodine, jiCi/gm of DE I-131 60 noble gas fuel defect level, %

1.0 Initial reactor coolant activity, accident initiated iodine spike iodine, ci/gm of DE I-131 1.0 noble gas fuel defect level, %

1.0 Concurrent iodine spike factor 335 Duration of concurrent iodine spike, hours 8

Initial secondary coolant iodine activity, IiCi/gm of DE I-13 1 0.1 Primary-to-secondary leakage to intact SG leak rate (cold conditions) 150 gal/day duration of leakage, hours 8

Mass of primary coolant, gm 1.247x108 Initial mass of secondary coolant, gm faulted SG 3.27x107 intact SG 3.27x10' Steam generator elemental iodine partition coefficients (mass-based)

Activity release from faulted SG via boiling of bulk water 100 via flashed break flow 1.0 Activity release from intact SG 100 Steam generator partition coefficient for organic iodide and noble gas release 1.0 Iodine species assumed in the reactor coolant and SG water elemental iodine 0.97 organic iodide 0.03 Summary of Radiological Analyses, Revision 1, 11/03 Page 56 of 81

TABLE 8.1 SGTR DOSE ANALYSIS ASSUMPTIONS Parameter Value Atmospheric dispersion, X/Q, sec/M3 EAB 0-2 hr 4.8 E-4 LPZ 0-8 hr 3.0 E-5 Breathing Rates, m3/sec EAB & LPZ 0-8 hr 3.47E-4 8-24 hr 1.75E-5 Table 8.2 Steam Releases and Rupture Flow Time periods, seconds Mass, 1000 Ibm 0-49 sec 49 sec-3492 sec-2 2 hrs - 8 hrs 3492 sec hours Ruptured SG to:

Condenser' 45.5 Atmosphere 62.4 0

31.6 Intact SG to:

Condenser 45.2 Atmosphere 60.0 147.5 459.9 Rupture flow 2.9 107.4 I

I I

49 sec:

Reactor trip.

3492 sec:

SG and RC pressures are equal, rupture flow is terminated.

8 hrs:

RHR operating conditions are achieved, steaming to the environment is terminated.

'The analysis conservatively treats steam released to the condenser the same as a direct release to the atmosphere, i.e., elemental iodine partition is 100.

Summary of Radiological Analyses, Revision 1, 11/03 Page 57 of 81

TABLE 8.3 CONTROL ROOM PARAMETERS Parameter Value Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

unfiltered in-leakage, cfm 300 Breathing rate, m3/sec 3.47E-4 Occupancy factor 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion, X/Q, sec/M3 0-2 hr 3.66E-3 2-8 hr 2.49E-3 8-24 hr 1.07E-3 Summary of Radiological Analyses, Revision 1, 11/03 Page 58 of 81

Table 8.4 Flow Rate and Iodine Removal Schedule Inleakage Recirculation Time, hours iodine removal iodine removal cfm efficiency, %

cfm efficiency, %'

0_0.12 2200 0/0/0 0

0/0/0 30.1-0.103 300 0/0/0 0

0/0/0

>0.103 300 0/0/0 5400 90/70/98 TABLE 8.5 RESULTS FOR SGTR EAB Max 2 hr LPZ, 8 hr Control Room 24 hr Accident Initiated Iodine Spike (TEDE) 0.22 0.017 0.14 Acceptance Criteria 2.5 2.5 5

Pre-Accident Iodine Spike (TEDE) 0.71 0.051 0.88 Acceptance Criteria 25 25 5

'Elemental/MethyllParticulate 20 to 360 seconds 3360 to 370 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 59 of 81

9.0 Locked Rotor Accident This calculation determines the offsite and Control Room doses for the LR accident. The analysis uses alternate source term and accompanying TEDE methodology and conservative Control Room X/Q values, that are calculated with the ARCON96 code. The LR analysis includes the following case:

Primary-to-secondary leakage with SG activity releases The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT nuclide data base discussed in Section 4.2 are used.

9.1 Assumptions Input parameters are listed in Table 9.1 and 9.2 below.

It is conservatively assumed 100% of the fuel rods experience DNB and are therefore assumed to release their gap activity into the reactor coolant system.

The initial reactor coolant iodine activity is based on a pre-accident spike discussed in Section 3.2. The concentrations are based on 60 uCi/gm of DE I-131. The noble gas activity is based on 1% fuel defects.

The initial secondary coolant iodine activity is based on 0.1 uCi of DE I-131.

The assumed post-accident primary-to-secondary leak rate is 500 gal/day per SG. This bounds the current limit of 144 gpd/SG and a future Technical Specification limit of 150 gpdISG.

A partition coefficient of 100 is assumed for elemental iodine in the secondary coolant. No partitioning is assumed for organic iodine or noble gas. No particulates are assumed to be released to the atmosphere with the secondary side steam.

