NL-04-0979, Revised Response to Selected Questions from Second NRC Request for Additional Information Related to Request to Revise Technical Specifications - Containment Equipment Hatch

Revised Response to Selected Questions from Second NRC Request for Additional Information Related to Request to Revise Technical Specifications - Containment Equipment Hatch

ML041670409
Person / Time
Site:	Farley
Issue date:	06/10/2004
From:	Stinson L Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-0979
Download: ML041670409 (27)

Text

L M.Stinson (Mike) Southern Nuclear Vice President Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 SOUTHERN COMPANY A

June 10, 2004 Energy to Serve YourWorld' Docket Nos.: 50-348 50-364 NL-04-0979 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Units I and 2 Revised Response to Selected Questions from Second NRC Request for Additional Information Related to Reqtuest to Revise Technical Specifications - Containment Equipment Hatch Ladies and Gentlemen:

By letter dated August 29, 2003, Southern Nuclear Operating Company (SNC) submitted a request to amend the Farley Nuclear Plant (FNP) Unit I and Unit 2 Technical Specifications (TS), to allow the equipment hatch to be open during core alterations and/or during movement of irradiated fuel assemblies within containment. By letter dated November 11, 2003, SNC submitted a response to a Request for Additional Information related to that submittal. By letter dated May 5, 2004, SNC submitted a response to a second Request for Additional Information related to that submittal. In the May 5, 2004 response, SNC stated that a follow-up response would be submitted when results from Main Control Room tracer gas testing were obtained. Enclosure I provides the NRC questions and the final SNC responses for RAI questions impacted by tracer gas testing. provides portions of the proposed fuel handling calculation related to the evaluation of the open equipment hatch.

(Affirmation and signature provided on the following page).

A ,o-2

U. S. Nuclear Regulatory Commission NL,04-0979 Page 2 Mr. L. M. Stinson states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments. If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY L. M. Stinson orn to at subscribedbefore me this AO!Lvday of , 2004.

Notary Publicl My commission expires:

LMS/IWAS/sdl Enclosure I - Revised Response to NRC Request for Additional Information - Excerpt from Proposed Design Bases Fuel Handling Calculation cc: Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President 'w/o Enclosures Ms. C. D. Collins, General Manager - Farley w/o Enclosures Mr. D. E. Grissette, General Manager - Plant Farley RTYPE: CFA04.054; LC# 14052 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. S. E. Peters, NRR Project Manager - Farley Mr. C. A. Patterson, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer

ENCLOSURE 1 Joseph M. Farley Nuclear Plant Units 1 and 2 Revised Response to Selected Questions from Second Request for Additional Information Related to Request to Revise Technical Specifications - Containment Equipment Hatch SNC Revised Response to NRC Request for Additional Information Response to RAI Page 1 of 5 NRC Question 1 Please provide the design-bases parameters, assumptions, and methodologies (other than those provided in the August 29, 2003, submittal) that were changed in the radiological design-basis accident analyses as a result of the proposed TS change. Also, please provide justification for the changes. If there are many changes it would be helpful to compare and contrast them in a table.

SNC Response:

The following previously identified changes were made consistent with the guidance of Regulatory Guide (RG) 1.195:

a. Gap activity was changed as shown in RG 1.195, Table 2.

b. Pool decontamination was changed as shown in RG 1.195, Appendix B.

c. Credit for isolation of the containment was deleted and all activity assumed exhausted.

d. Control room x/Qs were recalculated using ARCON96, and are discussed further below.

Although not a change to the analysis, it should be noted that the 10 cfm unfiltered inleakage flow previously provided is for ingress/egress. In addition, the control room HVAC model was changed to reflect flows which bound the inleakage preliminary test results as shown in the attached revised Table 1.

NRC Question 2 Based upon a preliminary review of the fuel handling accident for the proposed TS change, the reviewer is unable to match the calculated doses. Please provide the calculation for the design bases fuel handling accident.

SNC Response:

Replacement pages for portions of the current fuel handling calculation related to evaluation of the open equipment hatch, previously provided with our letter of May 5, 2004, are provided in Enclosure

2. This calculation has been updated with the new X/Q values based on the new meteorological tower data set. The assumed control room inleakage values shown in the current calculation were revised to reflect values which bound inleakage preliminary test results.

NRC Question 4 Appendix B to Title 10 of the Code of FederalRegulations (10 CFR) Part 50 establishes quality assurance requirements for the design, construction, and operation of those structures systems or components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. Appendix B, Criterion III, "Design Control," requires that design control measures be provided for verifying or checking the adequacy of design. Appendix B, Criterion XVI, "Corrective Action," requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, defective material and equipment, and nonconformances are promptly identified and corrected. Generic Letter (GL) 2003-01, "Control Room Habitability," addresses current issues with respect to previously assumed values of unfiltered inleakage. Generally, these issues can only be resolved by inleakage testing. In light of Appendix B requirements and GL 2003-01, provide sufficient justification to explain why a value of zero for the control room's unfiltered inleakage is appropriate for this proposed TS change request. Provide details regarding your control room, design, maintenance and assessments to justify the use of this number.

Response to RAI Page 2 of 5 SNC Response:

FNP has performed integrated control room envelope testing using the standard test method described in American Society for Testing and Materials (ASTM) standard E74 1, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution." As-tested inleakage has been determined for the normal HVAC, isolation and pressurization modes of operation. Preliminary inleakage test results are approximately 84 cfm, 33 cfm and 26 cfm respectively (final, verified values will be provided with our formal response to Generic Letter 2003-001). Bounding inleakage values have been used to update the calculations for this submittal. In addition, the calculation was updated with the new X/Q values based on the new meteorological tower data set previously provided with our May 5, 2004 letter. The control room was designed for low leakage using the technology and methods prevalent at the time of FNP licensing. The primary source of inleakage is expected to be through ductwork operated at a negative pressure (on the suction side of fans), and this ductwork is typically of welded construction, minimizing the potential for inleakage. Post-accident access to the control room would typically be through the non-rad area corridor from Unit I or through the Technical Support Center (TSC) from Unit 2. Inleakage from the TSC area would be filtered, and the non-rad area corridor of Unit 1 has two doors in series minimizing inflow which would be diluted by the air volume of the auxiliary building. In addition to the 10 cfm ingress/egress flow assumption; 600 cfm unpressurized, unfiltered inleakage is assumed during the first ten minutes until the pressurization flow is manually established. After pressurization with 450 cfm filtered outside air, 10 cfm ingress/egress flow plus 450 cfm unfiltered inleakage is assumed. Review of the calculated releases indicates most activity is released from the open containment during this first ten minute period when the unpressurized inleakage is the dominant source. The results of the current dose calculation are approximately 35% of the 50 REM thyroid limit, indicating substantially more unfiltered inleakage can be accommodated. Based on these results, the assumed unfiltered control room inleakage for the fuel handling accident could be increased by approximately a factor of 3 and doses would remain within the acceptance criteria in Regulatory Guide 1.195. Such leakage would be well in excess of results from recent industry tracer gas tests for pressurized control rooms and the LOCA inleakage acceptance criteria used in the control room inleakage testing performed at FNP in response to Generic Letter 2003-01.

NRC Question 6 Provide a detailed account of the timing and flow rates, and filtration of the control room Heating Ventilation and Air Conditioning as it responds to the accident. Please justify the assumption that one of the two emergency control room filtration trains are operating within 10 minutes of the accident. Is this action automatic or manual?

