NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 3

Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 3

ML16071A140
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	03/03/2016
From:	Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16071A108	List: ML16071A110 ML16071A126 ML16071A132 ML16071A136 ML16071A137 ML16071A140 ML16071A144 ML16071A146 ML16071A147 ML16071A149 ML16071A150 ML16071A152 ML16071A153 ML16071A158 ML16071A160 ML16071A166 ML16071A170 ML16071A174 ML16071A176 ML16071A180 ML16071A181 ML16071A184 ML16071A185 ML16071A187 ML16071A188 ML16071A193 ML16071A195 ML16071A198 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 1 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 2 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 3 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 4 of 4 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 3 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 3 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 3 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 6 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 7 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 8 of 8 NL-15-1898, Enclosure 5: Vogtle Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 4 ... further results
References
NL-15-1898
Download: ML16071A140 (247)

Text

Edwin I. Hatch Nuclear Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 3 Hatch EAL Calculations COMPANYJM Southern Nuclear Design Calculation

[Calculation Number: SMNH-;13-021 Plant:Hatch Nuclear Plant Unit:O-I 0 [ 2 E[] I & 2 fDiscipline:Mechanical Title: I

Subject:

Emergency NEi 99-01 Rev 6,EAL Calculations

.,.Action Level Setpoints"PurpoSe I Objective::

Document Emergency Action Level Values to support conversion to NEI 99-01 Rev 6 System or Equipment"Tag Number..:

'N/A .. .i..Contents Topic Page Attachments*

1. of (Computer Printouts, Technical Papers, Pages Sketches,, Correspondence)

Purpose 1 A -SNC EP Concurrence " ..1 Criteria 1 B -Reserved 0 conclusions 3 C -FuelClad Loss 4 .A Threshold calculation 19 Design Inputs ' 5 D -Fuel Clad Potential Loss 5.A Threshold 1.Assumptions*

10 11___References 14 !E -RCS Loss 4.A Threshold Calculation 16 Method of Solutions

...18i F -'Rc's Loss 5.A Thresh'old Calculation " .6 Body of Calculation

... ;: " 123 IG -Containment Potential Loss ,4A"Threshold 12 Calculation

"____ ._________

___H -Steam Table Validation

-... ' 'I- RSI Threshold Calculation 25 Total # of Pages including J8 -K Threshold Calculatio~n 4'1 cover sheet & Attachments:

18 K-G hrsodaclain1 L- E-HU.! Threshold Calculation

... 16 M -RA1 Threshold Calculation

.18 Nuclear Quality Level.0] Safety-Related IX] Safety Significant

.. I"1 Non- Safety --Significant Version Record _____.... ..nator .. Revioewer

'Approv'al I "

2 No. Description -__________

J htjSlrx D.S. 0. W. Wu A.T. Vieirs D.L. Lamlbert 1 Issued McCutcheon Sign. on file Sign. on file Sign. on file 2 Complete Revision 9I ooonakisn

'7 a oile KeithoDudy Notes: .. .1. Additional work and changes to this calculation are required.

The calculation is "APPROVED PENDING" NRC review and acceptance of the Emergency Plan (EP) submittal.

NMP-ES-039 NMP-ES-039-001 Purpose: The purpose of this calculation is to provide values/data/curves and bases for use in development of the Hatch Nuclear Plant Emergency Action Levels (EALs) using Nuclear Energy Institute (NEI) 99-01 Rev 6 guidelines.

This combined calculation includes all unique calculations required to support emergency action level thresholds as well as references to calculations used to create thresholds, but serve purposes beyond emergency action levels.The contents of this calculation are primarily taken from the calculation which supported the previous emergency action level scheme. The work performed in calculation SMNH-05-009 V2 is directly transposed into this document and edited to account for the differences between NEI 99-01 Rev 4and 6.The calculation Attachments L and M of this document contain calculations supporting new emergency action level thresholds and represent the only portion of this document which is not directly transposed from SMNH-05-009.

The unique work in this document includes those attachments and their associated content in the body of the calculation.