The steam release from the SGs is based on a LOFTRAN simulation of the LR followed by an energy balance to simulate the cooldown to RHR conditions. RHR System is assumed to be placed into service for heat removal 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the initiation of the LR.

Initially the Control Room HVAC is operating normally with a nominal 2000 cfm of makeup air. Isolation is assumed to occur at 60 sec. via the radiation monitors. A comparison of the nuclide concentration in the Control Room intake for the LR to the radiation monitor response showed a Control Room isolation signal would occur before the 60 sec. assumed in the calculations. CREATS is assumed operating at 70 sec. assuming a nominal 6000 cfm recirculation flow.

Summary of Radiological Analyses, Revision 1, 11/03 Page 60 of 81

The EAB and LPZ X/Qs are the new values calculated by K R PAVAN as discussed in Section 2.8.2.

Summary of Radiological Analyses, Revision 1, 11/03 Page 61 of 81

TABLE 9.1 LR Dose Analysis Assumptions Parameter Value Reactor power, Mwt(including 2% uncertainty) 1550 Failed Fuel, %

100 Initial reactor coolant activity, pre-accident iodine spike iodine, uCi/gm of DE I-131 60 noble gas fuel defect level, %

1.0 Initial secondary coolant iodine activity, uCi/gm of DE 1-131 0.1 Primary-to-secondary leakage (post accident) to SGs leak rate (cold conditions per SG, gpd 500 duration of leakage, hours 8

Mass of primary coolant, gm 1.247x108 Initial mass of secondary coolant in 2 SGs, gm 8.501E+7 Steam Releases (2 SGs), lb 0-10 min.

54,620 10-30 min.

14,446 0.5-8 hr.

685,229 Steam generator iodine partition coefficients (mass-based) elemental 100 methyl (organic) 1 Iodine fractions in the reactor coolant and SG water elemental iodine 0.97 methyl (organic) iodide 0.03 Atmospheric dispersion X/Q sec/m3 EAB 0-2 hr 2.98E-4 LPZ 0-8 hr 2.29E-5 Summary of Radiological Analyses, Revision 1, 11/03 Page 62 of 81

TABLE 9.1 LR Dose Analysis Assumptions Parameter Value Breathing rate m3/sec EAB & LPZ 0-8 hr 3.47E-4 8-24 hr 1.75E-4 TABLE 9.2 CONTROL ROOM PARAMETERS Parameter Value Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

Unfiltered in-leakage, cfm 300 Breathing rate, m3/sec 3.47 E-4 Occupancy factor 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion, X/Q, sec/M3 0-2 hr 3.66 E-3 2-8 hr 2.49 E-3 8-24 hr 1.07 E-3 Summary of Radiological Analyses, Revision 1, 11/03 Page 63 of 81

Table 9.3 Flow Rate and Iodine Removal Schedule Inleakage Recirculation Time, hours cfm iodine removal cfm iodine removal efficiency, %

efficiency, %'

0 - 0.01672 2200 0/0/0 0

0/0/0 30.0167 - 0.0194 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98 TABLE 9.4 RESULTS FOR LOCKED ROTOR EAB Max - 2 hr LPZ, 8 hr Control Room 24 hr TEDE TEDE TEDE Elemental iodide 1.150 0.209 1.370 Methyl iodide 1.022 0.254 2.115 Noble gas 0.578 0.090 0.232 Total 2.750 0.553 3.717 Acceptance criteria 2.5 2.5 5

'Elemental/Methyl/particulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 64 of 81

10.0 Rod Ejection Accident This calculation determines the offsite and Control Room doses (TEDE) for Rod Ejection Accident (REA). The analysis uses the alternate source term and the accompanying TEDE methodology and conservative control room X/Q values that are calculated with the ARCON96 code. The REA analysis includes the following cases:

Containment leakage

Primary-to-secondary leakage with SG activity release.

Doses are calculated for the following receptors:

Exclusion Area Boundary (EAB), maximum 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose

Outer boundary of the Low Population Zone (LPZ), 30 day dose (8 hr for secondary side transport)

Control Room, 30 day dose (24 hr for secondary side transport)

The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT nuclide data base described in Section 4.2 are used. Ten percent of the core is assumed to fail. This is based on a Ginna specific calculation (Reference 3, Section 15.4.5.3.5). The release fraction used in the analysis is the product of the core damage, the peaking factor, and the gap fraction. The input parameters are listed on Table 10.1.