SNC Response:

The control room is normally maintained at a slightly positive pressure by providing about 1350 cfm from the computer room Heating, Ventilation and Air Conditioning (HVAC) system, about 35% of which is outside air. However, for conservatism 3000 cfm of 100% outside air is assumed. This normal air supply is monitored for high radiation, and the setpoint is expected to be reached and generate an isolation signal within 20 seconds. For conservatism, 45 seconds isolation time is assumed, after which the control room is unpressurized with inleakage assumed to be 600 cfm plus 10 cfm ingress/egress flow. This continues for ten minutes, until the operator manually initiates Response to RAI Page 3 of 5 emergency filtered pressurization and recirculation filtration. Ten minutes is the currently assumed operator action time, and is a conservative assumption given the communication between the control room operators and the refueling floor and also the outside air intake radiation monitor alarm in the control room. Outside air flow for pressurization is assumed to be 450 cfm with 450 cfm unfiltered inleakage and 10 cfm ingress/egress flow. Recirculated filtered air flow is 2700 cfm for the duration of the analysis, based on technical specification limits.

Response to RAI Page 4 of 5 Table 1 (sheet I of 2)

PARAMETERS USED IN FUEL HANDLING ACCIDENT ANALYSIS (Accident in Containment with Equipment Hatch Open)

Core thermal power 2831 MWt Time between plant shutdown and accident 100 h Minimum water depth between tops of 23 ft Damaged fuel rods and water surface Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release from assembly Gap activity in ruptured rods per RG 1.195, Table 2 Radial peaking factor 1.7 Decontamination factor in water Elemental iodine (99.75%) 400 Organic iodine (0.25%) 1 Noble gases 1 Amount of mixing in building 6.6x10 5 ft3 ()

Exhaust flow rate 53,500 cfm(2 )

Isolation time N/A Iodine filtration system Containment purge system (not credited)

Filter efficiency (all species) N/A Atmospheric dilution factors Accident Offsite, FSAR Tablel 5B-2 Control Room, see sheet 2

Enclosure I Response to RAI Page 5 of 5 Table 1 (sheet 2 of 2)

CONTROL ROOM PARAMETERS USED IN FUEL HANDLING ACCIDENT ANALYSIS Previous value Current value(5)

Normrial HVAC unfiltered intake (ft 3 /min) 1350 3000 Unpri -ssurized unfiltered infiltration (ft 3 /min) 187.5 600 Filtered pressurization rate (ft 3 /min) 375 450 Press urized unfiltered infiltration (fl 3 /min) 0 450 Filter ed recirculation rate (ft 3 /min) 2700 No change Unfil tered ingress/egress rate (ft 3 /min) 10 No change Filter efficiencies (all forms of iodine) (%)

Pressurization air 98.5(3 No change Recirculation air 94.5(3) No change Volume (ft3 ) 114,000 No change Operator breathing rate (m3 /s) 3.47 x 104 No change Percent of time operator is in control room No changes following LOCA 0-1 day 100 1-4 days 60 4-30 days 40 Atmospheric dilution estimates (s/m 3)(4 0-2 h 8.42 x 104 8.39x 10 4 2-8 h 6.43 x 104 5.10 x 104 Notes (1) 90% of the volume above the operating deck and below the containment fan cooler registers.

(2) This flow discharges 9.7 times the mixing volume, resulting in complete release, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

(3) Filter efficiencies have been reduced by 0.5% for all forms of iodine to account for bypass leakage.

(4) Most limiting equipment hatch-control room intake values recalculated with ARCON96.

(5) Current HVAC flow values used to bound control room inleakage preliminary test results.

ENCLOSURE 2 Joseph M. Farley Nuclear Plant Units 1 and 2 Revised Response to Selected Questions from Second Request for Additional Information Related to Request to Revise Technical Specifications - Containment Equipment Hatch Excerpt from Proposed Design Bases Fuel Handling Calculation (To Be Implemented After NRC Approval of Technical Specification Change)

Southern Nuclear Design Calculations

[Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 2 of 252

==

Conclusions:==

The fuel handling accident in the containment with the equipment hatch open, analyzed in accordance with RG 1.195 results in offsite doses which meet the acceptance criteria:

Thyroid dose Skin dose Whole body (Rem) (Rem) dose (Rem)

Site Boundary 68.5 0.5 0.2 Low Population Zone 25.3 0.2 0.1 Control Room Laot 0.5 <0.1 The realistic estimate of the thyroid dose to the individual(s) designated to close the equipment hatch, with no credit for a respirator or other protective gear is 23.4 Rem, which meets the RG 1.195 limit for operators.

The maximum exposure time to remain within the 50 Rem limit would be about 65 minutes.

From Attachment 5, the FHA in the auxiliar building meets the acceptance criteria assuming credit for the operation of the penetration area filtration system with 0.5% bypass flow.

Thyroid dose Skin dose Whole body (Rem) (Rem) dose (Rem)

Site Boundary 21.6 0.8 0.4 Low Population Zone 7.9 0.3 0.1 Control Room 10.3 From Attachment 6, the FHA in the auxiliary building without PRF filters meets the "minimal increase" acceptance criteria IF the most recently discharged fuel in the spent fuel pool has decayed at least 676 hours0.00782 days <br />0.188 hours <br />0.00112 weeks <br />2.57218e-4 months <br /> since discharge from the reactor.

Site Boundary Low Population Zone Control Room Thyroid Dose (Rem) 25.7 9.5 12.3

[Project Farley Nuclear Plant Subject/ritle Southern Nuclear Design Calculations Calculation Number SM-96-1064-001 Sheet Fuel Handling Accident Doses 3 of 252 From Attachment 7, the FHA in the containment with the doors closed and no credit for the purge filter meets the acceptance criteria:

Site Boundary Low Population Zone Control Room Thyroid Dose (Rem) 12.1 4.5 28.6 Whole Body Dose (Rem) 0.4 0.1 0.8 Maior Equations: The calculations are prepared using the TACT 5 computer code, installed on a NEC P90 computer. Proper operation of the program was verified by running the test problems and comparing the output to that provided in reference 10. A listing of the TACT 5 directory is included in the Attachment 1.

The equations for the numerical solutions for activity transport, and the equations for the resultant doses are described and discussed in reference 10. Equations used to derive input values for the computer analyses are explained in the body of the calculation.

Assumptions: Offsite dose analyses are prepared using dose conversion factors from ICRP 30 as provided in the users manual for TACT 5 (reference 10). This is consistent with the NRC SERs for similar recent analyses (references 13 and 14).

Assumptions for the FHA in the auxiliary building or containment with the equipment hatch closed are consistent with the guidelines of Regulatory Guide 1.25, except as noted above, and are shown in Table 1. It is assumed that the air space above the spent fuel pool is maintained at a negative pressure relative to adjacent areas so that radioactive releases are processed through the penetration area filtration system--this implies that the doors and hatches into the area are closed or allow flow only into the spent fuel handling area, thereby preventing bypass of the filtration system.

Assumptions for the FHA in the containment with the equipment hatch open are consistent with the applicable portions of RG 1.195. The major differences are in the activity transport assumptions, which are dependent on the functioning of the various HVAC systems in the containment and the volume served.

it addition,control room MVA C is modeled to bound the Generic Letter 2003-001 test results. The normal intake rate is increasedto 3000 cfm wliclt includes the as-tested unfiltered inleakage. The isolation mode unnflftered inleakage is assumed to be 600 cfni plus 10 cfdn for ingress/eress. and the pressurizationflow is assumed to be 450 cfun with 450 cfm naflftered inleakage plus 10 cfht for hnrsseress Finally' the margin included in source terms is increasedto bound the U2C 7 core reload(lesi'n as describedin Action Item (Al) 200420003.

EProject Farley Nuclear Plant SubjectfTitle Southern Nuclear Design Calculations Calculation Number SM-96-1064-001 Sheet Fuel Handling Accident Doses 6 of 252

25. USNRC Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," 5/03.

26. Calculation BM-03-0018-001,4aerion "Control Room and Technical Support Center Air Intake X/Q Estimates," 4/I0.

27. Technical Specifications, Section 5.5.11

28. "Fuel Handling Accident (FHA) Drop Height Reference Information," Westinghouse Letter ALA-Ol-057, June 12, 2001.

