Text

to Calculation PNPS-1-ERHS-XXII.A-2, Radiological Consequences of Design Basis Fuel Handling Accident Using the Alternate Source Term, Attachment 2 to 2.04.003
	ML041130100
Person / Time
Site:	Pilgrim
Issue date:	04/14/2004
From:	Balduzzi M; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2.04.003 PNPS-1-ERHS-XXII.A-2, Rev 1
	Download: ML041130100 (49)
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Attachment 2 to 2.04.003 Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Plant Proposed Amendment to the Technical Specifications Entergy Calculation No. PNPS-1-ERHS-XXII.A-2, 'Radiological Consequences of a Design Basis Fuel Handling Accident using the Alternate Source Term", Rev. 1 (48 pages). (Computer run pages are not included) 1

CALCULATION COVER PAGE RType 64.65 EIP-2 EhIP-3 MJAF ZPNPS DVY Calculation No. PNPS-I-ERHS-XXII.A-2 This revision incorporates the following MERLIN DRNs or Minor Calc Changes: Sheet I of 181

Title:

Radioloeical Consequences of a Design Basis Fuel Handling Accident Q Usina the Altemative Source Term DNQR Design Basis Calculation?

Discipline: Systems and Safety Analysis E Yes El No This calculation supercedestvoids calculation: ERHS-XXII.A-2. Revision 0 Modification No./Task No/ER No: N/A El No software used El Software used and filed separately (Include Computer Run Summary Sheet). If "YES", Code: _

3 Software used and filed with this calculation. If "YES", Code: RADTRAD System NoJName: N/A Component No.Name: N/A (Attached additional pages ifnecessary)

Print / Sign STATUS OTHER REV (Prel. PREPARER REVIEWER/DESIGN REVIGN APPROVER DATE Pend. A, A VERIFIER DESIGN V, S) VERIFIER P. Compagnone P. T. Karatzas S. Wollman V

_ Pend. Ax ______ /_ 03

CALCULATION SHEET Calc No. PNPS-1-ERHS-XXll.A-2, Entergy Revision 1 Sheet 2 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST RECORD OF REVISIONS Calculation No. PNPS-ERHS-XXII.A-2 Revision No. Description of Change Reason For Change 1 Full revision Revised to reflect revisions in

__ referenced design documents

_ _ __ _ __ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I _ _ _ _ _ _ _ _ _

CALCULATION SHEET CalcNo. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 3 of 181.

Subiect: Radiological Consequences of a Desien Basis Fuel Handling Accident Using the AST CALCULATION

SUMMARY

PAGES Page 1 of 2 Calculation No. PNPS-1-ERHS-XMI.A-2 Revision No. 1 CALCULATION OBJECTIVE:

To determine the potential radiological consequences offsite and to Control Room personnel following a design basis fuel handling accident for various plant configurations using the alternative source term.

CONCLUSIONS:

With the SGTS in operation fuel may be moved 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown. Without SGTS in operation, fuel may be moved 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after shutown. In either case CRHEAFS is not required to be in operation.

The potential dose consequences of a FHA are as follows:

Operation Dose (Rem)

Case Fuel Decay SGTS CRHEAFS CR CR EAB LPZ (RI3 vent*) I(RI3 trucklock*)'

1 24 hrs YES YES 2.2E-05 2.2E-05 3.9E-01 1.3E-02 2 24 hrs YES NO 2.6E-05 2.6E-05 3.9E-01 1.3E-02 3a 24 hrs NO YES 2.7 1.5 6.5 6.5E-02 3b 48 hrs NO YES 2.4 1.3 4.9 4.9E-02 3c 72 hrs NO YES 2.1 1.1 4.2 4.2E-02 4a 24 hrs NO NO 3.3 1.8 6.5 6.5E-02 4b 48 hrs NO NO 2.9 1.5 4.9 4.9E-02 4c 72 his NO NO 2.5 1.4 4.2 4.2E-02 4d 96 hrs NO NO 2.3 1.2 3.8 3.7E-02 4e 120 hrs NO NO 2.1 1.1 3.4 3.4E-02 Regulatory limit 5 5 6.3 6.3 For Cases 3 and 4, only.

ASSUMPTIONS:

The FHA scenario is as detailed in RG 1.183.

GE14 fuel assemblies are involved in the FHA.

A conservative radial peaking factor of 2.1 is applied to all damaged fuel assemblies.

Inflow into the Control Room during normal ventilation operation is 500 cfm.

Inleakage into the Control Room during CRHEAFS operation is 500 cfm.

CALCULATION SHEEM Caic No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 4 of _181 .

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST CALCULATION

SUMMARY

PAGES (CONTINUED)

Page 2 of 2 DESIGN INPUT DOCUMENTS:

Regulatory Guide 1.183 Dept. Doc. OSD 00-088 Technical Specifications: 1.0, 3.7.A/B Calculation PNPS-1-ERHS-I.G-7, Rev. 0 Calculation PNPS-1-ERHS-X.G4-127, Rev. 0 Calculation PNPS-1-ERHS-XI.L-28, Rev. 0 Calculation PNPS-1-ERHS-II.A-1, Rev. I Calculation PNPS-1-ERHS-II.A-2, Rev. 1 Calculation PNPS-1-ERHS-II.B-4, Rev. 1 Drawing M23 Procedure 2.2.46 Calculation C15.0.3411 PNPS Memorandum C15-SMP-001 AFFECTED DOCUMENTS:

N/A METHODOLOGY:

The computer program RADTRAD was used to calculate the TEDE at the EAB and LPZ and to CR personnel assuming a design basis fuel handling accident in accordance with RG 1.183 Appendix B.

Various configurations of the SGTS and CRHEAFS were considered.

CALCULATION SHEET Caic No. PNPS-I-ERHS-XXII.A Entergy Revision I Sheet 5 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST TABLE OF CONTENTS Section Pase RECORD OF REVISIONS ............................ 2 CALCULATION SUM1rvIARY PAGES ....................... 3 TABLE OF CONTENTS ........................... 5 LIST OF FIGURES ............................ 8 LIST OF TBLES ..................... ... 8 LIST OF EFFECTIVE PAGES .................. . 9

1.0 BACKGROUND

.... 10 2.0 PURPOSE ...... 10 3.0 METHOD OF SOLUTION . . . .11 4.0 ASSUPIONS .. 11 5.0 INPUT AND DESIGN CRIIERIA .. 12 5.1 Release Activity ..................................................... 12 5.2 Dose Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

6.0 REFERENCES

...................................................... 16 7.0 CALCULATIO N/ANAjLYSIS ..................................................... 18 7.1 FIA Activity Release ..................................................... 18 7.2 Activity Release From Refuel Floor . . ................................................... 20 7.3 Case 1: FHA Dose Consequences With SGTS and With CRHEAFS ....................... 21 7.3.1 Case 1 Release Scenario......................................................... 21 7.3.2 Case 1 Dose Analysis......................................................... 23 7.3.3 Case I RADTRAD Input Data......................................................... 23 7.3.4 Case 1 RADTRAD Input Files. 26 7.3.5 FHA Case 1 Dose Consequences.27

CALCULATION SHEET Calc No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 6 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST TABLE OF CONTENTS (Continued)

Section Pase 7.4 Case 2 - FHA Dose Consequences With SGTS & NO CRHEAFS ........................... 27 7.4.1 Case 2 Release Scenario ............................ 27 7.4.2 Case 2 Dose Analysis ........................... 29 7.4.3 Case 2 RADTRAD Input Data....................................................................................29 7.4.4 Case 2 RADTRAD Input Files........................... 32 7.4.5 FHA Case 2 Dose Consequences........................... 32 7.5. Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS ... 33 7.5.1 Case 3 Release Scenario......................................................... 33 7.5.2 Case 3 Dose Analysis ......................................................... 35 7.5.3 Case3 RADTRAD Input Data......................................................... 35 7.5.4 Case 3 RADTRAD Input Files........................................................ 38 7.5.5 FHA Case 3 Dose Consequences........................................................ 39 7.6 Case 4: FHA Dose Consequences With NO SGTS and NO CRHEAFS .................. 40 7.6.1 Case 4 Release Scenario........................................................ 40 7.6.2 Case 4 Dose Analysis ........................................................ 42 7.6.3 Case 4 RADTRAD Input Data.............................. 42 7.6.4 Case 4 RADTRAD Input Files............................... 45 7.6.5 FHA Case 4 Dose Consequences.............................. 45 8.0 RESULTS ................................ 46 APPENDICES - RADTRAD Computer Program Files ....................................... 48 Appendix A - FHA Case 1 Computer Run ....................................... 49 Appendix B - FHA Case 2 Computer Run . . . .........................