10.1 Containment Leakage The containment leakage calculation assumes the gas activity is instantaneously released from the core to containment atmosphere. No credit is taken for removal of elemental or methyl iodine by the CRFC charcoal filters. The CRFCs only remove particulate iodine by the associated HEPA filters. The CRFCs are assumed to be operating at 53 seconds based on a 3 inch SBLOCA. The CRFCs are arbitrarily terminated after four hours since there is no longer a significant particulate concentration.

Containment spray is assumed not to actuate for the REA. Particulate removal is assumed by natural deposition. The removal coefficient is based on the correlations provided in Reference 8, Table 2.2.2.1-1. The first is for the time period 0 to 0.5 hr and the second is for 0.5 to 1.8 hr. The 10th percentile is the most conservative (smallest removal rate) and is used in this calculation. Only the smallest value is used and is held constant for the duration of the calculation.

10.2 Primary-to-Secondary Leakage The initial reactor coolant iodine activity is based on a pre-accident spike discussed in Section 3.2. The concentrations are based on 6OgCi/gm of DE I-13 1. The noble gas activity is based on 1% fuel defects. Gap (10% failed fuel rods) activity is released instantaneously and homogeneously mixed in the reactor coolant. The activity release Summary of Radiological Analyses, Revision 1, 11/03 Page 65 of 81

fraction is the product of core damage, the peaking factor, and gap fraction.

The initial secondary coolant iodine activity is based on 0.1 jIi of DE I-131.

The assumed post-accident primary-to-secondary leak rate is 500 gal/day per SG. This bounds the current limit of 144 gpd/SG and a future Technical Specification limit of 150 gpd/SG.

A partition coefficient of 100 is assumed for elemental iodine in the secondary coolant. No partitioning is assumed for organic iodine or noble gas. No particulates are assumed to be released to the atmosphere with the secondary side steam.

The steam release from the SGs is based on a LOFTRAN simulation of the REA followed by an energy balance to simulate the cooldown to RHR conditions. RHR system is assumed to be placed into service for heat removal 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the initiation of the REA.

Initially the Control Room HVAC is operating nornally with a nominal 2000 cfm of makeup air. Isolation is assumed to occur at 60 sec. via the radiation monitors. A comparison of the nuclide concentration in the Control Room intake for the REA to the radiation monitor response showed a Control Room isolation signal would occur before the 60 sec. assumed in the calculations. CREATS is assumed operating at 70 sec. assuming a nominal 6000 cfm recirculation flow.

Summary of Radiological Analyses, Revision 1, 11/03 Page 66 of 81

TABLE 10.1 REA CONTAINMENT PARAMETERS Parameter Value Reactor power, MwT(including 2% uncertainty) 1550 Failed Fuel, % of core 10 Gap fraction 0.10 Peaking factor, fraction 1.75 Initial primary coolant activity iodine 60liCi/gm of DE I-131 noble gas 1% fuel defects Iodine forms particulate 0.95 elemental 0.0485 organic 0.0015 Containment net free volume, ft3 I0E6 Containment leak rate, %/o/day 0-24 hr 0.2

>24 hr 0.1 Containment fan cooler flow and operation number of operating units 2

flow rate per unit, cfm 30,000 total filtered flow rate, cfm HEPA (2 units) 60,000 initiation delay CRFCs (HEPA) 53 sec termination of particulate iodine removal, hours 4

Containment fan cooler iodine removal efficiency, %

elemental 0

methyl 0

particulate 95 Natural deposition coefficient, 1/hr 0.023 Summary of Radiological Analyses, Revision 1, 11/03 Page 67 of 81

TABLE 10.1 REA CONTAINMENT PARAMETERS Parameter Value Atmospheric dispersion, X/Q, sec/M3 EAB 0-2 hr 4.8 E-4 LPZ 0-8 hr 3.0 E-5 8-24 hr 2.1 E-5 24-96 hr 8.6 E-6 96-720 hr 2.5 E-6 Breathing rate, m3/sec EAB & LPZ 0-8 hr 3.47 E-4 8-24 hr 1.75 E-4 24-720 hr 2.32 E-4 TABLE 10.2 PARAMETERS FOR REA SECONDARY SIDE ACTIVITY RELEASE Parameter Value Reactor power, MwT(including 2% uncertainty) 1550 Failed fuel, % of core 10 gap fraction 0.10 peaking factor, fraction 1.75 Initial secondary coolant iodine activity, ci/gm of DE 1-131 0.1 Primary-to-secondary leakage leak rate, gpd per SG 500 duration, hr 8

Mass of primary coolant, gm 1.247E8 Initial mass of secondary coolant, gm per 2 SGs 8.5E7 Steam released from S.S. to environment, gm/min 0-10 min 2.478E6 10-30 min 3.276E5 30 min - 8 hr 6.907E5 Summary of Radiological Analyses, Revision 1, 11/03 Page 68 of 81