E Project Farley Nuclear Plant SubjectfTitle Southern Nuclear Design Calculations Calculation Number SM-96-1064-001 Shed Fuel Handling Accident Doses 7 of 252 Bodv of Calculation:

A. Source Terms Source terms for the FHA are developed from Westinghouse input of core activities for the uprated power of 2775 MWt (reference 8), plus 2% for calorimetric errors (2831 MWt, reference 9). Parametric analyses were performed by Westinghouse to determine how these source terms might vary with different fuel designs (enrichment, blankets, burnup) to define a "bounding" factor to be applied to the base core inventory.

Because the resultant thyroid, whole body and skin doses are determined by iodines and noble gases, the parametric analyses for these isotopes were examined and the base core inventory is multiplied by 1.02 to bound the FHA source terms.

For the FHA in the containment with the equipment hatch open, the core inventori' margin is increasedto 1.03 (except is1.15 and 05)uLdthe gas gap inventory is 5 % of the core inventory except

-eaisnd,.

Kr85 is 10% and 1131 is 8 % consistent with reference 25. The TACT 5 code uses Ci/MWt as input, so the core inventory is then divided by 2775 and input without decay to file FNPGAP30:

Inventory x 1.03 x 0.05 (ex Kr85 = 0.1 and I131 = 0.08)

• 2775 = Ci/MWt. I The remaining data in FNPGPL25 is from file MLWRICRP.30 from reference lOa, Appendix E 2 3 3 0 0 1 1 WHOLEBDY SKIN THYROID I 135 HALOGENS NOBLES 2.864000E-05 2.784000E+03 I ELEM. ORG. PART. 2.490000E-01 7.860000E-02 3.100000E 04 I 131 0.0 0.0 0.0 9.963996E-07 2.227000E+03 0.0 0.0 0.0 I 5.590000E-02 3.070000E-02 1.100000E 06 5 1 0 0 0 0 0 0.0 0.0 0.0 KR 83M 0.0 0.0 0.0 1.035000E-04 1.00000E+02 I 1 1 0 0 0 0 0 1.270000E-05 0.OOOOOOE+00 0.0 I 132 0.0 0.0 0.0 8.269001E-05 2.041000E+03 0.0 0.0 0.0 3.550000E-01 1.100000E-01 6.300000E 03 6 2 0 0 0 0 0 I 0.0 0.0 0.0 KR 85M 0.0 0.0 0.0 4.385000E-05 3.897000E+02 I 2 1 0 0 0 0 0 2.310000E-02 4.970000E-02 0.0 I 133 0.0 0.0 0.0 9.219000E-06 2.969000E+03 0.0 0.0 0.0 9.110000E-02 8.900000E-02 1.800000E 05 7 2 0 0 0 0 0 I 0.0 0.0 0.0 KR 85 0.0 0.0 0.0 2.042000E-09 2.984000E+01 I 3 1 0 0 0 0 0 3.310000E-04 4.840000E-02 0.0 I 134 0.0 0.0 0.0 2.2280QOE-04 3.155000E+03 0.0 0.0 0.0 I 4.1100OOE-01 1.420000E-01 1.100000E 03 8 2 0 0 0 0 0 1 0.0 0.0 0.0 0.0 0.0 0.0 4 1 0 0 0 0 0

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subjectfritle Sheet Fuel Handling Accident Doses 8 of 252 XE 133 KR 87 1.522000E-06 2.838000E+03 1.519000E-04 7.423000E+02 4.960000E-03 9.670000E-03 0.0 1.330000E-01 3.360000E-01 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 13 2 0 0 0 0 0 9 2 0 0 0 0 0 XE 135M KR 88 7.400000E-04 5.56_000E+02 6.875000E-05 1.058000E+03 6.370000E-02 2.140000E-02 0.0 3.380000E-01 7.760000E-02 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 14 2 0 0 0 0 0 10 2 0 0 0 0 0 XE 135 XE 131M 2.091999E-05 6.495000E+02 6.680000E-07 1.559000E+01 3.590000E-02 6.320000E-02 0.0 1.250000E-03 1.330000E-02 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 15 2 0 0 0 0 0 11 2 0 0 0 0 0 XE 137 XE 133M 2.961000E-03 2.598000E+03 3.490000E-06 8.90_OOOE+01 2.830000E-02 4.590000E-01 0.0 4.290000E-03 2.960000E-02 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 16 2 0 0 0 0 0 12 2 0 0 0 0 0 XE 138 6.796001E-04 2.413000E+03 1.870000E-01 1.470000E-01 0.0 0.0 0.0 0.0 0.0 0.0 0.0 17 2 0 0 0 0 01

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subjectfritle Sheet Fuel Handling Accident Doses 12 of 252 C. Containment Building Model: Regulatory Guide 1.195 FHA in Containment (with Equipment Hatch Open)

For an FHA in the containment, the source terms are modeled as above in section A. The building space used for the dilution of the activity released is evaluated for the space served by the containment purge system.

During refueling, the containment may be purged using the main purge, drawing air from elevations 130 and 155 of the containment (references 7d, e, f). [NOTE that the pool sweep is typically not used, and has been disconnected from the purge exhaust system. If it were to be used, it would exhaust to elevation 139 below the operating deck.] Assuming mixing above the pool up to the elevation of the containment cooling fan header (but no mixing outside this envelope which might be induced by operation of the fans), and 90% free space, this represents a volume of approximately:

7ix652 x 55x0.9 6.6x105 ft 3 .

An FHA in the refueling canal will release the gap activity from one assembly (reference 11, 12) with the highest power, i.e. with a peaking factor of 1.7 (reference 8, which exceeds the reference 1 value), to the pool water which will scrub the iodines as they evolve into the building air. Release factors from reference 25 are 8% 131,10% Kr85, and 5% for other iodines and noble gases. The original partition between elemental and organic species of iodine and the cleanup by the pool water as described in reference 25 are combined to result in 0.5% of the iodine instantaneously released to the building, partitioned as 50% organic and 50%

elemental.

Releases are driven from the containment by the purge system flow. Although the initial release concentration in the containment:

895.9 Ci Kr85 x I E6 pCi/Ci -6.6 E5 ft3 ÷28317 cc/ft3 = 5 x 10-2 IpCi/cc is well in excess of the containment purge isolation radiation monitor setpoint of 1 x 10-4 pCi/cc, no credit for stopping the purge flow or purge filter is modeled. Then the main purge flow plus 10% exhausts 48,500 x 1.1 cfm x 120 min / 6.6 x 105 fl3 = 9.7 times the containment volume, which exceeds the requirements of reference 25. Releases are modeled as ground level, with X/Q values taken from reference 26. The annotated input file (FHAHTCH6.in) and offsite dose results are shown below, where control room parameters are taken from section D below.

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 l Subject/Title Shed Fuel Handling Accident Doses 13 of 252 FHAHTCH=6.in

'c:fnva 195

'c:LTAPE ','c:MTAPE ','c:NTAPE I RG 1.195 source terms from section A above.