64 Appendix C1- FHA Case 3a Computer Run . . . ........................

77 Appendix C2 - FHA Case 3b Computer Run ... .................................... 92 Appendix C3 - FHA Case 3c Computer Run . . . ........................

107 Appendix DI - FHA Case 4a Computer Run . . . .......................

122 Appendix D2 - FlHA Case 4b Computer Run . . . .......................

134 Appendix D3 - FHA Case 4c Computer Run . . . .......................

146 Appendix D4 - FHA Case 4d Computer Run . .. ......................................

158 Appendix D5 - FHA Case 4e Computer Run . . . .......................

170

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 7 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST TABLE OF CONTENTS (Continued)

Section Pane - Dept. Doc. OSD 00-088 (2 pages) ............................................... lAl-1 - Memorandum From S. Paranjape (1 page) ..................................... 1A2-1 - Calculation Design Verification (6 pages) ..................................... 1A3-1 - PNPS Memorandum C15-SMP-001 (1 page) ................................. 1A4-1

CALCULATION SHEET CaicNo. PNPS-I-ERHS-XX11.A-2 Entergy Revision 1 Sheet 8 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST LIST OF FIGURES Figure No. Page Figure 7-1 FHA Case 1 Scenario: SGTS & CRHEAFS . ........................................................

22 Figure 7-2 FHA Case 2 Scenario: SGTS & NO CRHEAFS ............................................ ..... 28 Figure 7-3 FHA Case 3 Scenario: NO SGTS & CRHEAFS . .................................................

34 Figure 7-4 FIA Case 4 Scenario: NO SGTS & NO CRHEAFS . ...........................................

41 LIST OF TABLES Table No. Paee Table 7-1 FHA Case 1: With SGTS & With CRHEAFS-Dose Consequences....................... 27 Table 7-2 FHA Case 2: With SGTS & Without CRHEAFS -Dose Consequences.................. 32 Table 7-3 FHA Case 3: Without SGTS & With CRHEAFS - Dose Consequences.................. 39 Table 7-4 FHA Case 4: Without SGTS & Without CRHEAFS - Dose Consequences............. 45 Table 8-1 Fuel HandlingAccident TEDE........................................................... 46

CALCULATION SHEET Calc No. PNPS-l-ERHS-XXII.A-2 Entergy Revision I Sheet 9 of 181!

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST LIST OF EFFECTIVE PAGES Calculation Number: PNPS-1-ERHS-XXII.A-2 Revision Number: I Page Revision ito 181 1 (Al - Al-2) I (A2-1) 1 (A3 A3-6) 1 (A4-1) I

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 10 of 181.

Subject:

Radiological Conseauences of a Design Basis Fuel Handling Accident Using the AST

1.0 BACKGROUND

GE14 (IOx10) fuel assemblies were loaded into the Pilgrim Nuclear Power Station (PNPS) reactor core during Reload 13 [1]'. It is anticipated that all future reloads will use GE14 fuel assemblies, as well.

The radiological consequences of a fuel handling accident (FHA) involving GE14 fuel assemblies, using the guidance of Regulatory Guide 1.25 [2] were evaluated in calculation PNPS-1-ERHS-XIII.Y-64, "Design Basis Fuel Handling Accident For RFO #13 (GE14, 10xl0) Fuel" [3].

The worst-case activity release for a FlHA results from the accidental dropping of a fuel assembly onto the reactor core during fuel movement [18A, 18B]. A fuel assembly dropped in the spent fuel pool would result in fewer fuel rods damaged due to a significantly smaller drop height [8]. This analysis evaluates the potential radiological consequences for a FEHA involving GE14 fuel in the PNPS core and applying the alternative source term (AST) in accordance with the guidance provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Plants" [4].

2.0 PURPOSE The purpose of this calculation is to estimate the radiological consequences to the public offsite, at the exclusion area boundary and the low population zone, and to personnel in the Control Room in the event of a design basis fuel handling accident at Pilgrim Nuclear Power Station using the alternative source term in accordance with RG 1.183 and for various plant configurations.

1 References are indicated in brackets [ 1.

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 11 of 181.

Subject:

Radioloaical Consequences of a Desien Basis Fuel Handlin2 Accident Usine the AST 3.0 METHOD OF SOLUTION The evaluation of the radiological consequences due to a design basis fuel handling accident (FHA) involving GE14 fuel assemblies in the PNPS reactor core and the alternative source term (AST) is based on Regulatory Guide 1.183 [4] assumptions and PNPS specific input. The number of damaged fuel rods was used to determine the fraction of the total core activity that is available for release to the environment. All the damaged fuel assemblies are assumed to leak the inventory specified in RG 1.183 to the coolant. The assumptions of RG 1.183, Appendix B are employed to determine the fraction of the released activity that eventually is available for release to the environment. The computer program RADTRAD [5] was used to calculate the doses at the exclusion area boundary (EAB), at the low population zone (LPZ), and in the control room (CR), based on the activity available for release to the environment and PNPS-specific atmospheric dispersion coefficients. Dose consequences are evaluated with and without the availability of the standby gas treatment system (SGTS) filters and the with and without the availability of the control room high efficiency air filtration system. For scenarios without SGTS available, release to the environment from the reactor building refuel floor is assumed to occur via the normal ventilation system out the reactor building vent or directly out of the reactor building trucklock.

4.0 ASSUMEPTIONS

1. The FHA scenario is as detailed in RG 1.183 Appendix B.

-2. GE14 fuel assemblies are involved in the fuel handling accident [1].

3. A conservative value of 2.1 for radial peaking factor is applied to ail damaged fuel rods [1].

4. For purposes of the radiological evaluation, additional inflow into the Control Room during normal ventilation equals 500 cfm, approximately equal to the intake flow rate into the Control Room. In reality, such a high inflow would not be acceptable for heating and air conditioning purposes.

5. During CRHEAFS operation the unfiltered inleakage is less than or equal to 500 cfm.

CALCULATION SHEET Calc No. PNPS- I-ERHS-XXII.A-2 Entergy Revision I Sheet 12 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 5.0 INPUT AND DESIGN CRITERIA The dose consequences due to a postulated design basis fuel handling accident (FHA) are based on the following data. The specific computer program parameter input values for receptor locations are listed in the subsequent sections.

5.1 Release Activity In the event of a fuel handling accident it is assumed that a number of fuel rods are damaged and radioactivity is released from the damaged rods. The number-of.rods assumed damaged and the consequent activity available fbr release is as given in Ref. I and RG 1.183.

1. Reactor power level = 2038 MWt.

This accounts for the reactor current licensed power level of 2028 MWt [6A] plus a measurement uncertainty of 0.5%. This is consistent with core inventory as determined in Reference 7.