TABLE 10.2 PARAMETERS FOR REA SECONDARY SIDE ACTIVITY RELEASE Steam generator iodine partition coefficient (mass-based) elemental 100 methyl 1

Iodine species assumed in the SG water elemental iodine 0.97 methyl iodide 0.03 TABLE 10.3 CONTROL ROOM PARAMETERS Habitable volume, ft3 36,211 Normal operating Mode make-Oup air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

Unfiltered in-leakage,cfin 300 Breathing rate, m3/sec 3.47E-4 Occupancy factors 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion X/Q, sec/m3 Containment Leakage ARV 0-2 hr 1.57E-3 3.66E-3 2-8 hr 1.12E-3 2.49E-3 8-24 hr 4.47E-4 1.07E-3 24-96 hr 3.69E-4 96-720 hr 3.10E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 69 of 8 1

Table 10.4 Control Room Flow Rate and Iodine Removal Schedule for REA Inleakage Recirculation Time, hours cfm iodine removal cfm iodine removal efficiency, %

efficiency, %I 0-0.01672 2200 0/0/0 0

0/0/0 0.0167-0.01943 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98 TABLE 10.5 REA DOSE SUMMATION, rem, TEDE EAB, max -2 hour LPZ, 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months Control Room, 720 (CNMT), 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months hours (CNMT), 24 (secondary side) hours (secondary side)

Containment 2.859E-01 4.825E-02 1.31 lE-0I Leakage Secondary Side, 4.539E-01 6.759E-02 3.196E-01 Elemental Iodine Secondary Side, 3.263E-01 4.125E-02 8.132E-02 Noble Gas Secondary Side, 4.032E-01 8.244E-02 5.043E-01 Methyl Iodide TOTAL 1.47E+00 2.40E-01 1.04E+00

'Elemental/Methyl/Particulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 70 of 81

11.0 Tornado Missile in Spent Fuel Pool 11.1 This calculation determines the offsite and Control Room doses (TEDE) for a tornado missile accident (TMA). The analysis uses the alternate source term and accompanying TEDE methodology and conservative Control Room X/Q values calculated as discussed in Section 2.7.

The AST defined in Reference 5 is used. The HABIT code (Reference 6) and HABIT nuclide data base as discussed in Section 4.2 are used. The analysis assumes 9 fuel assemblies are damaged (5 fuel assemblies decayed for 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and four fuel assemblies decayed for 60 days) based on the size of a telephone pole missile. The nuclide inventory in the damaged assemblies is estimated by applying a power peaking factor of 1.75 to the average assembly inventory. Activity from the damaged assemblies is assumed to be instantaneously released to the pool water. After applying decontamination factors of the pool water, the resulting elemental and organic fractions above the water are 0.57 and 0.43. The activity above the pool is assumed to be released to the environment. No iodine removal is assumed. Reference 5 suggests using a two-hour activity release. Since the duration of the tornado is uncertain, and may be less than two hours, two cases were run.

Case 1)

All activity was released over two hours. The activity released over the first hour was at tornado conditions. The activity released over the second hour was at normal atmospheric conditions.

Case 2)

All activity was released over one hour at tornado conditions. Case 2 resulted in slightly higher Control Room doses.

The TMA dose analysis assumptions are listed on Table 11.1. The activity released from the pool is listed on Table 11.5. The Control Room assumptions are listed on Table 11.2. The Control Room is assumed to be isolated within 60 seconds via the radiation monitors. A comparison of the nuclide concentration in the Control Room intake for the TMA to the radiation monitor response showed a Control Room isolation signal would occur before the 60 seconds assumed in the calculation. The resulting doses are presented on Table 11.4. Since the release from the TMA is assumed to end after one or two hours, the dose calculations are terminated after all contributions are accounted for; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for EAB and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the Control Room.

Summary of Radiological Analyses, Revision 1, 11/03 Page 71 of 81

TABLE 11.1 TMA DOSE ANALYSIS ASSUMPTIONS Parameter Value Reactor power, MwT(including 2% uncertainty) 1550 Power Peaking Factor 1.75 Number of damaged fuel assemblies Hot 5

Cold 4

Fission product inventory in damaged assemblies after decay Values calculated Time after reactor shutdown hot assemblies 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months cold assemblies 60 days Fuel rod gap fractions I-131 0.08 other halogens 0.05 Kr-85 0.1 other noble gases 0.05 Iodine species above water elemental iodine 0.57 organic iodine 0.43 Pool DF elemental iodine 500 organic iodide 1

particulate co Overall Pool DF 200 Exhaust flow rate, cfm 1-hour activity release 1.545E5 2 -hour activity release 7.685E4 Iodine removal efficiency for all forms 0

Atmospheric dispersion, X/Q, sec/m3 EA.