I Farley FHA in the Containment with Equip Hatch Open (RG 1.195) 1 0, 0, 1, 1, 0 2, 2

'Contnmnt','CtrlRoom' 25 1.7675E+1, -1.OOE+02 2775 MWt/157 assys, 100 hr decay 8.500E-03, 1.700E-00 1.7/400 iodine and 1.7 noble gas 0.000E+00, O.OOOE+00 release fractions 5.OOOE-01, 5.OOOE-01, 0.OOOE+00 50% elemental, 50% organic iodine 1.000E+00, 0.OOOE+00, O.OOOE+00 100% elemental noble gases 6.600E+05, 1.140E+05 Containment refueling area and

'TIME INTERVAL ',0,0,0,0,2, 0. OOOE+00, 2.7778E-06 control room volumes

'INITIAL FRACTION',0,0,0,0,2, 1. OOOE+00, 0. OOOE+00

'FILTER EFF 1,1,0,1,3, 0. OOOE+00, 0. OOOE+00, 0.OOOE+00

'FILTER EFF ',1,2, 0, 1, 3, 0. OOOE+00, 0. OOOE+00, 0.OOOE+00

'FILTER EFF ',1,3, 0, 1, 3, 0. OOOE+00, 0. OOOE+00, 0.OOOE+00

'TRANSFER CFM '0,0,0, 1,3, 5.350E+04, 0. OOOE+00, 6. 434-E+01 HVAC flow rates I

'TRANSFER CFM ,0, 0, 0,2, 3, 1.360E+03, 0. OOOE+00, 1.OOOE+00 Normal CR exhaust flow

'DOSE PARAMS 10, 0,0,0,5, 7.600E-04, 3.470E-04, 2.800E-04, 3.470E-04, O.OOOE+00

'TIME INTERVAL ,0,0,0,0,2, 2.778E-06, 2.778E-04

'TIME INTERVAL '0,0,0,0,2, 2.778E-04, 8.333E-03

'TIME INTERVAL ,0,0,0,0,2, 8.333E-03, 1.250E-02

'TIME INTERVAL '0, 0,0,0,2, 1.250E-02, 2.500E-02 45 sec, CR auto-isolation

'FILTER EFF ',1,1,0, 1,3, 0.OOOE+00, 0. OOOE+00, 0.OOOE+00

'FILTER EFF ',1,2,0, 1,3, 0.OOOE+00, 0.OOOE+00, 0.OOOE+00

'FILTER EFF ',1, 3, 0, 1, 3, 0. OOOE+00, 0.OOOE+00, 0.OOOE+00

'TRANSFER CFM 0,0,0,1,3, 5.350E+04, 0.OOOE+00, 18. 08-E+00 Unpressurized CR intake

'TRANSFER CFM ',0,0,0,2,3, 6.100-E+02, 0.OOOE+00,, 1.OOOE+00 3.470E-04, 2.800E-04, 3.470E-04, O.OOOE+00 flow and exhaust I

'DOSE PARAMS ',0,0,0,0,5, 7.600E-04,

'TIME INTERVAL ',0,0,0,0,2, 2.500E-02, 4.167E-02

'TIME INTERVAL ',0,0,0,0,2, 4.167E-02, 8.333E-02

'TIME INTERVAL ',0,0,0,0,2, 8.333E-02, 1.667E-01

'TIME INTERVAL ',0,0,0,0,2, 1. 667E-01, 3.333E-01 10 min, manual pressurization

'FILTER EFF ',1,1,0,1,3, 0. OOOE+00, 0. OOOE+00, 4.871-E+OlEquivalent pressurization

'FILTER EFF ',1,2,0,1,3, 0. OOOE+00, 0.OOOE+00, 4.871-E+01 filter efficiency

'FILTER EFF ',1,3,0,1,3, 0. OOOE+00, 0.OOOE+00, O.OOOE+00

'FILTER EFF ',1,1,0,2,3, 0. OOOE+00, 0.OOOE+00, 9.450E+01 Recirculation filter

'FILTER EFF ',1,2,0,2,3, 0. OOOE+00, 0.OOOE+00, 9.450E+01 efficiency

'FILTER EFF ',1,3,0,2,3, 0. OOOE+00, 0.OOOE+00, O. OOOE+00

'TRANSFER CFM 1,0,0,0,1,3, 5.350E+04, 0.OOOE+00, 19.52-E+00 Pressurized CR intake

'TRANSFER CFM 1,0,0,0,2,3, 9.100-E+02, 0.OOOE+00 2.700E+03 flow and exhaust

'DOSE PARAMS ',0,0,0,0,5, 7. 600E-04, 3.470E-04, 2.800E-04, 3.470E-04, O.OOOE+00

'TIME INTERVAL ',0,0,0,0,2, 3.333E-01, 6.666E-01

'TIME INTERVAL ,0,0,0,0,2, 6. 666E-01, 7.500E-01

'TIME INTERVAL ',0,0,0,0,2, 7.500E-01, 1.OOOE+00

'TIME INTERVAL ',0,0,0,0,2, 1. OOOE+00, 1.250E+00

'TIME INTERVAL ',0,0,0,0,2, 1.250E+00, 1.500E+00

'TIME INTERVAL ',0,0,0,0,2, 1. 500E+00, 1.750E+00

'TIME INTERVAL ',0,0,0,0,2, 1. 750E+00, 2.OOOE+00

'TIME INTERVAL ',0,0,0,0,2, 2. OOOE+00, 2.250E+00

Southern Nuclear Design Calculations

[Project Calculation Number

[Farley Nuclear Plant Subjectfritle Fuel Handling Accident Doses SM-96-1064-001 Sheet 14 of 252

'TRANSFER CFM ',0,0,0,1,3, 5.350E+04, 0.OOOE+00, 14.96-E+00 Change CR X/Q and flow I

'DOSE PARAMS ',0,0,0, 0,5, 2.900E-04, 3.470E-04, 1.100E-04, 3.470E-04, 0.OOOE+00

'TIME INTERVAL ',0, 0, 0, 0, 2, 2.250E+00, 2.500E+00

'TIME INTERVAL ',0, 0, 0, 0, 2, 2.500E+00, 3.OOOE+00

'TIME INTERVAL ,0, 0, 0, 0, 2, 3.OOOE+00, 4.OOOE+00

'TIME INTERVAL ',0, 0, 0, 0, 2, 4.000E+00, 5.000E+00

'TIME INTERVAL ,0, 0,0,0,2, 5.OOOE+00, 6.OOOE+00

'TIME INTERVAL ,0, 0,0,0,2, 6.000E+00, 7.OOOE+00

'TIME INTERVAL ,0, 0,0,0,2, 7.OOOE+00, 8.OOOE+00

'TIME INTERVAL 0, 0, 0, 0,2, 8.OOOE+00, 24.OOE+00

'TRANSFER CFM ,0, 0,0, 1,3, 5.350E+04, 0.OOOE+00, 7.379-E+00 Change CR X/Q and flow I

'DOSE PARAMS ',0, 0,0,0,5, 3.300E-05, 3.470E-04, 1.OOOE-05, 3.470E-04, 0.OOOE+00

'END 110,0,0,0,0, 0.OOOE+00, 0.OOOE+00, 0.OOOE+00

SUMMARY

OF OFF-SITE DOSES Farley FHA in the Containment with Equip Hatch Open (RG 1.195)

CALCULATION FOR WHOLEBDY DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF START EXCLUSION RADIUS LOW POPULATION ZONE TIME EACH ACCUM. EACH ACCUM.

(HRS) STEP STEP 0.OOOE+00 2.705E-06 2.705E-06 9.967E-07 9.967E-07 2.778E-06 2.677E-04 2.704E-04 9.861E-05 9.961E-05 2.778E-04 7.683E-03 7.953E-03 2.830E-03 2.930E-03 8.333E-03 3.857E-03 1.181E-02 1.421E-03 4.351E-03 1.250E-02 1. 1E-02 2.292E-02 4.094E-03 8.445E-03 2.500E-02 1.381E-02 3.673E-02 5.086E-03 1.353E-02 4.167E-02 2.997E-02 6.670E-02 1.104E-02 2.457E-02 8.333E-02 4.447E-02 1.112E-01 1.638E-02 4.096E-02 1.667E-01 4.934E-02 1.605E-01 1.818E-02 5.914E-02 3.333E-01 3.166E-02 1.922E-01 1.166E-02 7.080E-02

6. 666E-01 2.598E-03 1.948E-01 9.571E-04 7.176E-02 7.500E-01 3.653E-03 1.984E-01 1.346E-03 7.310E-02 1.OOOE+00 1.084E-03 1.995E-01 3.995E-04 7.350E-02 1.250E+00 3.239E-04 1.998E-01 1.193E-04 7.362E-02 1.500E+00 9.857E-05 1.999E-01 3.632E-05 7.366E-02 1.750E+00 3.158E-05 2.000E-01 1.164E-05 7.367E-02

2. OOOE+00 4.382E-06 2.OOOE-01 1.662E-06 7.367E-02 2.250E+00 2.019E-06 2.000E-01 7.657E-07 7.367E-02 2.500E+00 2.168E-06 2.000E-01 8.225E-07 7.367E1-02 3.OOOE+00 2.567E-06 2.000E-01 9.739E-07 7.368E-02

4. OOOE+00 1.560E-06 2.000E-01 5.917E-07 7.368E-02 5.OOOE+00 9.606E-07 2.0001E-01 3.643E-07 7.368E-02 6.OOOE+00 5.9161-07 2.0001-01 2.244E-07 7.368E-02 7.000E+00 3.644E1-07 2.000E-01 1.382E-07 7.368E-02 8.OOOE+00 6.649E-08 2.000E-01 2.015E-08 7.368E-02 TOTAL 2.000E-01 TOTAL 7.368E-02

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 15 of 252 CALCULATION F'OR SKIN DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF START EXCLUSION RADIUS LOW POPULATION ZONE TIME EACH ACCUM. EACH ACCUM.