Several EAL thresholds were removed as they do not support the NEI 99-01 Rev 6 scheme. Thresholds removed include specific water elevations which previously supported l~s RU2, RA1, RA2, BFL2, and RCS Barrier. The transposed material in this calculation has been further altered to reflect the new language and organization of NEI 99-01 Rev 6. These changes in language and organization are administrative in nature and have no impact on the calculation output.Criteria: The calculation performed will support the development of guidelines for NEI 99-01 initiating conditions (ICs) RU1, RU2, RA1, RA2, RS1, RGI, CAl, CS1, CG1, E-HU1, Fuel Clad Barrier, RCS Barrier, Containment Barrier, and SU3.1. Declaration of an emergency, when such a declaration is not required, involves risk to the public as does the failure to make such a declaration, should one be warranted.

Therefore, this calculation shall develop a "best estimate" value for the dose rates or curie concentrations sensed at the monitors chosen for the Emergency Action Level (EAL) set points. When judgments are necessary, these judgments shall be as close to anticipated conditions as possible.2. If a particular monitor is to be used for an EAL, then the dose rate or curie concentration set point developed for that specific monitor shall be within the range of the monitor, or the monitor shall not be cited as applicable for the EAL.3. In accordance with the guidance of Regulatory Guide 1 .97, Revision 2, post-accident radiation monitors must read within a factor of 2 of actual radiation conditions.

Therefore, changes in the set points of this revision that are within a factor of 2 of the previous revision's set point for the same EAL do not invalidate the previous set point.It is up to the ultimate user of these calculations to determine if change to the EAL set point guidance document(s) is warranted.

4. Methods and Assumptions shall comply with the guidance of NEI 99-01 Revision 6.Note: NEI 99-01 Rev. 6 states that the "A" Recognition Category designation may be changed to "R" provided the change is carried through for all of the associated IC identifiers.

As such, the Hatch Nuclear Plant Emergency Action Levels use the Recognition Category designation of "R" for the Abnormal Radiation Recognition Category.

==

Conclusions:==

This calculation contains thresholds for the Hatch Nuclear Plant Emergency Action Levels. Site specific information used to develop the scheme is included in the Body of Calculation section beginning on page 23. In this document, ten thresholds are calculated.

The results of these calculations are as follows.Initiating Condition RA1 Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6E-02 pCi/cc 8.1E+01 pCi/cc Initiating Condition RS1 Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6E-01 pCi/cc 8.1E+02 pCi/cc Initiating Condition RGI Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6 E+OO pCi/cc 8.1E+03 pCi/cc Initiating Condition CG1 The supporting calculation for this threshold concludes that there are no monitors able to provide on-scale indications of core uncovery.Initiating Condition E-HU1 Greater than any of the following on-contact radiation readings serve as thresholds for EAL 1.

Southern Nuclear 0perating Cornpany SOU1HURN Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-04 COMPANY Unit: l&2SH T4 Table El Location of Dose Rate Total Dose Rate (Neutron + Gamma mR/hr)HI-TRAC 125 Side -Mid- height 450 Top 110 HI-STAR 100 or HI-STORM 100 Side -60 inches below mid- 80 height Side -Mid- height 80 Side -60 inches above mid- 30 height Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 Fuel Clad Barrier Loss Threshold 4.A A Drywell Wide Range Radiation Monitor (DWRRM) reading greater than 1,400 R/hr.Fuel Clad Barrier Loss Threshold 5.A The supporting calculation for this threshold concludes that the applicable monitors will be off scale.RCS Barrier Loss Threshold 4.A A DWRRM reading greater than 40 R/hr.RCS Barrier Loss Threshold 5.A The reading on drywell fission product monitor D11K630 of 1E+06 cpm will indicate a potential loss of the RCS barrier. Per SX18062 page 34, the monitor K630 range is 10 to 1E+06 cpm. A reading of 5.OE+05 cpm is an appropriate threshold to ensure an accurate reading is possible.Primary Containment Barrier Potential Loss Threshold 4.A A DWRRM reading greater than 26,000 R/hr.