Tornado conditions 1.74E-6 Normal conditions 4.8E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 72 of 81

TABLE 11.1 TMA DOSE ANALYSIS ASSUMPTIONS Parameter Value Breathing rate, m3/sec I

EA. 0-8 hr 3.47E-4 TABLE 11.2 CONTROL ROOM PARAMETERS Parameter Value Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode Recirculating air iodine removal efficiency, %

elemental 90 methyl 70 particulate 98 flow rate, cfm 6000-10%

Unfiltered in-leakage, cfm 300 Breathing rate, m3/sec 3.47E-4 Occupancy factor 0-24 hr 1

24-96 hr 0.6 96-720 hr 0.4 Atmospheric dispersion, X/Q, sec/M3 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months release 4.36E-5 4.36E-5 (Case 1) 1.45E-3

_ (Case 2)

Summary of Radiological Analyses, Revision 1, 11/03 Page 73 of 81

Table 11.3 Flow Rate and Iodine Removal Schedule Inleakage Recirculation Time, hours cfm iodine removal cfm iodine removal efficiency, %

efficiency, %I 0-0.01672 2200 0/0/0 0

0/0/0 0.0167-0.01943 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 5400 90/70/98 TABLE 11.4 TMA DOSE, Rem, TEDE TMA EAB, Max - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s4 Control Room, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1-hour release 2.01 E-2 5.87 E-2 2-hour release 7.41 E-2 5.44 E-2 Acceptance Criteria 6.3 5

'Elemental/Methyl/Particulate 20 to 60 seconds 360 to 70 seconds 4The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months LPZ dose is bounded by the 2-hour dose at the EAB, as such, only the EAB dose is evaluated.

Summary of Radiological Analyses, Revision 1, 11/03 Page 74 of 81

TABLE 11.5 Spent Fuel Pool Activity l Aloo l A,,d T

n Xgap Xpeak DF lAmcud 1-131 2.98E+07 2.432E+05 121 0.08 1.75 200 8.676E+02 1-132 2.52E+07 0.OOOE+00 121 0.05 1.75 200 4.557E+02 1-133 3.12E+06 1.261E-13 121 0.05 1.75 200 5.640E+01 1-134 0.00E+00 0.00E-0 121 0.05 1.75 200 0.0 1-135 2.23E+03 0.00 121 0.05 1.75 200 4.028E-02 Kr-85m 2.15E+00 0.00 121 0.05 1.75 1

7.774E-03 Kr-85 4.98E+05 4.934E+05 121 0.1 1.75 1

6.456E+03 Kr-87 4.58E-17 0.0 121 0.05 1.75 1

1.656E-19 Kr-88 7.48E-04 0.0 121 0.05 1.75 1

2.705E-06 Xe-131m 4.42E+05 3.084E+04 121 0.05 1.75 1

1.687E+03 Xe-133m 1.1OE+06 2.416E-02 121 0.05 1.75 1

3.977E+03 Xe-133 5.71E+07 3.662E+04 121 0.05 1.75 1

2.066+05 Xe-135m 3.57E+02 0.0 121 0.05 1.75 1

1.291E+00 Xe-135 1.09E+05 0.0 121 0.05 1.75 1

3.941E+02 Xe-138 0.00E+00 0.0 121 0.05 1.75 1

0.0 Total core activity @ 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months (A 1o): Ci Total core activity @ 60 days (A6w): Ci Core assemblies (n)

Gap Fraction (Xgap)

Peaking factor (Xpeak)

Overall pool DF Summary of Radiological Analyses, Revision 1, 11/03 Page 75 of 81

Activity released from the pool to the environment (Ar,¢,,,.d):

A.t Ah

=o= A 100 *5 n

Acold Acold=

A60d *13 n

Atotal A4otali Ahol + A.id A~tota *

X Atotal

=

DF Summary of Radiological Analyses, Revision 1, 11/03 Page 76 of 81

12.0 Waste Gas Decay Tank Rupture 12.1 Analysis This analysis calculates the Control Room and off-site doses for a release of a Gas Decay Tank (GDT) into the Auxiliary Building Atmosphere 12.2 Assumptions The source term is 100,000 Ci of equivalent Xe-133. The assumed source will be 100,000 Ci of actual Xe-133.

Activity, from the ruptured tank, is released considering two different release rates Released to the environment over 2-hours. The flow rate corresponds to a 2-hour activity release.

A puff release assuming a 5 second release duration.

The 2-hour activity release assumption is consistent with that of the Fuel Handling Accident.

Activity from the ruptured tank is released into the Auxiliary Building and assumed to diffuse from the building to the environment. The Control Room dose calculation 'will use x/Qs for the Auxiliary Building area source.

The dose conversion factor for Xe-133, contained in HABIT library MLWRGE.1465.pwr, will be used. The DCF values in this library were derived from FGR 11 and 12.