(HRS) STEP STEP 0.000E+00 5.717E-06 5.717E-06 2.106E-06 2.106E-06 2.778E-06 5.657E-04 5.714E-04 2.084E-04 2.105E-04 2.778E-04 1.624E-02 1.681E-02 5.982E-03 6.192E-03 8.333E-03 8.152E-03 2.496E-02 3.004E-03 9.196E-03 1.250E-02 2.349E-02 4.845E-02 8.653E-03 1.785E-02 2.500E-02 2.918E-02 7.762E-02 1.075E-02 2.860E-Q02 4.167E-02 6.335E-02 1.410E-01 2.334E-02 5.194E-02 8.333E-02 9.398E-02 2.350E-01 3.463E-02 8.656E-02 1.667E-01 1.043E-01 3.392E-01 3.842E-02 1.250E-01 3.333E-01 6.692E-02 4.062E-01 2.466E-02 1.496E-01 6.666E-01 5.492E-03 4.117E-01 2.023E-03 1.517E-01 7.500E-01 7.722E-03 4.194E-01 2.845E-03 1.545E-01 1.O0OE+00 2.293E-03 4.217E-01 8.446E-04 1.554E-01 1.250E+00 6.852E-04 4.224E-01 2.524E-04 1.556E-O1 1.500E+00 2.087E-04 4.226E-01 7.690E-05 1.557E-01 1.750E+00 6.707E-05 4.226E-01 2.471E-05 1.557E-01 2.OOOE+00 9.367E-06 4.226E-01 3.553E-06 1.557E-01 2.250E+00 4.361E-06 4.226E-01 1.654E-06 1.557E-01 2.500E+00 4.742E-06 4.227E-01 1.799E-06 1.557E-01 3.OOOE+00 5.654E-06 4.227E-01 2.145E-06 1.557E-01 4.OOOE+00 3.439E-06 4.22W-OX 1.304E-06 1.55W-OX 5.OOOE+00 2.119E-06 4.227E-01 8.036E-07 1.557E-01 6.OOOE+00 1.305E-06 4.227E-01 4.951E-07 1.557E-01 7.OOOE+00 8.042E-07 4.227E-01 3.051E-07 1.557E-01 8.OOOE+00 1.468E-07 4.227E-01 4.450E-08 1.557E-01 TOTAL 4.227E-01 TOTAL 1.557E-01

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM96-1064-001 SubjectVTitle Shed Fuel Handling Accident Doses 16 of 252 CALCULATION FOR THYROID DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF START EXCLUSION RADIUS LOW POPULATION ZONE TIME EACH ACCUM. EACH ACCUM.

(HRS) STEP STEP 0.00OE+00 9.264E-04 9.264E-04 3.413E-04 3.413E-04 2.778E-06 9.166E-02 9.258E-02 3.377E-02 3.411E-02 2.778E-04 2.631E+00 2.723E+00 9.692E-01 1.003E+OO 8.333E-03 1.321E+00 4.044E+00 4.867E-01 1.490E+OO 1.250E-02 3.805E+00 7.850E+00 1.402EfO0 2.892E+OO 2.500E-02 4.728E+00 1.258E+01 1.742E+00 4.634E+OO 4.167E-02 1.027E+01 2.284E+01 3.782E+00 8.416E+OO 8.333E-02 1.523E+01 3.807E+01 5.611E+00 1.403E+Ol 1.667E-01 1.690E+01 5.498E+01 6.228E+00 2.025E+O1 3.333E-01 1.085E+01 6.583E+01 3.997E+00 2.425E+O1 6.666E-01 8.903E-01 6.672E+01 3.280E-01 2.458E+O1 7.500E-01 1.251E+00 6.797E+01 4.610E-01 2.504E+O1 1.000E+00 3.707E-01 6.834E+01 1.366E-01 2.518E+Ol 1.250E+00 2.099E-01 6.845E+01 4.048E-02 2.522E+O1 1.500E+00 3.263E-02 6.848E+01 1.202E-02 2.523E+Ol 1.750E+00 9.720E-03 6.849E+01 3.581E-03 2.523E+O1

2. OOOE+00 1.113E-03 6.849E+01 4.222E-04 2.523E+Ol 2.250E+00 3.390E-04 6.849E+01 1.286E-04 2.523E+O1 2.500E+00 1.417E-04 6.849E+01 5.374E-05 2.523E+O1 3.OOOE+00 2.212E-05 6.849E+01 8.391E-06 2.523E+O1 4.OOOE+00 1.761E-06 6.849E+01 6.681E-07 2.523E+Ol 5.000E+00 2.699E-07 6.849E+01 1.024E-07 2.523E+01 6.OOOE+00 4.338E-08 6.849E+01 1.645E-08 2.523E+01 7.000E+00 6.988E-09 6.849E+01 2.650E-09 2.523E+O1 8.OOOE+00 1.527E-10 6.849E+01 4.627E-11 2.523E+Ol TOTAL 6.849E+O1 TOTAL 2.523E+O1 D. Control Room Model The control room normally operates with an air supply of 1350 cefm (ref. 17) coming from the computer room A/C. Approximately 35% of this is outside air; however for conservatism, a total air flowsof i is assumed to be outside air. TACT 5 will not allow transfer from the environment back into a node and allows only one transfer path for both filtered intake and unfiltered inleakage, so an equivalent direct transfer from the containment via the environment to the control room is modeled as:

Release Rate x X/Q x Unit Conversion Factors x Intake Rate.

Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 SubjectTitle Shed Fuel Handling Accident Doses 17 of 252 The effective filter efficiency is modeled as:

1 - [(1- efficiency) x Filtered Flow + Unfiltered Flow] / [Filtered Flow +Unfiltered Flow].

The control room intake radiation monitor will isolate the control room HVAC, but manual action is required to start the pressurization and recirculation mode of operation. The pressurization and recirculation flows are assumed to be 450 cfm andtheTechnical Specification (ref. 27) value of 3000 +/- 10%. The control room filter efficiency modeled includes a reduction of 0.5% allowed for bypass testing yielding:

CR volume = 114,000 ft3 (ref. 16)

Pressurization flow = 15 cfin, efficiency = 98.5%

Recirculation flow = 2700 cfm, efficiency==4.5%

Normal flow = 3M00 cfm, efficiency= 0%

Pessuriz~edunfilteredinleakage = 1 cfm, efficiency= 0% (assumed)

For the period between automatic isolation and manual pressurization, the control room unfiltered inleakage is assumed to be 60ASm unfiltered inleakage plus 10 cfm for ingress/egress= 600 + 10 =61 cfm. Then I for a containment purge flow rate of 53,500 cfm, the direct transfer rates and filter efficiencies are:

Release Z'Q Unit Conversion Intake Direct Effective Filter Efficiency flow Factors flow transfer (cfim) (s/m 3 ) (min/s) (m3 /ft') (cfni) (cfm) 53,500 8.42E-4 1/60 1/35.3 3000 64.34 0 610 DIN0 0 1%52 1-[ (1 -0.985) +A450+l 0]/(WJQ) =481Z%

6.5E-4_ 14.96 Ir 3.2tE-4 Ir I r 7~3Z2 (Checker's note: The X/Q values used aresightli' different from reference 26. but give conservative results.)