Southern Nuclear 0perating Cornpany ISOUTHERNEI Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY' Unit: 1&2 SHEET 5 Design Inputs (additional design inputs and references are listed in the attachments as needed): Attachment C: 1. The core inventory source terms are listed in Enclosure 1 of NL-06-1637 page 16.Core Inventory Isoope Curies/MWt Kr-83m 3.30E+03 Kr-85 3.78E+02 Kr-85m 6.92E+03 Kr-87 1.32E+04 Kr-88 1.86E+04 Kr-89 2.26E+04 Xe-131m 3.03E+02 Xe-133 5.27E+04 Xe-133m 1.58E+03 Xe-135 1.89E+04 Xe-135m 1.09E+04 Xe-137 4.81E+04 Xe-138 4.52E+04 1-131 2.72E+i04 1-132 3.93E+'04 1-133 5.52E+04 1-134 6.05E+04 I- 135 5.16E+04 Per NUREG-1301 pg. 6, DE1-131 only considers the following iodine isotopes (1-131, 1-132, 1-133, 1-134, 1-135). Therefore other iodine isotopes have been excluded from this calculation

2. Dose Equivalent 1-131 is defined as that concentration of 1-131 (pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, 1-135 (NUREG-1 301 page 6).3. Core Power for radiological evaluations of 2818 MWt is provided in NL-06-1 637 page 14 Enclosure 1.4. RCS inventory of 9965 ft^3 is provided in NL-06-1 637 Enclosure 1 page 47 Table 34.5. The release fraction into the containment for noble gases and halogens (i.e. lodines) is 0.05 per RG-1.183 page 13.6. The dry well free volume (UI: 146010 ftA3 U2: 146266 ftA3) is listed in NL-06-1637 Enclosure 1 page 47 table 34.7. The suppression pool free volume (UI: 112900 ftA3 U2: 109800 ftA3) is listed in NL-06-1637 Enclosure 1 page 47 table 34.

8. UI: The Main feed water temperature is 392.4 °F and core recirculation temperature is 532.0 0 F. Per GE-NE-0000-0003-0634-01 pg. 15.9. U2: The Main feed water temperature is 425.7 °F and the core recirculation temperature is 535.0 0 F. Per GE-NE-0000-0003-0634-01 pg. 16.10. According to PDMS the Drywell Wide Range detector is MPL 1 D11 N003A&B and 2D11IN003A&B.

The detectors are connected to D11 K621 Recorders as described in the following table: MPLDraingInstrument Tag Range'MLSheet No R/hr Model No 1D1IN003A A16481 DuIB QD11-RE-N003A 1EO- 1E7 877-1 1D11N003B A16481 D11C QD11-RE-N003B 1E0 -1E7 877-1 2D11N003A A26481 D11A Q2D11-RE-NO03A lE0- 1E7 877-1 2D11NOO3B A26481 D11B O.2Dl1-RE-NOO3B lEO -1E7 877-1 Attachment D I11. Attachment D Factor Allocation 0.225 and Factor Safety 0.5 per 64CI-OCB-004-1/2 Page 56.12.0Off-Gas Post Treatment Radiation Monitor D11 K61 5 efficiency Factor 3.8E5 cps/pCi/cc per e-mail attached to attachment D 13. The High Purity Waste Stream leakage inside the Drywell is obtained from SMNH-93-029 page 6 and is 3900 gal/day.14. The Low Purity Waste Stream leakage inside the Drywell is obtained from SMNH-93-029 page 6 and is 700 gal/day.15. Fission Product Monitor D11K630 for Xe133 = 2.7E7 cpm/IpCi/mL and Kr85 = 2.29E8 cpm/pCi/mL.

The response factor was obtained from SX1 8062 pg. 32 and SX27520 pg.32 for equipment tag number P010.Attachment E 16. The radiation field from RPV at elevation 153.6 for Unit 1 and Unit 2 is listed below (BH2-M-V999-0047 Table 2 and 3).

Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-FO13A 6.44 2B21-AOV-FO13A 2.46 1B21-AOV-FO13B 4.41 2B21-AOV-F013B 9.96 1B21-AOV-FO13C 12.3 2B21-AOV-FO13C 9.96 1B21-AOV-F013D 13 2B21-AOV-F013D 2.46 1B21-AOV-FO13E 12.3 2B21-AOV-F013E 2.46 1B21-AOV-F013F 12.3 2B21-AOV-F013F 9.96 1B21-AOV-F013G 12.3 2B21-AOV-FO13G 9.96 1B21-AOV-FO13H 4.41 2B21-AOV-FO13H 2.46 1B21-AOV-F013i 4.41 2B21-AOV-FO13K 9.96 1B21-AOV-FO13K 12.3 2B21-AOV-FO13L 9.96 1B21-AOV-FO13L 12.3 2B21-AOV-F013M 9.96 Average = 9.7 7.2 17. The radiation field from Main Steam Lines (BH2-M-V999-0047 Table 5 and 6).Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-FO13A 17.57 2B21-AOV-FO13A 17.39 1B21-AOV-FO13B 17.28 2B21-AOV-FO13B 17.28 1B21-AOV-F013C 17.08 2B21-AOV-FO13C 17.08 1B21-AOV-F013D 17.01 2B21-AOV-F013D 17.39 1B21-AOV-F013E 17.16 2B21-AOV-F013E 17.28 1B21-AOV-F013F 16.13 2B21-AOV-FO13F 17.08 1B21-AOV-F013G 16.13 2B21-AOV-F013G 15.43 1B21-AOV-F013H 16.62 2B21-AOV-F013H 17.28 1B21-AOV-F013J 16.33 2B21-AOV-FO13K 15.43 1B21-AOV-F013K 16 2B21-AOV-F013L 14.78 1B21-AOV-F013L 13.4 2B21-AOV-F013M 14.78 Average = 16.4 16.5 18. The radiation field from Recirculation Lines (BH2-M-V999-0047 Table 9 and 10).'S Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-F013A 7.5 2B21-AOV-F013A 11.97 1B21-AOV-F013B 8.23 2B21-AOV-F013B 6.23 1B21-AOV-F013C 6.19 2B21-AOV-F013C 6.85 1B21-AOV-F013D 5.2 2B21-AOV-F013D 10.08 1B21-AOV-F013E 5.89 2B21-AOV-F013E 15.22 1B21-AOV-F013F 5.15 2B21-AOV-F013F 6.19 1B21-AOV-F013G 5.89 2B21-AOV-F013G 6.85 1B21-AOV-FO13H 7.5 2B21-AOV-F013H 15.96 1B21-AOV-F013J 7.5 2B21-AOV-F013K 6.85 1B21-AOV-F013K 6.22 2B21-AOV-F013L 7.8 1B21-AOV-FO13L 6.22 2B21-AOV-F013M 7.8 Average = 6.5 _______ 9.3 Attachment F 19.The Normal RCS Isotopes concentration is taken from FSAR Table 11 .1-1, and are listed in the attachment F.20. The core flow (UI: 78.5E6 Ibm/hr U2: 77E6 lbm/hr) can be found in GE-NE-0000-0003-0634-01 Figures Ia & lb.21. Per SX1 8062 page 34 the monitor K630 range is 10 to 1 0A6 cpm.Attachment G3 22. DWRRM is D11K621 NMP-EP-110-GL02 pg. 66.23. The source terms for noble gases and iodines are provided in NL-06-1 637 and are shown below. We will only consider the isotopes that have release fraction above 0 as described in RG 1.183 Table I and Table 5.24. The 20% is the core fraction gap release as described in NEI 99-01.Attachment H 25. Radiation from a HOLTEC Overpack from table I and 2 of S55932.26. The attenuation coefficient for 6 MeV gamma was used. The lowest attenuation was used thus giving us most conservative (lowest) shielding factor. Attenuation

= 0.234 cm^-l.27. MPC lid thickness 9.5 inches per $55932 pg. 7 bullet 9.Attachment I

28. The Core Release Fractions are taken from RG 1.183 Table 1. This fraction, describes release of isotopes to RCS water. Iodines 0.3 and Noble Gases 1.0.29. The Partition Coefficient is the ratio of the concentration of a nuclide in the gas phase to the concentration of that nuclide in the liquid phase when the liquid and gas are at equilibrium.

It is assumed that 100% Noble gases are released into the steam.According to NUREG-0016 Table 2-7 page 2-13 the Iodines partition coefficient is 0.004.30. X_Q = 8.37E-6 sec/m^3 ground release per Table 3-4 of HNP ODCM page 3-17 31. Reactor building ventilation flow rate 1 .42E8 mL/sec per Table 3-4 of HNP ODCM page 3-17 32. The Dose Conversion Factor (DCF) for Effective Dose Equivalent (EDE) was taken from FGR12 "Effective Column" of Table II1.1.33. The Dose Conversion Factor (DCF) for Committed Effective Dose Equivalent (CEDE)was obtained from FGR11 Table 2.1 column labeled "Effective".