Table 12.1 Atmospheric Dispersion (sec/m 3)

EAB lLPZ l

I 2.98E-4 2.29E-5 Control Room l0-2 hours l2-8 hours l 8-24hours l24-96 l96-720 l

3.89E-3 2.99E-3 9.63E-4 8.98E-4 8.23E-4 Summary of Radiological Analyses, Revision 1, 11/03 Page 77 of 81

Table 12.2 Control Room Parameters Parameter Value Habitable volume, ft3 36,211 Normal Operating Mode make-up air flow rate, cfm 2000+10%

Accident Operating Mode This analysis considers only noble gas, as such, iodine removal efficiencies and recirculation flow have no effect on the calculated doses.

300 Unfiltered in-leakage, cfm Table 12.3 Flow Rate and Iodine Removal Schedule Time, hours Inleakage Recirculation cfm iodine removal cfm iodine removal efficiency, % (

efficiency, %

0 - 0.01672 2200 0/0/0 0

0/0/0 0.0167 - 0.01943 300 0/0/0 0

0/0/0

>0.0194 300 0/0/0 0

0/0/0 Note: The isolation and recirculation times, shown above, are consistent with those provided for other accidents (excluding SGTR).

'Elemental/MethyllParticulate 20 to 60 seconds 360 to 70 seconds Summary of Radiological Analyses, Revision 1, 11/03 Page 78 of 81

The iodine removal efficiencies and recirculation flow rates are not applicable to the GDT rupture, which assumes only Xe-133 in the source term (no iodine).

Table 12.4 Offsite and Control Room Doses rem, TEDE EAB l

LPZ Control Room Max. 2-hour l

30 days 30 days I

2-hour release without 1.718E-1 1.320E-2 6.632E-2 CR isolation 2-hour release with CR 1.718E-1 1.320E-2 9.559E-2 isolation Puff release without CR 1.720E-1 1.332E-2 6.664E-2 isolation i Acceptance Criteria 1

0.5 1

5 I

I I

I Summary of Radiological Analyses, Revision 1, 11/03 Page 79 of 81

13.0 References l 1.

NUREG/CR-6331, Rev. 1 "Atmospheric Relative Concentrations in Building Wakes", J.

V. Ramsdell, C. A. Simonen, Pacific Northwest National Laboratory, 1997

2.

Draft Regulatory Guide DG-1 I 1, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, December 2001

3.

Ginna UFSAR, Revision 17, 10/02

4.

NUREG - 1465, "Accident Source Terms for Light-Water Nuclear Power Plants",

February 1995

5.

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

6.

HABIT Version 1.1, "Computer Codes for Evaluation of Control Room Habitability",

TACT5 and CONHAB Modules, NUREG/CR-6210, Supplement 1

7.

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

A. Powers, et al., Sandia National Laboratories, June 1993

8.

NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation", S.L. Humphreys, et. al., Sandia National Laboratories, April 1998. (See Section 10)

9.

NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations", T.J.

Bander, USNRC, 1982

10.

Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion",

Keith F. Eckerman, et al., Oak Ridge National Laboratory, 1988

11.

Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil", Keith F. Eckerman, et al., Oak Ridge National Laboratory, 1993

12.

A. G. Croff, "A User's Manual for the ORIGEN2 Computer Code", ORNL/TM-7175, Oak Ridge National Laboratory, July 1980

13.

RG&E Design Analysis DA-NS-2001-060, Atmospheric Dispersion Factors for the Control Room Intake, Rev 0 Summary of Radiological Analyses, Revision 1, 11/03 Page 80 of 81

14.

RG&E Design Analysis DA-NS-2003-004, Atmospheric Dispersion Factors for the Exclusion Boundary and Low Population Zone, Rev 0

15.

RG&E Design Analysis DA-NS-2001-063, Iodine and Noble Gas Activity in the Primary Coolant and Iodine Activity in the Secondary Coolant, Rev 1

16.

RG&E Design Analysis DA-NS-2001-064, Iodine Appearance Rates, Rev 1

17.

RG&E Design Analysis DA-NS-2001-087, Large Break LOCA Offsite And Control Room Doses, Rev 2

18.

RG&E Design Analysis DA-NS-2002-004, Fuel Handling Accident Offsite and Control Room Doses, Rev 1

19.

RG&E Design Analysis DA-NS-2002-007, Main Steam Line Break Offsite and Control Room Doses, Rev 2

20.

RG&E Design Analysis DA-NS-2001-084, Steam Generator Tube Rupture Offsite and Control Room Doses, Rev 1

21.

RG&E Design Analysis DA-NS-2002-054, Locked Rotor Offsite and Control Room Doses, Rev 0

22.