The switch from normal HVAC intake (30 cfin) to unpressurized unfiltered inleakage (A cfm) occurs I when the intake radiation monitor responds to the released activity. The 1E-4 pCi/ml monitor setpoint (ref.

15) is exceeded within about 1 second of the activity reaching the monitor:

Released activity = .L2E6 piCi X/Q = 8.42E-4 (s/m3 ) l (1 42OE6 pCi / 1 sec) x (8.49E-4 s/m3 ) / (1E6 ml/m3 ) 1I E-1aCi/ml Based on a review of reference 20, this should result in an isolation signal within about 15-20 seconds, so use 45 seconds to conservatively bound the AOV (HV3 622-3 629) stroke time.

EProject Farley Nuclear Plant SubjectlTitle Southern Nuclear Design Calculations Calculation Number SM-96-1064-001 Sheet Fuel Handling Accident Doses 18 of 252 The resultant control room activities are entered into an EXCEL spreadsheet to calculate the control room thyroid, skin and whole body doses. Thyroid doses are calculated for each time step from (Average Ci-hr) x (3600 s/hr) x (35.3 ft3 /m3 ) x (3.47E4 m3 inhaled/s) x DCF (REM/Ci)/l 14,000 ft3 and whole body and skin doses are calculated from (1/Geometry factor) x (Average Ci-hr) x (35.3 ft 3 /m3 ) x DCF (REM-m 3/Ci-hr)/1 14,000 ft3 where the geometry factor is applied only to the whole body dose and is 1173/ 1140000.338 = 22.91 (ref 25).

FHAHTCHB.out Control Room Inventory I (Curies)

Time (hr)

ISOTOPE 2.78E-06 8.33E-03 1.25E-02 2.50E-02 4.17E-02 8.33E-02 1 131 1.90E-06 5.57E-03 8. 25E-03 9. 8OE-03 1. 17E-02 1.58E-02 1131 1.90E-06 5.57E-03 8.25E-03 9. 80E-03 1. 17E-02 1.58E-02 1 132 2.94E-19 8. 59E-16 1.27E-15 1.51E-15 1.79E-15 2.39E-15 1132 2.94E-19 8. 59E-16 1.27E-15 1.51E-15 1.79E-15 2.39E-15 1 133 1.31E-07 3.84E-04 5. 70E-04 6. 77E-04 8. 08E-04 l1.09E-03 1133 1.31E-07 3.84E-04 5. 70E-04 6. 77E-04 8. 08E-04 1. 09E-03 1 134 0. OOE+O0 0. 00E+OO 0. OOE+00 0. 0OE+OO 0. OOE+00 0. 00E+00 1134 0. OOE+00 0. OOE+OO 0. OOE+OO 0. OOE+00 0. OOE+00 0. OOE+00 1 135 1.13E-10 3.31E-07 4.91E-07 5. 82E-07 6.95E-07 9.34E-07 1135 1.13E-1O 3.31E-07 4.91E-07 5.82E-07 6. 95E-07 9.34E-07 KR 83M 5. 78E-21 1. 69E-17 2. 50E-17 2.96E-17 3.51E-17 4. 67E-17 KR 85M 2. 65E-l1 7. 76E-08 1.15E-07 1.36E-07 1. 63E-07 2.18E-07 KR 85 1.4 6E-05 4.27E-02 6. 33E-02 7.52E-02 S. 98E-02 1.21E-01 KR 87 6.46E-28 1.89E-24 2. 79E-24 3.29E-24 3. 90E-24 5.14E-24 KR 88 9.21E-15 2. 69E-11 3.99E-11 4. 73E-1l 5. 62E-1 1 7. 51E-1 l XE 131M 5.98E-06 1. 75E-02 2. 60E-02 3. 09E-02 3. 69E- 02 .98E-02 XE 133M 1. 24E- 05 3. 63E-02 5.38E-02 6.39E-02 7. 63E-02 1. 03E-Ol XE 133 B. 01E-04 2.35E*O0 3.48E+00 4.13E+OO 4.94E+0O 6. 67E+00 XE 135 1. 70E-07 4.98E-04 7.38E-04 8. 76E-04 1.05E-03 1. 41E-03

Southern Nuclear Design Calculations Plant: Unit: Calculation Number:

Farley Nuclear Plant I 01 0 2 E 1 & 2 I SM-96-1064-001

Title:

Fuel Handling Accident Doses l Sheet:

19 of 252 Control Room Inventory (Curies) 1.67E-01 3.33E-01 6.67E-01 7.50E-01 1.00E+00 1.25 1.5 1.75 2 2.25 2.5 3 I I

216E-02 2,Q5E-02 136E.02 IA9E-02 7.85E-03 506FE-03 3.24E-03 2.06E-03 131E-03 8.28E-04 5.5E-04 2.E04 I 216F-02 2,05E-2 1.36Em-0 1.19E.02 785E-03 5,O6E-03 3-24E-03 206E-0 1.31E-03 .28E4 5 25E-04 2J.E.04 I 3,19E15 2.8E-15 1.73E;15 148E-AS 905E16 5,43E-16 322E-16 1.90E-16 112E-16 661FE-17 3.89E-17 1.35E-17 I 3.19E-15 2 15 1.73E 15 1.48E-15 9.05E-1 5.43E-16 3,22E-1 1.90E-16 1.12E-16 6.61E-17 3.89E-17 1.3E. I 1,49E.03 IA4E-+/-03 9.18E-04 8l.04E 04 5.26E-04 337E-04 214E-04 135E-04 8.50E-05 535E-05 337E-5 1.33E.5 I 1.49E-03 1.40E-3 9.18E 04 &04E-04 526E-04 337E-04 214E-04 1.35E-04 s50Eo05 5-35E-05 337E-05 133E-05 I Q.ODE+/-00 QO-OEO DOE+/-+M D O.O O OE+OO O.OOE+OO O.ODE+OO OOOE+OO O.OOE+/-+O I DXDDE+/-00 fDODE+/-DO O.OOE+OO D.ODE+OO O.OOF+OO OOE+/-+O D.OOE+/-+0 O.OOE+/-+OO D.O0E+/-ODl0E+/-+O ODDE+0 0120E+/-D0 I 1.27E-06 1-iFI F- 75fiE-0Z &58E 07 4.23E-0 26E-07 1,66E.07 1.03E-07 638E08 3.94E-08 244E-08 9.1ZE-09 I 7.56E-OZ 4.23E-07 266-EO0 1.66E-07 1t03E-0 6.38FB 3.94E-08 2.44E-08 931E-9 I 1.27E-06 1.18E-06 7.54E-17 715EA-Z 5.93E-1 4.83E17 3.92E-17 317E-1 256E-17 2.07E.17 1.67F-17 l O9E-Z E I 2.95E-07 398E-0Z 3.85E 07 3.37E-072 290E-0 2,48E-07 212Ek07 1.81E-07 ?54E-07 1.1E-07 9,55E-08 I 1.66E-01 2,30E01 2A3E-01 ,238E.01f 2,17EF01 1.94E-01 1.73E-01 1.53E-01 1.36E-01 1.21E.01 1.07E-01 8-2E-02 I 6l72E-24 852E-24 L50E-24 L0IE-24 5.56E-24 434E-24 337E-24 2A1E-24 Z2E-24 1556E-24 11E-24 7L24EW25 I 1i01E-10 1,34E.10 ,31E-10 125E.10 1.07E.1 9,OOE-11 7.53E-11 6.28E--11 5.24E-11 4.37E-11 3.65E-1 2.54E-11 I W82E092 .6 E 02 9L98E--02 9.76E-02 8,88E-02 794E02 7.06E02 6.27E-02 5.56E-02 4.93E-02 4.3EF-2 3-44E-02 I 141E-01 1.95E Z2.5E-01 ,2.00E-O 1.82E-01 1.62EF-01 144E-01 1.27E-01 1.13E-01 9.97E-02 W82E2 fi.90E-02 I 9.13E+0O 1.27E+01 1.33E+01f 1.30E+01 1A 9E+/-01 I06E+/-+D 9A.4E+OO &I5E+/-00 7.40E+00 6.55E+00 5BIE+/-DD 4.56E+OO I 122E+/-0I 1.91E-03 262E-3 ,2.fiE-Q3 2.35E-03 26E03 180E-03 1.57E-03 137E-03 1.19E-03 1.04E 03 7.84E-04 I