34.The exposure time is provided in NEI 99-01, T~lhr.35.The breathing rate for persons offsite is listed in section 4.1.3 page 1.183-16 of RG-1.183. BR=3.5E-4 mA3/sec 36. Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE). However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent for external exposure, if the body is irradiated uniformly.

37. The Thyroid Committed Dose Equivalent (CDE) from inhalation was obtained from FGR 11 Table 2.1 column labeled "Thyroid".

38. X_Q = 4.10E-8sec/m^3 elevated release per Table 3-4 of H NP ODCM page 3-17.39. Main stack flow 9.44E6 mL/sec per Table 3-4 of HNP ODCM page 3-17.40. Recombiner building vent flow 2.36E5mL/sec per Table 3-4 of HNP ODCM page 3-17.Attachment J 41. Radiation field above SFP when water is at TOAF is 2.68E5 R/Hr per BH2-M-V999-0048 pdf page 39 column 3.42. The material of RPV is steel per FSAR U2 Table 5.2-6.43.The RPV inside diameter is 218 inches per FSAR U2 Table 5.4-1 and $15213.44. RPV vessel thickness is 5.38 inches per FSAR U2 Table 5.2-10.Attachment K 45. Roof elevation of 281 ft 9 in DWG H25963.

Assumptions (additional assumptions are listed in attachments as necessary):

Attachment C 1. The higher temperature of RCS will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-Q000-0003-0634-01 page 19 the reactor pressure is 1060 psia. According to http://www .spiraxsarco.com/resources/steam-tables .asp, the specific volume for 1060 psia and 480.35 °F conditions is 0.01 991 61 ft^3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to attachment C.2. According to http://www.spiraxsarco.com/resources/steam-tables.asp, the specific volume for 44.7 psia and 343 0 F conditions is 10.4690 ftA3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to Attachment C. The drywell pressure, and temperature were obtained from NL-06-1 637 Enclosure 1 table 11 and table 12.3. The Drywell Wide Range detectors are 1 D11N003A&B and 2D11IN003A&B.

Per H16241 and H26417 the instrument location is at elevation of 156 ft and approximately 27 ft from the center of RPV. As the result the geometry of detector in this volume will be modeled as a cylinder with of 41.5 feet high. This number corresponds to the elevation above drywell floor. The radius of the cylinder was adjusted to the drywell free volume that is in the line of site of the detector.

The geometry factor will be evaluated for the point P1. The uncollided fluxes at interior and exterior points of a non-absorbing cylindrical volume source are provided in Engineering Compendium on radiation shielding pages 381 and 382.T. .........

-" Fig'. 6.4.-?. Geometry3 for non-absorbing cylindrical, volume source; interior, surface and exterior points.At P 1 : 4. The mean free path was calculated for the containment air (Figure 1). It can be seen that the mean free path is over 650 feet, which is much larger than the reactor building of 71.75 ft+71 .75 ft = 143.5 ft (H25941).

Therefore it was concluded that the Southern Nuclear 0peratinl Cornpany OUHr 1 Plant: HNP SMNH-1 3-021 nt & Title: NEI 99-01 Rev 6 EAL Calculations SET1 containment atmosphere does not provide any appreciable gamma shielding and that the non-absorbing cylindrical model should be used.Figure 1: Mean Free Path for Drywell Steam..... .pa/_!=1.