2. FHA AST nuclide activities at 2038 MWt [7, Table 7-4]

Core Activity l 1Core Activity Core Activity Nuclide (CiIMWt) Nuclide I (Ci/MWt) I Nuclide I (CiIMWt)

Kr-85 3.14E+02 1-131 2.63E+04 Xe-133 5.52E+04 Kr-85m 7.94E+03 __1-132 3.81E+04 Xe-135 2.50E+04 Kr-87 1.55E+04 I-133 5.51E+04 =

Kr-88 2.19E+04 I_-134 6.07E+04 Cs-134 3.06E+03

_ 1-135 5.14E+04 _ 2s-136 1.31E+03 Rb-86 3.56E+01 I _I _ I I Cs-136 3.14E+03

3. . Number of GE 14 fuel rods damaged = 151 [1]

4. Total number of fuel rods in core = 45512 [1]

5. Radial peaking factor= 2.1 [1]

This factor is conservatively applied to all damaged fuel rods.

6. Delay time before fuel movement - assumed.

CALCULATION SHEET Caic No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 13 of 181.

Subiect: Radiological ConseQuences of a Design Basis Fuel Handling Accident Using the AST 5.0 INPUT AND DESIGN CRITERIA (Continued) 5.1 Release Activity (Continued)

7. Fraction of fission product inventory in gap that is released to coolant [4, Table 3]

Group Fraction I-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other lodines 0.(5 Alkali metals (Cs, Rb) l 0.12 Note: the maximum burnup at the end of Cycle 15 is predicted to be 51.7 GWDIMTU [19].

Therefore, the condition of Footnote 11 of RG 1.183 [4, page 1.183-14] is met and the fractions given in RG 1.183, Table 3 apply.

8. Iodine species fractions released to pool [4, Appendix B]:

Aerosol = 95%

Elemental = 4.85%

Organic=0.15%

9. Depth of water above fuel in refuel cavity > 23 ft [8].

Core is approximately at elevation 70 ft. Water level during refueling is at elevation 116 ft. (See

§1.0).

10. Effective iodine decontamination factor = 200 [4, Appendix B]

11. Noble gas decontamination factor = 1.0 [4, Appendix B]

12. Particulate decontamination factor = infinite [4, Appendix B]

13. Release to environment occurs over a 2-hr time period [4, Appendix B]

14. Iodine species fractions above water [4, Appendix B]

Elemental = 57%

Organic = 43%

CALCULATION SHEET Caic No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 14 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 5.0 INPUT AND DESIGN CRITERIA (Continued) 5.2 Dose Analyses

1. Refuel floor air volume = 7.0 x 105 ft3 [9, page 141

2. Standby gas treatment system (SGTS) iodine filter efficiency = 99% [6B3 [17]

3. Normal operation control room ventilation envelope comprises the Control Room (CR), the cable spreading room, computer room, storage room and corridor. Not including the CR, the net volume of the envelope = 28,620 ft3 [10, page 63].

4. Control Room net volume = 34,280 ft3 [20].

5. Total normal operation control room ventilation envelope net volume = (28,620 ft3 + 34,280 ft3 )

62,900 ft3 [from #3 and #4, above).

6. Control Room normal ventilation inflow/exhaust rate = 1100 cfm

a. Normal operation intake/exhaust rate (for ventilation envelope) = 1090 cfm [10, page 60-63]

This includes a normal ventilation intake of 1000 cfm and 90 cfm inflow, due to access/egress through doorways in the total envelope, to the total volume of 62,900 ft3 . The portion going to the CR volume of 34,280 ft3 = [(34,280 ft3 /62,900 ft3 )(1090 cfm)] _ 600 cfm.

b. Assumed additional inflow from surrounding ventilation zones = 500 cfm [§4.0, #5]

c. Theref6re, the total inflow to the CR = 600 cfm + 500 cfm = 1100 cfm.

d. Intake iodine filter efficiency = 0 (no iodine filters)

7. Control room filtered intake - 0.5 hr [ 1, page 11] - 720 hrs ( post-accident):

Filtered intake rate = 1100 cfm [6C][11, page 12]

Intake iodine filter efficiency = 95% [6C]

8. Unfiltered inflow to CR (0.5 hr - 720 hrs, post-accident):

Unfiltered inleakage = 10 cfm (doorway) [12]

Unfiltered inleakage = 500 cfm (other locations) [§4.0, #5]

9. Control Room occupancy factors [4, pg 1.183-18]:

Time Occupancy Factor 0 - 24 hrs 1.0 1 - 4 days 0.6 4 - 30 days 0.4

CALCULATION SHEET Calc No. PNPS-I.ERHS-XXll.A-2 Entergy Revision I Sheet 15 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 5.0 INPUT AND DESIGN CRITERIA (Continued) 5.2 Dose Analyses (Continued)

10. Breathing rates:[4, pg 1.183-16]: BrahngRt Time

Breathin-, Rate 0 - 8 hrs 3.5 x 10 Tm3/s (0 - 720 hrs for CR) 8 - 24 hrs 1.8 x 104 m3 /s 24 - 720 hrs 2.3 x 10-4m3/s

11. Atmospheric dispersion factors (x/Q):

_. Eievated Ground-level - Reactor Building (Main Stack)

Vent Trucklock Time Interval X/Q xi /Q (hrs) * (sImr) (s/m Wm)

EAB fumigation 5.85E-04

[13] (0-2) 0 -2 2.61E-03 LPZ Fumigation 1.91E-05

[13] (0-4)

_0-8 2.58E-05 CR 0-2 1.85E-03 l 9.87E04 7.32E-07

[141 I 1

12. Regulatory dose limits:

TEDE Ref.

CR 5 rem [15]

EAB 6.3 rem [4, Table 6]

LPZ 6.3 rem [4, Table 6]

Time intervals are relative to time post-accident.

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXH.A-2 Entergy Revision I Sheet 16 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST

6.0 REFERENCES

1. Office Memorandum Dept. Doc. OSD 00-088, "Design Criteria Input Request for Cycle 14 Reloaa Core Design Change, PDC 01-03," from S. Paranjape - Attachment 1 (2 pages).

2. USNRC Regulatory Guide 1.25 (Safety Guide 25), "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 12, 1972.

3. Calculation PNPS-1-ERHS-XIII.Y-64, Revision 0, "Design Basis Fuel HandlingAccident For RFO

13 (GE14, 10x10) Fuel"

4. USNRC Regulatory Guide 1.183, 'Altemative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Plants," July 2000.

5. S&SACP35, "RADTRAD Qualification Conformance Matrix," Revision 0. Software Catalog

01812.

6. PNPS Technical Specifications A. Amendment 201, "1.0 Definitions - Design Power."

B. 3.7.B.1 "SGTS" C. 3.7.B.2 "CHREAFS"

7. Calculation PNPS-1-ERHS-XXI.A-1, "PNPS AST Nuclide Inventory," Revision 0.

8. Drawing M23, Equipment Location Sections DD &LL, Rev. E8.

9. Calculation PNPS-1-ERHS-X.G4-127, "Environmental Qualification Airborne Radiation Levels in the Reactor Building Secondary Containment," Revision 0.

10. Calculation PNPS-1-ERHS-XI.L-28, Revision 0, "CR Habitability Radiological Reassessment for a DBA LOCA."

11. Procedure 2.2.46, "Control Room Cable Spreading Room, and Computer Room Heating, Ventilation, and Air Conditioning System," Revision 38.

12. USNRC, NUREG-0800, Standard Review Plan 6.4, "Control Room Habitability System," Rev. 2, July 1981

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXIT.A-2 Entergy Revision I Sheet 17 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 6.0. REFERENCES (Continued)

13. Calculation A. PNPS-1-ERHS-II.A-1, "Accident X/Q's for Main Stack and Turbine Building Releases,"

Revision 1.