Southern Nuclear Design Calculations Plant: Unit: Calculation Number:

Farley Nuclear Plant 0 1 0 2 E 1&2 I SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

I 20 of 252 Control Room Inventory (Curies) Exposure (Ci - hr)

Time 4 5 6 7 8 (hr) 2.78E-06 8.33E-03 1.25E-02 3A4QE-5 5.AZE 06 &82,Q-07 1A2E-OZ Z29E+/-QB I 131 5S12EA-2 4-64E-05 5&6EA05 3.40E-05 5.7E-06f 2E 07 14A2E-07 2.29E-08 I 132 BM6E-25 Z716E-18 1.62E.1B 1.95E-19 32 Q ZBIEm 33,E--22 I 133 3S64E-3 3.20E-06 3.98E-06 1-62E-18 I.95E-19 ZtEi2Q ,2.81E-21 3JZE--22 I 134 OeOOEOO 0 DQOE+OO D.OOE+/-+Oo L2lE-Q 3.fiED0 5AlOE-8 Z.98E-9 1s2E-Q2 I 135 3.14E-16 2ZfiEZ09 3-43E-09

,2.08E-O 31.2E-07 510E-08 Z.98EfJ9 1.25E-9 O.OOEiQQ O.OOE+OO O.OOE0+Q DOSOE+/-OQ D.OJE+OO O.DOE+OO O.ODE+OQ, L.OE-+OO KR 83M 803E-27 7.04E-20 .73E-20

.36E-9 19E-10 289E 11 4.22E-12 6&15E-13 KR 85M 3.68E17Z 323EA1O IM6E-09 1.t98E-10 289E4 ~4.22E.12 fi 5E-13 KR 85 ZD2E-1I 178E-4 2.21E-04A 4_6ZE+/-2_ I.9E-18 850E 1M 3S3EA-9 155E-49 KR 87 &2Z7E-34 7.5E-2Z 9JA4E.2Z 5.05E-08 2.67E-08 I1AE4IS 7.48E-09 39flED9 KR 88 12iE-20 1AE13 L39E13 512E-02 323E-2 2.ODE-0M 1.24E-02 7LflE-M XE131M 8231E-12 731E-05 9.07E 05 2XWOEA25 9.31E-2f 3.34E-26 I20E-26 4_29EL2 XE133M 1t72E41 1I51E-04 I1.EB-04 1.23E.11 5-Q3E-12 2.87E-2 1 39E-12 ,6,71E+/-13 XE 133 1IE-9 ,9.8E-03 1.21E-0D2 212E-02 1.1E-02 BAIE-a 5.01E3-O 3 10E3- XE 135 2,36E-13 ZO7E-0f 2 58E-06 4.22Ek02 25E-02 1.58E-2 9.65E-03 5.91E-Q3 2,BIE+O0 17E+OO 1QZE+OQ 65E-01 4.04E-Ol 4.51E-04 2.59E-4 1A9E.04 &54E-05 4.91E-05

Southern Nuclear Design Calculations Plant: Unit: Calculation Number:

Farley Nuclear Plant 0 1 02 [1 &2 I SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

1 21 of 252-Exposure (Ci - hr) 2.50E-02 4.17E-02 8.33E-02 1.67E-01 3.33E-01 6.67E-01 7.50E-01 1.OOE+00 1.25E+00 1.50E+00 1.75E+00 2.OOE+00 3,.59E!04 1.15E-03 3.12E-03 L02E 03 14E-02 2.2E-03 4.94E-03 3.23E-03 2ZE-M03 1.32E-03 .A4 E.04 5.50E-17A 3,47E7f SM5EAZ7 1.74E-f6 4-fiSE+/-1fi 1I.E-15 1.53E-15 2.67E-16 5-9fiE+/-Thi 3S.62E-1 215E-16 1.28E-16  ?.56E.17

.56E-05 Z15E.04 4.65E-16 4.81E-04 7.4E-04 3 33E.04 2AiE-A04 1.38E-04 144E-04 5.96E-16Q 8Z2E--05 5.5QE--05 O.OOE+OQ O.OOE+OO O.OOE+OO O 0OE+ O O.OOE+OO OOE+OO O.OOE+OO O.OOE+OO .0OE+/-+OO O.OOE+OO D.QOE+/-00 2.13E-08 fi78E.08 1.84E-07 4.08E-07 6A.4E.07 I1A8E 07 2170E07 1.72E-0Z 10BE+/-07 ,.3E.08 4A.7E.0 1.70E-18 1.37E-03 3.41E49 2A59E-9 4.53E-18 I19E1AZ 2.60DE1Z fl12E-4 1i35E-17 1Q9E-1z &R6EA-f 7.1E-8 1.57E-O9 512E--O 7.92E-09 2iA4E-8 5,ZE-08 1.33E-0Z 3.27E-0 9.03E-08 7.84E-08 6.72E-08 5J4E 08 4.90E-08

&865E-04 40E-03 L20E-02 3.30E-02 L7.9E-02 2.01E-02 5.68BE-02 5LIE-02 4.58E-02 4.0ZE-02 3.6i1E 02 3AO2E-26 5-99EW26 4.94E-25 1.2ZEA24 26iZE-,24 fi.O5E-25 1t57Et24 1.24E-24 9Q64E.25 L4BEA25 5.79E.25 5.45E-13 8&63E.13 274E-12 L33E-12 196fiE.1 4A.1E.11 1.0ZE.11 2.90E.11 2A6E--1 2.Q7E-11 1.73E.11 1.44E-11f 3.56E-0O4 5L.5EO4 1.81E.03 4,92E-03 1.36E.02 3.24E-2 &23E-3 2.33E-02 2I0E02 1.8E-02 .7E-02 1A8E-02 7.36E-04 1.1ZE03 3.74E-03 1.02E-02 21EWEkQ2 67E-02 1.9E-02 41Z8E 02 4.30E-02 3S.BEU02 3.39E-02 3.00E-Q2 4.76E02 7.56E-2 2.42E.0 fi,.58E-0 1I.E+OO 4.332E+O I11E+OO ZBIE+OO 2.50E+/-0+ 2.22E+OO 1.9ZE+OO tsIE-05 1.60E-05 5.11E05 1.38E.04 31.8E-04 &BZE.04 21.2E.04 .f2E+/-04 5.51E-04 4.82E-04 4.21E04 3.67E!04

Southern Nuclear Design Calculations Plant: Unit: Calculation Number:

Farley Nuclear Plant I 01 0 2 01 1 & 2 I SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