RG&E Design Analysis DA-NS-2002-050, Control Rod Ejection Accident Offsite and Control Room Doses, Rev 0

23.

RG&E Design Analysis DA-NS-2002-019, Tornado Missile Accident Offsite and Control Room Doses, Rev 1

24.

RG&E Design Analysis DA-NS-2000-057, Gas Decay Tank Rupture Offsite and Control Room Doses, Rev 2

25.

RG&E Design Analysis DA-NS-2002-037, HABIT Code Nuclear Data Library, Rev 0 Summary of Radiological Analyses, Revision 1, 11/03 Page 81 of 81 Particulate Removal by Sprays PARTICULATE REMOVAL BY SPRAYS NUREG/CR-5966 Input h:= 8911 Average spray nozzle elevation above the operating deck Diam:= 10511 Flow:= 1300 g min Inside diameter of the containment Total spray flow rate Calculation Diam2 A:= n-.

4 Area of containment operating deck Flow q:=-A Q~~~

cm sec) cm3 spray flux ft3 1

mm H := h.-

~q

=O0.02-ft t

/ /

10th Percentile, 90% confidence lamlo:= 5.575+ 0.94362*ln(Q) - 6.9821-10 3.Q 2-H - 7.327-IC 7.Q-H2 + 3.555 107 6Q2.H2

,o:=e lamlohr 50th Percentile, 50% confidence lamso:= 6.83707+ 1.00741n(Q) - 4.1731-10 3.Q2.H - 1.247&Q - 2.40451,0 5-H + 9.00610- 8.Q-HQ A, am50

_ I X5 0 := e lrohr7 90th Percentile, 90% Confidence lanio:= 7.10927-8.086810 4-Q2-H + 0.925491n(Q)

Iaxnj0

-I go9:= e hr Input Summary h = 89ft Diam= 105fI Flow = 1.3x I9 g min Output X9 0 = 17.551hr 1

-50 = 9.047hr 1 I0 = 3.48hr 1

.2/2 MSLB/Locked Rotor Two Hour Steam Release MSLB/Locked Rotor Two Hour Steam Release Main Steam Line Break (MSLB)

From Attachment 2, Table 7.1:

Steam Release 0 - 610 seconds (intact SG) = 37,780 lb From Reference 19 of Attachment 2, Section 7.1.6:

Steam Release Rate 610 seconds - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months = 7.29E+5 gm/min Calculation:

From 610 seconds to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months adds an additional 6590 seconds, or 109.83 minutes.

Therefore: (7.29E+5 gm/min) x (109.83 min) = 8.007E+7 gm Converted = 176,514 lb Total steam release (0 - 2 hr, intact SG) = 37,780 + 176,514

= 214,294 lb

2.

Locked Rotor From Attachment 2, Table 9.1:

Total Steam release 0 - 10 min = 54,620 lb 10 min - 30 min = 14,446 lb From Reference 21 of Attachment 2:

Steam release rate 30 min - 8 hr = 6.907E+5 gm/min Calculation:

From 30 min to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months adds an additional 90 minutes Therefore: (6.907E+5 gm/min) x (90 min) = 6.22E+7 gm Converted = 137,045 lb Total steam release (0 - 2hr, both SG) = 54,620 + 14,446 + 137,045

=206,111 lb 2,/2 Clarification of SGTR Steam Releases

A0teCJL-?*h e,0jt 6r CLARIFICATION OF SGTR STEAM RELEASES Steam releases (Ibm) and break flows (Ibm), for the affected and unaffected SGs, are shown in the table below. Westinghouse calculated these mass values and time intervals, with LOFTRAN.

Table I Steam Releases and Rupture Flow Time Periods, seconds Mass, 1000 Ibm 0-trip trip -break Break - 2 2hrs - RHR hours Affected SG to:

Condenser 45.5 Atmosphere 62.4 0

31.6 Unaffected SG to:

Condenser 45.2 Atmosphere 60.0 147.5 459.9 Rupture flow 2.9 107.4 trip:

break:

RHR:

reactor trip (49 sec.)

SG and RC pressures are equal, rupture flow is terminated (3492 sec)

RHR operating conditions are achieved, steaming to the environment is terminated (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

The steam releases, from the affected SG, represent the total mass released to the environment. The components of the total steam release are:

steam produced by boiling of the bulk SG water steam produced by flashing rupture flow

//c-

/tackoLw.Z a

The first part of the calculation is to convert the integrated steam and break flows (provided in the previous table) into flow rates.