B. PNPS-1-ERHS-II.A-2, "Accident x/Q's for Reactor Building and Yard Area Releases,"

Revision 1.

14. Calculation PNPS-1-ERHS-II.B-4, Revision 1, "Control Room and Technical Support Center Accident x/Q's Using ARCON96."

15. Code of Federal Regulations Title 10 Part 50.67 (IOCFR50.67).

16. NUREG/CR-6604, "RADTRAD: A Simplified Model For RADionuclide Transport and Removal And Dose Estimation," December 1997.

17. Calculation S&SA 154, "SGTS Train Efficiency For Radioiodine Removal by Dual Bank Carbon Adsorbers," Revision 0.

18. PNPS Final Safety Analysis Report:

A. Section 14.5.5, "Fuel Handling Accident", Rev 22, October 1999 B. Appendix R, Section R.3.4, "Refueling Accident", Revision 6, July 1986 C. Section 7.18, "Reactor Building Isolation and Control System", Revision 10, July 1989

19. Memorandum from S. Paranjape to P. Compagnone, August 12, 2003 - Attachment 2 (1 page).

20. Calculation C15.0.3411, "Control Room Volume, 'Revision 0.

21. PNPS Memorandum, C15-SMP-001, dated 4/01/04, from S.M. Paranjape to Phil Karatzas, "Fuel Handling Accident Calculation ERHS-XXII.A-2, Question on Cycle 13 Data Versus Current Cycle 15 Fuel Configuration".

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 18 of 818.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS The potential radiological consequences of a postulated fuel handling accident are evaluated for various plant configurations to determine acceptable plant configurations during fuel movement. The data and assumptions listed in Section 5.0 are used with the computer program RADTRAD to calculate the doses in the Control Room (CR) and offsite, at the EAB and LPZ, as a result of the postulated releases to the environment.

7.1 HA Activity Release The design basis fuel handling accident, as described in Regulatory Guide 1.183 "Alternative Radiological Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors," [4] is modeled as an essentially instantaneous release of the radioactivity from the damaged fuel to the coolant and then to the refuel floor. Release from the refuel floor to the environment occurs at a rate such that essentially all the available radioactivity is released to the environment in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months The activity available for release from the refuel floor is determined as follows.

1. Reactor power = 2038 MWt [§5.1 #1]

2. Nuclide specific activities (Ci/MWt)- [§5.1, #2].

3. Fraction of iodine activity released to coolant = 0.05 [§5.1, #7]

4. 1-131 specific activity adjusted to correspond to 0.05 release = 4.21E+04 Ci/MWt

Fraction of iodine activity in coolant = 0.05

Fraction of I-131 activity in coolant = 0.08 [§5.1, #7]

1-131 specific activity is adjusted by a factor of (0.08)/(0.05) = 1.6.

Core A113 1 = 2.63E+04 Ci/MWt [ item #2, above]

Adjusted A1131 = 2.63E+04 Ci/MWt (1.6) = 4.21E+04 CiJMWt

CALCULATION SHEET CaIcNo. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 19 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.1 FHA Activity Release (Continued)

5. Fraction of NG activity released to pool = 0.05 [§5.1, #7]

6. Kr-85 specific activity adjusted to correspond to 0.05 release = 6.28E+02 Ci/MWt

Fraction of noble gas activity in coolant = 0.05

Fraction of Kr-85 activity in coolant = 0.10 [§5.1, #7]

Kr-85 specific activity is adjusted by a factor of (0.10)1(0.05) = 2.0 Core AC8 5 = 3.14E+02 Ci/MWt [item #2;-above]

Adjusted A>f 5 = 3.14E+02 Ci/MWt (2.0) = 6.28E+02 Ci/MWt

7. FHA effective power (PFHA) = 14.2 MWt This is the effective power generated by the damaged fuel rods at a radial peaking factor of 2.1.

Power=2038 MWt [§5.1,#1]

Number of damaged fuel rods = 151 [§5.1, #3]

Total numberof fuel rods = 45512 [§5.1, #4]

Radial peaking factor = 2.1 [§5.1, #5]

PCRDA (2038MWtXlslX2.1) -14.2 MWt 45512

CALCULATION SHEET CaIc No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 20 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.2 Activity Release From Refuel Floor

1. Fraction of iodine inventory available for release from the refuel floor (F1 ) = 2.5E-04

Fraction of iodine activity in pool = 0.05 [§5.1, #7]; fraction of 0.08 for I-131 accounted for by adjusted 1-131 specific activity [§7.1, #4]

Pool effective iodine decontamination factor = 200 [§5.1, #10]

.. F1 = (0.05)/(200) = 0.00025

2. Fraction of noble gas activity available for release from the refuel floor (FNG) = 5.OE-02

Fraction of noble gas activity in pool = 0.05 [§5.1, #7], fraction of 2.0 for Kr-85 accounted for by adjusted Kr-85 specific activity [§7.1, #6]

Pool effective noble gas decontamination factor = 1.0 [§5.1, #11]

FNG = (0.05)/(1.0) = 0.05

3. Fraction of alkali metal activity available for release from refuel floor (Fp) 0 Particulates are retained in the pool [§5.1, #12].

4. Activity release rate from the refuel floor = 6.908/hr = 80600 cfm The activity released to the refuel floor must be released to the environment in a two-hour time period [§5.1, #13]. The activity remaining in the building at any time, as accounted for by the computer program RADTRAD follows an exponential decay curve. In order to release essentially all the activity in a 2-hr time period, the activity release rate for use with RADTRAD is determined as follows. Assume that at the end of the 2-hr time period only 106 of the initial activity at t = 0 hr.

AO, remains.

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 21 of 181.

Subiect: Radioloaical Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.2 Activity Release From Refuel Floor (Continued) 10 AO=Aoe~

10 = ek In10- 6 = -A(2hr) 6.9081hr = A This release rate corresponds to a volumetric release rate (RR) of 80600 cfm.

(700000ft3X6-908/

RR= . hr = 80600cfm 60 7/hr 7.3 Case l: FHA Dose Consequences With SGTS and With CRHEAFS 7.3.1 Case I Release Scenario For this scenario, activity is released to the environment through the standby gas treatment system (SGTS) filters and out the main stack. As indicated in Reference ISC, in the event of a fuel handling accident, isolation of the Reactor Building occurs before any radioactivity can be released directly from the Reactor Building to the environment [18C]. The control room high efficiency air filtration system (CRHEAFS) is in operation within 30 minutes of accident initiation. For this case it is assumed that fuel is moved 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shut down. Additional delay in fuel movement provides additional decay time and, thus, lower doses. The activity release scenario is shown on the following sketch. The plant scenario file used with the computer program RADTRAD is detailed in the following sections.

CALCULAriON SIHEET CaIc No. PNPS-I-ERHS-XXIL.A-2 Entergy Revision I Sheet 22 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: MIlA Dose Consequences With SGTS and CRLIEAFS (Continued) 7.3.1 Case I Release Sceniario (Continztied)

Figure 7-1 FIIA Case 1 Scenario: SGTS & CRHEAFS

CALCULATION SHEET CaicNo. PNPS-I-ERHS-XXII.A-2... Entergy Revision I Sheet 23 of 181.

Subiect: Radioloeical Consequences of a Design Basis Fuel Handline Accident Usine the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: FHA Dose Consequences With SGTS and CRHEAFS (Continued) 7.3.2 Case 1 Dose Analysis

1. Activity decay time = 24 hrs.

Any additional decay would result in lower doses.