Sum 0-2 hr Exposure (Ci - hr) Sum 2-8 hr (Ci inhaled) (Ci inhaled) 2.25E+00 2.50E+00 3.OOE+00 4.OOE+00 5.OOE+00 6.OOE+oo 7.OOE+00 8.OOE+oo 146E.05 5.34E-04 3.38E-04 245E+/-04 3-94E-05 6.35E-0 102E-06 I1.5E-0Z 5.92E-07 I 1.91E-1B 4A6E-1Z 2,62E.17 1.51E-Z 1.82E-1B 2.18E-19 2262E-20 3.A4E-2 4-42E-20 I 9.94E-0Z 3A.E 05 2BE-+/-05 235E-05 1.54E-05 241EE-06 3 77E-0z 5.90E-08 923E-09 3.BEOBw I O.OOE+OO O.OODE+OO O.ODE+OO O.OOE+/-O0 O.OOE+OO O.OOE+OO O.OOE+OQ O.OOE+0O O.OOE+OO 0.OOE+OO I

&22EA0 2.58E-08 1 68E-08 1.0E-08 156E 09 227E+/-10 32.1 E-11 4.3E-12 2,75E-11 I (Ci-sec/m 3) (Cl-sec/m 3 )

1.60E-08 120E-16 579E-18 4.68E-18 6.92E-1B 78E-18 3.33E-18 1.42E-18 6.06E-19 2.59E-19 3.43E-1Z I 6.68E 0z 4.1BE-08 3,57E-0B 5.67E-08 730E-08 3,86E-08 2.04E-0B 08E-08 572E09 3.15E-07 I 4.26E-Ol 321E-02 2.85E+/-02 4.78E-02 6B.2E-02 422E02 2.52EQ02 12E-02 1 GE--02 3-02E-0M I I.i6E.23 4A.8E-25 3AZE.25 4.84E25 4.92E-25 I1.6E-25 632E-226 221E-26 8 13E-27 Z27E-24 I 2i4E-10 1.20E-1i .00E-I1 tBBEtf 9.10E 12 4A.0E.12 2.13E-12 I103E12 1E-11 I 1.Z5E-01 .3IE-02 I1.iEs02 1.95E-02 2J.BE-02 17i2EO2 10i6E-02 6.56E-03 4.05E-03 123E01 I 3.58E-01 265E-02 2,35Ek02 3.93E-02 55IM-02 3.40E-02 2SDBEk02 1.27E-02 7SBE03 245E1 ' I Z23E+/-0I I.74E+OO 1.55E+OO 2,59Ea+O3QBE+/-00 2.27E+OO I.40E+OO 8.IE-01f 5 30Ef01 1fi2E 1 I 4.63E-03 3.19E-04 2,78E-04 4,55EO4 6.17E-04 3.55E-04 2.a4E-04 1.17E 04 6.73E-05 Z9E03 I

Southern Nuclear Design Calculations Calculation Number:

2 I SM-96-1064-001 Dose Results Thyroid Thyroid Thyroid Skin Skin Body Body DCF Dose Dose Exposure DCF DCF (Rem/Ci) (REM) (REM) 0-720 hrs (REM-M3/ Dose (REM-M3/ Dose 0-2 hr 2-8 hr (Ci-Sec/M3) Sec-Ci) (REM) Sec-Ci) (REM),

I 131 1.61E+01 6.5IE.01 4239EQ02 3.0ZE-02 135E-03 5.529E02 t0QZE--04 132 IAOE+/-0fi I 1.20E-14 2.78E-16 563E-15 I.0E.01 fi69E-f6 3,55E-01 &72EA-Z I 133 1.Z9E-01 fi.4E-03 29ZE7E03 .90E-02 2.65E-04 9.11E-02 IA18E-05 134 IAOE+/-05 I 0LQ0E400 0.0QE+0O OOOE+OO 1-42E-0I 0.00E+/-00 4.11E-01 0.0DE-00 i-f E+/--Q4 I 135 2.55E-05 &52E-0Z 2A54E-06 7.8--02 1.92E-0Z 2A9E-1 2SMfE-Mf TOTAL 1.63E01 658E-01 TOTAL .61E-03 TOTAL 1.19E-04 I Skin DCF NG Skin Body DCF NG Body (REM-M3/ Dose (REM-M3/ Dose Sec-Ci) (REM) Sec-Ci) (REM)

KR 83M O.OOE+00 D0.DOE+/-0 1.27E-05 &SHE23 KR 85M 4.97E-02 4.89E+/-0B 2.31 E-02 9-92E40M KR 85 4.84E-02 3.52EM- 3.31 E-04 L.05E-05 KR 87 3.36E-01 4.68E!24 1.33E-01 &0flE-2f6 KR 88 7.76E-02 2.29E--ff 3.38E-01 4a6fE--12 XE131 M 1.33E-02 3.9fiE-30 1.25E-03 t62E-05 XE133M 2.96E-02 t78E-02 4.29E-03 IA2E--04 XE133 9.67E-03 3.8 3-01f 4.96E-03 857E-03 XE135M 2.14E-02 1.57E+/-04A 6.37E-02 2.03E--

TOTAL 4AOE01( 4-42E-0I TOTAL =1E-03 8W5EQ03 (w/ iodine) (w/ iodine)

Southern Nuclear Design Calculations Plant: l Unit: Calculation Number:

Farley Nuclear Plant I 01 02 E 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

1 24 of 252l E. Realistic Dose Estimate for the Individual Designated to Close the Equipment Hatch As noted in reference 23 and supporting documentation, the thyroid dose to the individual(s) designated to close the equipment hatch may be excessively high if he were to remain in the containment with design basis FHA parameters. Thus a realistic estimate of the thyroid (and whole body and skin) dose to this individual is made based on the following parameters.

Power - reference 28 indicates that fuel drops from up to 20 feet are not expected to cause damage to a fuel assembly that result in radioactive releases. The FSAR (ref. 11 a) indicates only the outer row of fuel pins might be assumed to fail, so the power is taken to be 2775 MWt / 157 assy / 17 rows = 1.04 MWt.

Decay Time - historically FNP has started fuel movement at about 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> due to limitations in spent fuel pool heat removal capability. To bound future improvements, the accident is assumed at 120 hrs.

Release Fraction - the average peaking factor for the core is 1.0. Per ref 23 the pool DF for elemental iodine is expected to be 500. Ref 1 e indicates the expected gap fraction for I13 (by far the limiting isotope) is 2.2% as compared to 8% used in input file FNPGAP31 (earlierversion of FNPGP9Swith a constant core iventori inarghi of 1.02). With a gap activity 99.75% elemental and .25% organic the expected iodine release fraction to the containment air will be (.9975/500 + .0025) x 2.2/8 = 1.236 x 10-3. Although reference 1ie indicates the expected gap fraction for noble gases (except Kr85) is less than the noble gas release modeled, it is unchanged for this evaluation.

Containment Volume - Operation of the containment cooling system will quickly dilute the released activity throughout the free volume of the containment, so use 2 x 106 ft3 . For the whole body dose, the individual is assumed exposed to the free volume above the operating deck (2 x 106 ft3 ) x (0.822),

which results in a Geometry Factor of 1173 / [(2 x 106)(0.822)](0 38)= 9.3.

Purge Rate - a slower purge will maintain the released activity inside the containment longer. Use of the slow speed exhaust rate will maximize the operator dose. Note however, the containment purge radiation monitor will still rapidly reach the isolation setpoint, thus purge will be rapidly terminated and the operator dose will be calculated based on the initial activity released into the containment.

ACTIVITY RELEASED TO ENVIRONMENT AND IN EACH NODE AT END OF... 2.778E-06 (HRS)

ISO NAM ENV. Contnmnt CtrlRoom I 131 1.887E-06 8.235E-01 9.950E-10 1131 is clearly the limiting I 131 2.363E-06 1.031E+00 1.246E-09 isotope. The total I'3' -

I 132 8.056E-22 3.515E-16 4.248E-25 inventory released into the I 132 1.009E-21 4.402E-16 5.319E-25 containment is 1.9 Ci.

I 133 6.967E-08 3.040E-02 3.673E-11 I 133 8.724E-08 3.807E-02 4.600E-11 I 135 1.480E-11 6.456E-06 7.801E-15 I 135 1.853E-11 8.085E-06 9.769E-15