Table 2 5U I R STELM J01EASE FIROM AFF-LITJ 515 AND RUPT IUXU-FLOW I

I I

1-I1 LOt-1IRAN I ime Steps Tor Steam Melease and breaK Hlow Start time, Start time, End of time Steam now, Steam flow Break flow, Break flow Break flow hr sec step, sec diuatin. min ibm rate. grnrrin Ibm rate, Ibmrmin rate, grrn~in O

-U-4 55M-+M Z53ETU-ZuwtE=+3 3.55M

.t61 E+%

0.014 49

-g2 57.33 2OE24 M3EE05 1T1U74Et T87E+03 8--YE+05 Q97C 3492 7i'~~~~~UO tJ1=+SUU1UOGE+'M U

.WtUtO.UU_+M

.t i 7200 Z88t+04 "6U.W I

.16UEt-+L0 3.93E+O U.WOEt+W _U.tW QE The flow rates are shown in both lb/min and grams/min. The analysis uses the gm/min values. Interval time is converted to hours (TACT5 time step unit is hours. TACT5 flow unit is gm/min).

The second part of the calculation is to split the rupture flow into flashed (steam) flow and non-flashed (liquid) flow and calculate the net boil-off steam release.

The total break flow is taken from Table 2. The integrated flashed flow is provided by Westinghouse.

The TACT5 input (start time in hours, flashing break flow rate, non-flashing break flow rate, and net steam flow rate) are calculated in Table 3.

Table 3 Star time, End time, duration, minHashing Flashing Flashing Flashing Non-Flashing Net steam sec sec break flow at break flow at break flow for break flow break flow flow rate, start time, end time, time step, rate. gm/min rate. gm/min gm/min ibm Ibm Ibm 49 0.8~

2 o2-.WE+W-2 4.24E=+2 2-7ME+05 1.3E3+06 2.boE+D 0.01 424-

-3 4E=+

4.B3E+=

4.1BE=+3

.79E+T 71T tE-33E5EU ON4 32 2

4E 3

4.bEO0 O.OE+

WO.OtO B4EDS 4EU5 2.vful 7200 I=

61.8U 4.53Et+03 4.bt+0 oEU

0.

Ot+0EW O.Out l Z.U00 1eUU I ;!At+ 04 I 60 4.Ddt+U03 4.001:+J 03U.WE+U I

0.Vcot+U 00 O.

OOE+003.UE

/9 tz~aCtA,K ewef 'e The following times are represented:

49 sec reactor trip 2004 break flashing stops 3492 break flow is stopped 7200 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: end of dose interval 2.88E4 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s: RHR operation begins, steam release is terminated The flashing break flow for the time interval t (integrated mass in lb), is calculated as follows:

FFT =FF1 2 - FFtI Where:

FFT

=

integrated flashing break flow for time interval T FF2=

integrated flashing break flow at time t2 FFtj

=

integrated flashing break flow at time tl The flashing break flow rate (steaming rate in grams/min) is calculated as follows:

FFT =FF TR DurT Where:

FFTR

=

flashing break flow rate for time interval T DurT

=

duration of time interval T This steam is released directly to the environment without mixing with the SG water or partition (partition =1).

The non-flashing break flow rate (gm/min) is calculated as follows:

NFTR = Wbrea.k -

FFTR Where:

NFTR

=

non-flashing break flow rate for time interval T Wbak =

total break flow rate for time interval T 2/^s

This water is assumed to homogeneously mix with the bulk SG water.

The net steam flow rate (gm/min) is calculated as follows:

W.e I=

Wg,,, -FFTR Where:

Wnet

=

net steam flow rate to environment due to SG boiling Wgross =

gross steam flow rate to environment (total due to boiling and flashing)

A partition of 100 is applied to the net steam elemental iodine activity release. A partition os I is applied to the methyl iodide activity release.

The steam released from and break flow to the affected SG are summarized as follows:

The last time interval is 7200 sec to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The rate of activity release from the SG to the environment is proportional to the activity concentration is the SG water, the steaming rate and the iodine partition coefficient.

Steam Release Start time, Break Flow sec Flash Total Steam, (not-flashed)

Steam, lb Boil-Off, lb lb lb 0

4.24FE+02 4.51 E+04

.5+4 28E0 49 4.16E+03 3.13E+U4 3.54E+U4 5.68E+04 2__4 O.DOE+UO 2.70E+4 2.70E+4 E

3492 O.OOE=+UU O.OOE+00 O.OOE+00 O.OEO 7200 O.OOZ+OO

.3.165+04 E3.16E-+04 O.0OE+OO 1.40E+05 1.06E+05

'/76

19t4.cA m r-Aid g'-

The following iodine partition coefficients are used:

Elemental iodine:

boil-off from bulk SG water flashing activity release Methyl iodide - all releases The TACT5 activity transport model, for the affected and unaffected SGs, is shown below.

Unaffected SG TACT5, SGTR Transport Model 67 PWR Natural Deposition Coefficients

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