2. SGTS iodine filter efficiency = 99% [§5.2, #2]

3. Release is at the elevated level from the main stack.

4. Control Room intake/exhaust rate:

- 0 - 30 minutes = 1100 cfm [§5.2, #6]

- 0.5 hr - 720 hrs: intake =,1100 cfm, filter efficiency = 95% [§5.2, #7]

5. Unfiltered inleakage into CR (0.5 hr- 720 hrs) = 510 cfm [§5.2, #8]

7.3.3 Case I RADTRAD Input Data The computer program RADTRAD is used to determine the dose consequences offsite at the exclusion area boundary (EAB) and at the low population zone (LPZ) and to the control room (CR) based on the case 1 scenario. The control room high efficiency air filtration system (CRHEAFS) will not be in operation until 30 minutes after accident initiation. For the first 30 minutes post-accident the normal ventilation system, which provides outside air for the CR, the cable spreading room, the computer room, storage room and corridor, will be in operation. The input is described in the following sections. All input times are post-accident.

CALCULATION SHEET CaicNo. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 24 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handline Accident Usine the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: MIA Dose Consequences With SGTS and CRHEAFS (Continued) 7.3.3 Case I RADTRAD Input Data (Continued)

Compartments

1. Refuel floor (3-Other)

a. Volume = 700,000 ft3 [§5.2, #1]

b. Source term fraction = 1.0 (all the activity corresponding to the FHA equivalent power)

c. Compartment features - none

2. Atmosphere (Environment)

3. Control Room volume = 34,280 ft3 [§5.2, #4]

Transferpathways

1. Refuel floor to atmosphere

a. Transfer mechanism - filter

b. Flow rate and Filter efficiencies (percent) [§7.2, #4][§7.3.2, #2]

Time (h) Flow Rate (cfm) Aerosol Elemental I Organic I 0.0 80600 99 99 99 2.0 0 0.0 0.0 0.0

2. Atmosphere to CR (intake)

a. Transfer mechanism - filter

b. Flow rates and Filter efficiencies (percent) [§5.2, #6, #7]

Time (h) Flow Rate (cfm) Aerosol l Elemental Il Organic I 0.0 1100 0.0 0.0 0.0 0.5 1100 95 95 l 95 720 0.0 0.0 0.0 0.0

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A Entergy Revision I Sheet 25 of 181 Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: FHA Dose Consequences With SGTS and CRHEAFS (Continued) 7.3.3 Case I RADTRAD Input Data (Continued)

3. Atmosphere to CR (inleakage)

a. Transfer mechanism - filter

b. Flow rate and Filter efficiencies (percent) [§5.2, #8]

Time (h) Flow Rate (cfn) Aerosol Elemental I Organic I l 0.0 0 0 0 0 0.5 510 0 0 0 720 0 0 0 0

4. Control Room to Atmosphere (exhaust)

a. Transfer mechanism - filter

b. Flow rate and Filter efficiencies (percent)

Time (h) Flow Rate (cfm) Aerosol Elemental I Ormnic I 0.0 1100 O 0 0 0.5 1610 0 0 0 720 0 0 0 0 Dose Locations Location: EAB LPZ Control Room X/Q (s/m): 0 hr: 5.85E-04 0 hr: 1.9 1E-05 0 hr: 7.32E-07 from MS [§5.2, #II1 2 hrs: 0.0 4 hrs: 0.0 2 hrs: 0.0 Breathing rate (m3/s) 0 hr: 3.5E-04 0 hr: 3.5E-04 0 hr: 3.5E-04

[§5.2, #10] 8 hrs: 1.8E-04 8 hrs: I.SE-04 720 hrs: 0.0 24 hrs: 2.3E-04 24 hrs: 2.3E-04 720 hrs: 0.0 720 hrs: 0.0 0 hr: 1.0 Occupancy '4 hrs: 0.6

[§5.2, #9] 96 hrs: 0.4 720 hrs: 0.0

CALCULATION SHEET CaicNo. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 26 of 181.

Subiect: Radioloaical Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: FHA Dose Consequences With SGTS and CRHEAFS (Continued) 7.3.3 Case 1 RADTRAD Input Data (Continued)

Source Term and Dose Conversion Factors

1. Nuclide inventory - User inventory: file "fha.nif' File consists of PNPS specific activities given in §5.1, #2, with the activities for I-131 and Kr-85 adjusted as indicated in §7.1, #4 and #6 in the RADTRAD file format [16, Table 1.4.3.2-3].

2. Plant power = 14.2, corresponding to damaged fuel rods [§7.;, #7j

3. Decay and Daughter Products - decay/daughter products

4. Iodine Chemical Fractions above water [§5.1, # 14]

Aerosol = 0.0 Elemental I = 0.57 Organic I = 0.43

5. Release Fractions and Timing - user "rft" file - "fha.rft" Duration of release to refuel floor - essentially instantaneous = .OE-04 hr Noble gas fraction = 0.5000E-01 [§7.2, #2]

Iodine fraction = 0.25000E-3 [§7.2, #1]

Others = 0.0 [§7.2, #3]

Delay (hours) = 24

6. Dose Conversion Factors - "fgrl l&12.inp" (default) 7.3.4 Case I RADTRAD Input Files The above data are used to define the RADTRAD plant scenario file for the fuel handling accident -

case 1, "fhal.psf". The scenario file reads the additional files for nuclide inventory, "fha.nif", the release fractions and timing, "fha.rft", and the dose conversion factors, "fgrll&12.inp". A copy of the computer run file is given in Appendix A.

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2 - Entergy Revision I Sheet 27 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST.

7.0 CALCULATION/ANALYSIS (Continued) 7.3 Case 1: FHA Dose Consequences With SGTS and CRHEAFS (Continued) 7.3.5 FHA Case 1 Dose Consequences The estimated dose consequences due to EHA activity release with the SGTS and CRHEAFS in operation are:

Table 7-1 FHA Case 1: With SGTS & With CRHEAFS - Dose Consequences Decay Time I Dose EAB TEDE LPZ TEDE I CR TEDE (hrs) Interval (rem) (rem) (rem) 24 0 - 2 hrs 3.91E-01 0 - 720 hrs _ 1.28E-02 2.20-05 7.4 Case 2 - FHA Dose Consequences With SGTS & NO CRHEAFS 7.4.1 Case 2 Release Scenario For this scenario, activity is released to the environment through the standby gas treatment system (SGTS) filters and out the main stack. The control room high efficiency air filtration system (CRHEAFS) will not be put into.operation during the accident. The normal ventilation system will be in operation throughout the accident. For this case it is assumed that fuel will not be moved for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shut down. Additional delay in moving the fuel will result in lower doses. The activity release scenario for Case 2 is shown on the following sketch. The plant scenario file used with the computer program RADTRAD is detailed in the following sections.

CALCULATION SIIEET CaIc No. PNPS-I-ER! S-XXII.A-2 Entergy Revision I Sheet 28 of 181.

Subject:

Radiological Consegucnces of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATI ON/ANALYSIS (Continued) 7.4 Case 2: FIIA Dose Consequences With SGTS & NO CRIIEAFS (Continued) 7.4.1 Case 2 Release Scenario (Continued)

Figure 7-2 FHIA Case 2 Scenario: SGTS & NO CRHEAFS Unfiltered intake +

infeakage

CALCULATION SHEET Calc No. PNPS-1-ERHS-XXII.A-2. . Entergy Revision I Sheet 29 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.4 Case 2: FHA Dose Consequences With SGTS & NO CRHEAFS (Continued) 7.4.2 Case 2 Dose Analysis

1. Activity decay time = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. SGTS iodine filter efficiency = 99% [§5.2, #2]

3. Release is elevated from the main stack.

4. Comtrm! Room intake/exhaust rate = I 100 cfm [§5.2, #6]

7.4.3 Case 2 RADTRAD Input Data The computer program RADTRAD is used to determine the dose consequences to the control room (CR) and offsite at the exclusion area boundary (EAB) and at the low population zone (LPZ) based on the case 2 scenario. The input is described in the following sections.

Compartments

1. Refuel floor (3-Other)

a. Volume = 700,000 ft3 [§5.2, #1]

b. Source term fraction = 1.0 (all the activity corresponding to the FHA equivalent power)

c. Compartment features - none

2. Atmosphere (Environment)

3. Control Room Volume = 34,280 ft3 [§5.2, #4]

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 --- Entergy Revision I Sheet 30 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.4 Case 2: FHA Dose Consequences With SGTS & NO CRHEAFS (Continued) 7.4.3 Case 2 RADTRAD Input Data (Continued)

Transferpathways

1. Refuel floor to atmosphere

a. Transfer mechanism - filter

b. Flow rate and filter efficiencies (percent) [§7.2, #4] [§7.4.2, #2]

!Elemental I I I

Time (h) Flow Rate (cfm) Ake~tosol Organic I 0.0 80600 99 99 99 2.0 0O 0.0 _ 0.0 0.0

2. Atmosphere to CR (intake)

a. Transfer mechanism - filter

b. Flow rate and filter efficiencies (percent) [§5.2, #6]

Time (b)

I720 Flow Rate (cfm) 1100 0.0 -

Aerosol 0.0 0.0 Elemental I 0.0 0.0 Organic I 0.0 0.0 I

3. Control Room to Atmosphere (exhaust)

a. Transfer mechanism - filter

b. Filter efficiencies (percent)

Time (h) Flow Rate Win) Aerosol Elemental I Organic I 0.0 1100 0.0 0.0 0.0 720 0.0 0.0 0.0 0.0

CALCULATION SHEET Calc No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 31 of 181.

Subiect: Radiolomical Consequences of a Desien Basis Fuel Handling Accident Usine the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.4 Case 2: FHA Dose Consequences With SGTS & NO CRHEAFS (Continued) 7.4.3 Case 2 RADTRAD Input Data (Continued)

Dose Locations Location: EAB LPZ Control Room XIQ (s/rn3 ): 0 hr: 5.85E-04 0 hr: 1.91E-05 0 hr: 7.32E-07 from MS [§5.2, #11] 2 hrs: 0.0 4 hrs: 0.0 2 hrs: 0.0 Breathing rate (m 3 Is 0 hr: 3.5E-04 0 hr: 3.5E-04 0 hr: 3.5E-04

[§5.2, #10] 8 hrs: 1.8E-04 8 hrs: 1.8E-04 720 hrs: 0.0 24 hrs: 2.3E-04 24 hrs: 2.3E-04 720 hrs: 0.0 720 hrs: 0.0 O hr: 1.0 Occupancy 24 hrs: 0.6

[§5.2,$#9] 96 hrs: 0.4 720 hrs: 0.0 Source Term and Dose Conversion Factors

1. Nuclide inventory - User inventory: file "fha.nif' - see §7.3.3

2. Plant power = 14.2, corresponding t6 darnaged fuel rods [§7.1, #7]

3. Decay and Daughter Products - decay/daughter products

4. Iodine Chemical Fractions above water [§5.1, # 14]

Aerosol = 0.0 Elemental I = 0.57 Organic I = 0.43

5. Release Fractions and Timing - user "rft" file - "fha.rft" Duration of release to refuel floor- essentially instantaneous = 1.OE-04 hr Noble gas fraction = 0.5000E-01 [§7.2, #2]

Iodine fraction = 0.25000E-3 [§7.2, #1]

Others = 0.0 [§7.2, #3]

Delay (hours) = 24

6. Dose Conversion Factors - "fgrl 1&12.inp" (default)

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 32 of 181.

Subiect: Radioloeical Consequences of a Design Basis Fuel Handling Accident Usine the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.4 Case 2: FHA Dose Consequences With SGTS & NO CRHEAFS (Continued) 7.4.4 Case 2 RADTRAD Input Files The above data are used to define the RADTRAD plant scenario files for the fuel handling accident

- case 2, "fha2.psf'. The scenario file reads the additional files for nuclide inventory, "fha.nif', the release fractions and timing, "fha.rf", and the dose conversion factors, "fgrll&12.inp". A copy of the input for case 2 and corresponding output file is provided in Appendix B.

7.4.5 FHA Case 2 Dose Consequences The estimated dose consequences due to FHA activity release after 24-hr decay, with the SGTS in operation and CRHEAFS NOT in operation are:

Table 7-2 FHA Case 2: With SGTS & Without CRHEAFS - Dose Consequences Decay Dose EAB TEDE LPZTEDE CRITEDE Time Interval (rem) (rem) (rem) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0 - 2 hrs 3.91E-01 I ___0 I - 720 hrs _ 1.28E-02 2.62E-05

CALCULATION SHEET CaicNo. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 33 of 18 1.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS 7.5.1 Case 3 Release Scenario For this scenario, it is assumed that the SGTS is not available and activity is released to the environment at ground level from the reactor building (RB) vent or from the RB trucklock. The control room high efficiency air filtration system (CRHEAFS) is in operation within 30 minutes of accident initiation. For this case fuel movement will be considered from 1 to 3 days aver the reactor is shut down.

The activity release scenario is shown on the following sketch. The plant scenario file used with the computer program RADTRAD is detailed in the following sections. The only difference beween release from the RB vent and release from the RB trucklock is the X/Q to the control room (CR). Therefore, computer runs are made only for the RB vent case. For the RB trucklock release case, the CR dose will be determined by the ratio of the appropriate X/Qs.

CALCULATION SHEET Calc No. PNPS-1-ERH1S-XX11.A-2 Entergy Revision I Sheet 34 of 181.

Subiect: Radiiological Conse(qucces of al Design Basis F1uel Handling Accident Using Ie AST 7.0 CA LC ULA'T'ION/ANALYSIS (Continued) 7.5 Case 3: FIIA Dose Consequences With NO SGTS anti With CRIIEAFS (Continued) 7.5.1 Case 3 Release Scenario (Continued)

I Figure 7-3 FHA Case 3 Scenario: NO SGTS & CRIIEAFS

CALCULATION SHEET CaIcNo. PNPS-I-ERHS-XXII.A-2. Entergy Revision I Sheet 35 of 181.

Subiect: Radioloeical Consequences of a Desien Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS (Continued) 7.5.2 Case 3 Dose Analysis

1. Activity decay time = 24, 48, 72 hrs.

2. Release is at ground level from the reactor building, unfiltered.

3. Control Room intake/exhaust rate

0 - 0.5 hr: intake rate = 1100 cfm [§5.2. #6], no filters

0.5 hr - 720 hrs: intake rate = 1100 cfm, filter efficiency = 95%[§5.2, #7]

0.5 hr-720 hrs - unfiltered inleakage = .510 cfm [§5.2, #8]

7.5.3 Case 3 RADTRAD Input Data The computer program RADTRAD is used to determine the dose consequences to the control room (CR) and offsite at the exclusion area boundary (EAB) and at the low population zone (LPZ) based on the case 1 scenario. The control room high efficiency air filtration system (CRHEAFS) will not be in operation until 30 minutes after accident initiation. For the first 30 minutes post-accident the normal ventilation system, which provides outside air for the CR, the cable spreading room, the computer room, storage room and corridor will be in operation. The input is described in the following sections.

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 36 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS (Continued) 7.5.3 Case 3 RADTRAD Input Data (Continued)

Compartments

1. Refuel floor (3-Other)

a. Volume = 700,000 ft3 [§5.2, #1]

b. Source term fraction = 1.0 (all the activity corresponding to the FHA equivalent power) c Compartment features - none

2. Atmosphere (Environment)

3. Control Room Volume = 34,280 ft3 [§5.2, #4]

Transferpathways

1. Refuel floor to atmosphere

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) [§7.2, #4l[§7.5.2, #2]

Time (hr) Flow Rate (cfm) Aerosol Elemental I _Org0nic I 0.0 1 80600 0.0 0.0 _0.0I 2.0 . 0 0.0 0.0 0.0 I i

2. Atmosphere to CR - intake

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) 1§5.2, #6, #7]

Time (h) l Flow Rate (cfm) Aerosol Elemental lI Organic I 0.0 1100 0 0 0 0.5 11100 95 95 I 95 720 0 0 0 l 0

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXH.A-2 Entergy Revision I Sheet 37 of 181

Subject:

Radioloical Consequences of a Desisn Basis Fuel Handlin2 Accident Usina the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS (Continued) 7.5.3 Case 3 RADTRAD Input Data (Continued)

3. Atmosphere to CR - inleakage

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) [§5.2, #8]

Time (h) Flow Rate (cfm) Aerosol Elemental I Organic I 0.0 0 0 0 0 0.5 510 0 0 0 720 0 0 0 0

4. Control Room to Atmosphere - exhaust

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent)

Time (h) Flow Rate (cfm)

F Aerosol Elemental I Organic I 0.0 1100 0 0 0 0.5 1610 0. 0 0 720 0 0 0 0 Dose Locations Location: EAB . LPZ Control Room X/Q (s/m3 ): 0 hr: 2.61E-03 Ohr: 2.58E-05 0 hr: 1.85E-03 from RB vent 2 hrs: 0.0 8 hrs: 0.0 2 hrs: 0.0

(§5.2, #11]

Breathing rate (m3/s) 0 hr: 3.5E-04 0 hr: 3.5E-04 0 hr: 3.5E-04

[§5.2, #10] 8 hrs: 1.8E-04 8 hrs: 1.8E-04 720 hrs: 0.0 24 hrs: 2.3E-04 24 hrs: 2.3E-04 720 hrs: 0.0 720 hrs: 0.0 Occupancy Ocup2 #91c 0 hr:hrs: 0.6 24 1.0

[§5.2. t1 96 hrs: 0.4 720 hrs: 0.0

CALCULATION SHEET Caic No. PNPS-I-ERHS-XII.A-2. Entergy Revision I Sheet 38 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS (Continued) 7.5.3 Case 3 RADTRAD Input Data (Continued)

Source Term and Dose Conversion Factors

1. Nuclide inventory - User inventory: file "fha.nif" - see §7.3.3

2. Plant power = 14.2, corresponding to damaged fuel rods [§7.1, #7]

3. Decay and Daughter Products - decay/daughter products

4. Iodine Chemical Fractions above water [§5.1, # 14]

Aerosol = 0.0 Elemental I = 0.57 Organic I = 0.43

5. Release Fractions and Timing - user "rft" file - "fha.rft" Duration of release to refuel floor - essentially instantaneous = 1.OE-04 hr Noble gas fraction = 0.5000E-01 [§7.2, #2]

Iodine fraction = 0.2500E-3 [§7.2, #1]

Others = 0.0 [§7.2, #3]

Delay (hours) = 24, 48, 72

6. Dose Conversion Factors - "fgrl 1&12.inp" (default) 7.5.4 Case 3 RADTRAD Input Files The above data are used to define the RADTRAD plant scenario files for the fuel handling accident

- case3, "fha3a.psf', "fha3b.psf", and "fha3c.psf". The scenario file reads the additional files for nuclide inventory, "fha.nif', the release fractions and timing, "fha.rft", and the dose conversion factors, "fgrl 1&12.inp". Copies of the files for case 3 are given in Appendix C1 - Appendix C3.

CALCULATION SHEET Caic No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 39 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.5 Case 3: FHA Dose Consequences With NO SGTS and With CRHEAFS (Continued) 7.5.5 FHA Case 3 Dose Consequences The estimated dose consequences due to FHA activity release without the SGTS and with CRHEAFS in operation are given below. The CR TEDE values from Appendix C are for release from the RB vent. The corresponding values for release from the RB trucklock are obtain by multiplying the run value by RB trucklock to CR X/Q 9.87 x 104 , 3 -=0.534 RB venttoCR X/Q 1.85xi(f3 in3 Table 7-3 FHA Case 3: Without SGTS & With CRHEAFS- Dose Consequences Decay Dose EAB TEDE LPZ TEDE CRTEDE Time Interval (rem) (rem) (rem)

. . RB vent RB trucklock 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0 - 2 hrs 6.54 0-720 hrs 6.46E-02 2.72 1.45 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0 - 2 hrs 4.93

_ - 720 hrs 4.87E-02 2.35 1.25 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 0 - 2 hrs 4.23 0 - 720 hrs 4.18E-02 2.10 1.12

CALCULATION SHEET CaIc No. PNPS-1-ERHS-XX11.A-2 _ Entergy Revision I Sheet 40 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FIA Dose Consequences With NO SGTS and NO CRHEAFS 7.6.1 Case 4 Release Scenario For this scenario, it is assumed that neither the SGTS nor the CRHEAFS is available. Activity is released to the environment directly from the reactor-building vent or from the RB trucklock, at ground level. Fuel movement from 1 day to 5 days after reactor shut down will be considered. The activity release scenario is shown on the following sketch. The plant scenario file used with the computer program RADTRAD is detailed in the following sections. The only difference between release from the RB trucklock and release from the RB vent is the X/Q to the Control Room. Therefore, computer runs are made only for the RB vent case. For the RB trucklock release case, the CR dose is determined from the RB vent release case by the ratio of the appropriate x/Qs.

CALCULATION SI 1EET Caic No. 13NPS-I-ERIIS.XXII.A-2 Entergy Revision I Sheet 41 or 181 .

Subject:

Radiological ConseC(Itlefces of a Design Basis Fuel Handling Accident Using tie As'r 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FIIA Dose Consequences With NO SGTS and NO CRHEAFS (Continued) 7.6.1 Case 4 Release Scenario (Contfintedl)

Figure 7-4 FIHA Case 4 Scenario: NO SGTS & NO CRHEAFS

CALCULATION SHEET CaIc No. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 42 of 181.

Subiect: Radiological Conseauences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FHA Dose Consequences With NO SGTS and NO CRHEAFS (Continued) 7.6.2 Case 4 Dose Analysis

1. Activity decay time = 24, 48, 72, 96, 120 hrs.

2. Release is unfiltered at ground level from the reactor building.

3. Control Room intake/exhaust rate = 1100 cfm [§5.2, #6]:

7.6.3 Case 4 RADTRAD Input Data The computer program RADTRAD is used to determine the dose consequences to the control room (CR) and offsite at the exclusion area boundary (EAB) and at the.low population zone (LPZ) based on the case 4 scenario. The normal ventilation system, which provides unfiltered outside air for the CR, the cable spreading room, the computer room, storage room, and corridor, will be in operation for the entire 30-day period. The input is described in the following sections.

Compartments

1. Refuel floor (3-Other)

a. Volume = 700,000 ft3 [§5.2, #1]

b. Source term fraction = 1.0 (all the activity corresponding to the FHA equivalent power)

c. Compartment features - none

2. Atmosphere (Environment)

3. Control Room = 34,280 ft3 [§5.2, #4]

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 43 of 181 Subiect: Radioloeical Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FHA Dose Consequences With NO SGTS and NO CRHEAFS (Continued) 7.6.3 Case 4 RADTRAD Input Data (Continued)

Transfer pathways

1. Refuel floor to atmosphere

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) [§7.2, #4][§7.6.2, #2]

Time (h) Flow Rate (cfm) Aeroscd l EMmental I Organic I 0.0 80600 0.0 0.0 0.0 2.0 0 0.0 0.0 0.0

2. Atmosphere to CR - intake

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) [§7.6.2, #3]

Time (h) l Flow Rate (cfm) Aerosol Elemental I Organic I 0.0 1100 0.0 0.0 0.0 720 0.0 0.0 0.0 0.0

3. Control Room to Atmosphere - exhaust

a. Transfer mechanism - filter

b. Flow rates and filter efficiencies (percent) l Time (h) . Flow Rate (cfm) ' Aerosol l Elemental I . Organic I 0.0 1100 0.0 0.0 0.0 720 10.0 I 0.0 0.0

CALCULATION SHEET CaIcNo. PNPS-1-ERHS-XXII.A-2 Entergy Revision I Sheet 44 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FHA Dose Consequences With NO SGTS and NO CRHEAFS (Continued) 7.6.3 Case 4 RADTRAD Input Data (Continued)

Dose Locations Location: EAB LPZ Control Room XIQ (s/rm): 0 hr: 2.61E-03 0 hr: 2.58E-05 0 hi: 1.85E-03 from RB vent 2 hrs: 0.0 8 hrs: 0.0 2 his: 0.0

[§5.2,#111 Breathing rate (m3/s) 0 hr: 3.5E-04 0 hr: 3.5E-04 0 hr: 3.5E-04

[§5.2, #10] 8 his: 1.8E-04 8 hrs: 1.8E-04 720 hrs: 0.0 24 his: 2.3E-04 24 his: 2.3E-04 720 his: 0.0 720 his: 0.0

.n O hr:

Occupancy 24 hrs: 1.0 0.6

[§5.2,#9] 96 hrs: 0.4 720 hrs: 0.0 Source Term and Dose Conversion Factors

1. Nuclide inventory - User inventory: file "fha.nif' - [§7.3.3]

2. Plant power= 14.2, corresponding to damaged fuel rods [§7.1, #7]

3. Decay and Daughter Products - decay/daughter products

4. Iodine Chemical Fractions above water [§5.1, # 14]

Aerosol = 0.0 Elemental I = 0.57.

Organic I = 0.43

5. Release Fractions and Timing - user "rft" file - "fha.rft" Duration of release to refuel floor - essentially instantaneous = 1.OE-04 hr Noble gas fraction = 0.5000E-01 [§7.2, #2]

Iodine fraction = 0.2500E-3 [§7.2, #1]

Others = 0.0 [§7.2, #3]

Delay (hours) = 24, 48, 72, 96, 120

6. Dose Conversion Factors - "fgrl I &12.inp" (default)

CALCULATION SHEET CaIc No. PNPS-I-ERHS-XXII.A-2. Entergy Revision I Sheet 45 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 7.0 CALCULATION/ANALYSIS (Continued) 7.6 Case 4: FHA Dose Consequences With NO SGTS and NO CRIEAFS (Continued) 7.6.4 Case 4 RADTRAD Input Files The above data are used to define the RADTRAD plant scenario files for the fuel handling accident

- case 4, "fha4a.psf' through "fha4e.psf". The scenario file reads the additional files for nuclide inventory, "fha.nif", the release fractions and timing, "fha.rft", and the dose conversion factors, "fgr 1&12.inp".

Copies of the computer files for case 4 are given in Appendix DI - D5.

7.6.5 FHA Case 4 Dose Consequences The estimated dose consequences due to FHA activity release with NO SGTS and NO CRHEAFS in operation are given below. The CR TEDE values from Appendix D are for release from the RB vent. The corresponding values for release from the RB trucklock are obtain by multiplying the run value by 0.534 [§7.5.5]

Table 7-4 FHA Case 4: Without SGTS & Without CRHEAFS - Dose Consequences Decay Dose EAB TEDE LPZ TEDE CR TEDE Time

Interval ** (rem) (rem) (rem)

RB vent RB Trucklock 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0 - 2 hrs 6.54 0 - 720 hrs 6.46E-02 3.29 1.76 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0 - 2 hrs 4.93 0 - 720 hrs 4.87E-02 2.85 1.52 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 0 2 hrs 4.23 0-720 hrs 4.18E-02 2.54 1.36 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 0 - 2 hrs 3.78 0 - 720 hrs 3.73E-02 2.30 1.23 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months 0 - 2 hrs 3.42 0-720 hrs 3.38E-02 2.10 1.12

Times are relative to the time after reactor shutdown when fuel movement commences.

- Intervals are post-accident

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 46 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handlinf Accident Using the AST 8.0 RESULTS The radiological consequences of the fuel handling accident (FHA) assuming GE14 10 x 10 fuel assemblies are summarized below for the four combinations of the SGTS and CRHEAFS:

Case 1: SGTS is in operation, CRHEAFS is in operation Case 2: SGTS is in operation, CRBEAFS is not in operation Case 3: SGTS is not in operation, CRHEAFS is in operation, Case 4: SGTS is not in operation, CRHEAFS is not in operation The FHA total effective dose equivalentsmfor each case are given below. -

Table 8-1 FuelHandlingAccidentTEDE Operation Dose (rem)

Release Point RB RB Vent

trucklock*

Case Decay SGTS CRHEAFS EAB LPZ CR CR 1 24 hrs YES YES 3.9E-01 1.3E-02 2.2E-05 2.2E-05 2 24 hrs YES NO 3.9E-01 1.3E-02 2.6E-05 2.6E-05 3a 24 hrs NO YES 6.5 6.5E-02 2.7 1.5 3b 48 hrs NO YES 4.9 4.9E-02 2.4 1.3 3c 72 hrs NO YES 4.2 4.2E-02 2.1 1.1 4a 24 hrs NO NO 6.5 6.5E-02 3.3 1.8 4b 48 hrs NO NO 4.9 4.9E-02 2.9 1.5 4c 72 hrs NO NO 4.2 4.2E-02 2.5 1.4 4d 96 hrs NO NO 3.8 3.7E-02 2.3 1.2 4e 120 hrs NO NO 3.4 3.4E-02 2.1 1.1 Regulatory Limit 6.3 6.3 5

For Cases 3 and 4, only.

CALCULATION SHEET Calc No. PNPS-I-ERHS-XXII.A-2 Entergy Revision I Sheet 47 of 181.

Subiect: Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST 8.0 RESULTS (CONTINUED)

It is seen in Table 8-1 that the estimated radiological consequences due to a design basis fuel handling accident at PNPS are below the regulatory limits in all cases if fuel is moved after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months post-reactor shutdown, i. e., if fuel is moved 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after reactor shutdown, neither SGTS nor CRIEAFS is required to be in operation. If SGTS is in operation, fuel movement is acceptable 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown with or without CRHEAFS in operation (case 1 and case 2). Without SGTS in operation, fuel movement prior to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after reactor shutdown is not acceptable (case 3a and case 4a).

CALCULATION SHEET Caic No. PNPS-1-ERHS-XXII.A Entergy Revision I Sheet 48 of 181.

Subject:

Radiological Consequences of a Design Basis Fuel Handling Accident Using the AST APPENDICES - RADTRAD Computer Program Files Pae Appendix A - FHA Case 1 Computer Run ..............................................................................

AG~

Appendix B - FHA Case 2 Computer Run ............................................................................ (A...............6 Appendix C1- FHA Case 3a Computer R ........................................................................... ...............

77 Appendix C2 - FHA Case 3b Computer R .......................;................................................. -.................

92 Appendix C3 - FHA Case 3c Computer Ri ......................................................................... ...............

107 Appendix DI - FHA Case 4a Computer Ri .............................. . .......................... 122 Appendix D2 - FHA Case 4b Computer R ......................................................................... ..................

134 Appendix D3 - EHA Case 4c Computer Ri ....................................................................... ...146 Appendix D4 - FHA Case 4d Computer R ..... ...158 I

Appendix D5 - EHA Case 4e Computer Ri ....................................................................... ...170

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