Mass Absorption Coefficient (cm^2/g) --..... =Absorption Coefficient_(1l/cm_)

... .....;L= iMean Free Path (cm) !........... I ...' ........L -.....I .... .....Steam @ 44.7 psia & 343degF L---... .... ...... .....-- -...... ..... ... ... ....... ... ...- -... .-!.... L Calculated in AttachmentC p.= 11.53E-03 IEnergy J.aPLia (MeV) g)^2 (11cm) (cm) (ft)0.5 0.0330 5.05E-05 1.98E+04 6.50E+02 1.0 0.0311 4.76E-05 2.10E+04 6,89E+02 1.5 0.0285 4.36E-05 2.29E+04 7.52E+02 2.0 0.0264 4.04E-05 2.48E+04 8,12E+02 3.0 0.0233 3.56E-05 2.81E+04 9.20E+02 4.0 0.0213 3.26E-05 3.07E+04 1.01E+03 glcm^~3 to--.. .... ... .. .......- ..... ... ... .. .... ..... ..... .. ... .... .. .. ..... ......$

Reference:

Table 11.5, page 649, Lamarsh, "Introduction Nuclear Engineering," 2nd edition, 1983 -, I Attachment D 5. Since the D11 P010 gas detector is G-M counter ($30523 pdf page 76 section 1-4) the Xe1 33 response factor will applied to all isotopes of Xenon, while the Kr85 response factor will be applied to all isotopes of Krypton. For G-M detectors all gammas above detector energy threshold would initiate charged particles inside the detector.Attachment F 6. Noble gases are not typically retained in the RCS water, but are continuously released via offgas system. The source terms for noble gases are taken from Table 11.1-1. It is assumed that 100% of noble gases will leave the solution.Attachment I

Southern Nuclear 0 eratin9 Cornpany SOUTHERN Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET13-2 COMPANY Unit: 1&2 SET1 7. The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used in HNP FSAR Chapter 15 evaluations:

Iodine's and noble gases Table 15.3-4. It is assumed that only iodine's and noble gases needs to be considered, because particulates will be retained in the primary containment water.8. The RCS will have some equilibrium noble gases and some iodine in the RCS water from normally operating reactor. According to HNP FSAR U2 Table 11.1-2 the levels of iodines is on the order of 1 E-1 pCi/g which is negligible to calculated above order of magnitude of 1 E5 pCi/g. The reason these isotopes are in low concentrations because they are continually removed in the power plant steam condensers.

Thus initial equilibrium noble gases and iodines will be neglected.

Attachment J 9. The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top of Active Fuel. The calculation pdf page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Since the RPV core is assumed to be a columnized source, the same radiation level is assumed to be at the edge of the RPV. This is conservative and according to the BH2-M-V999-0048 pdf page 39 column 2 on the edge of the SFP the radiation level is 2.29E5 R/hr. As can be seen the radiation level does not very much from the middle of the core to the edge of the core.10. The page B-9 of BH2-M-V999-0048 provides the gamma source strength broken down by energy groups. It can be seen that the maximum gamma strength is spread between 0.4 to 1.8 MeV, therefore this calculation will use linear attenuations that are associated with 1 MeV gamma rays.Attachment K 11. The reflected dose rate at the operating deck area radiation monitors will be calculated using the methods of Davission's "Gamma Ray Dose Albedos" (copy in Attachment K).The calculation will be based on an iron reflector at the top of the secondary containment, with a diameter equal to the drywell, and a distance R from the reflector to the radiation monitor equal to the hypotenuse of the triangle formed by the difference in elevations of the reflector and the monitor and the distance from drywell center to the approximate detector placement.

The iron reflector is selected because the containment roof has metal roof deck. The reflected dose rate is proportional to the area of the reflector.

Assuming the reactor vessel functions as a collimator with reduced RCS inventory will reduce the reflected area. This in turn reduces the dose rate at the radiation monitor and, therefore, the EAL threshold for reduced RCS inventory.

f 4 Southern Nuclear 0 eratinl Cornpany SOTEN4 Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-1 3-021 COMPANY Unit: 1&2 SHEET 14

References:

1. NEI 99-01 "Methodology for Development of Emergency Action Levels" Rev 6.2. H16145 V18.0 "HNP UI Nuclear Boiler System P&ID Sheet 3" 3. H261 89 V21 .0 "HNP U2 Nuclear Boiler System P&ID Sheet 3" 4. NUREG-1301 "Offsite Dose Calculation Manual Guidance:

Standard Radiological Effluent Controls for Pressurized Water Reactors." April 1991.5. NL-06-1637 "Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term." August 29, 2006.6. GE-NE-0000-0003-0634-01 "Edwin I. Hatch Nuclear Plant, Units I and 2, 10-PSI Dome Pressure Increase." July 2003.7. S16039, V2.0 "Instrument Data Sheet Process Radiation Monitoring System" 8. A16481 Sheet D11IA V2.0 "STACK GAS VELOCITY PROBE AND TRANSMITTER DATA SHEET" 9. A16481 Sheet D11B V0.0 "AREA RADIATION MONITORING SYSTEM" 10.A16481 Sheet D11C V0.0 "AREA RADIATION MONITORING SYSTEM" 11 .A26481 Sheet D11A V0.0 "AREA RADIATION MONITORING SYSTEM" 12.A26481 Sheet D11B V0.0 "AREA RADIATION MONITORING SYSTEM" 13. H16566 V7.0 "PROCESS RADIATION MONITORING SYSTEM P&ID (SHEET 4)" 14. H26017 V22.0 "FISSION PRODUCTS/POST-LOCA MONITORING SYSTEMS P&ID" 15. H16032 V5.0 "EQUIP LOCATION REAC BLDG SECT A-A" 16. H16241 V29.0 "INSTRUMENT

& PRIMARY PT LOC REAC&RADWASTE BLDG EL 130" 17. H26417 V17.0 "INSTRUMENT

& PRIMARY POINT LOCATIONS-REACTOR BUILDING-EL.

130'-0" & RADWASTE BUILDING-EL.

132'-4"" 18. Engineering Compendium of Radiation Shielding, NY 1968, Volume 1.19.64CI-OCB-004-1 V7.0 "UNIT ONE POST TREATMENT RADIATION MONITORS" 20. 64CI-OCB-004-2 V7.0 "UNIT TWO POST TREATMENT RADIATION MONITORS" 21.S $6039 V2.0 "INSTRUMENT DATA SHEET PROCESS RADIATION MONITORING SYSTEM" 22. SX27520 VI.0 "IDS -PROCESS RADIATION MONITORING SYSTEM" 23. SX1 8062 V2.0 "IDS PROCESS RADIATION MONITORING SYSTEM" 24. S30523 VO.2 "IODINE -NOBLE GAS SAMPLE PANEL 133D9025G1-G6 INSTRUCTION MANUAL (OPERATION AND MAINTENANCE)" 25.S$19207 V2.0" INSTRUCTION MANUAL RADIATION MONITORING SYSTEM VOLUME 4" 26. SX29455 V1 .0 "RADIATION MONITORING SYSTEM VOLUME IV"

27. H 16568 V5.0 "REACTOR PROTECTION SYSTEM P&ID" 28. BH2-M-V999-0047 V2.0 "DRYWELL EQUIPMENT EQ DOSES FOR EXTENDED POWER UPRATE FOR REA HT-96660" 29. HNP Technical Specifications 273/218 01 16 30. NUREG-0016 "Calculation of releases of radioactive materials in gaseous an liquid effluents from boiling water reactors (BWR GALE Code)" April 1976.31. RG 1.183 "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." July 2000 32. BH2-CS-52-2P33-01 V3.0 "Containment Hydrogen Analyzer" 33. BH2-CS-52-2P33-02 V2.0 "Containment Oxygen Analyzer" 34. Regulatory Guide 1 .7 "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident" Rev 1, Sept 1976.35. SS2102005 V6.0 "Furnishing

& Delivery of Reactor Drywell & Suppression Chamber-Containment Systems" 36. CALC F-86-03 V0.0 "COMPUTER CODE: VERIFICATION OF THE F GRODEC COMPUTER PROGRAM" 37. Deleted 38. 64CI-OCB-008-0 V8.1 "PLANT SERVICE WATER RADIATION MONITORS" 39. 64CI-OCB-009-0 V5.3 "LIQUID RADWASTE RADIATION MONITORING" 40. H16564 V29.0 "PROCESS RADIATION MONITORING SYSTEM P&ID SHT. 2" 41. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2" 42.64CI-OCB-002-1 V12.0 "UNIT ONE REACTOR BUILDING VENT RADIATION MONITORING" 43. 64CI-OCB-002-2 V16.0" UNIT TWO REACTOR BUILDING VENT RADIATION MONITORING" 44. H26013 V7.0 "PROCESS RADIATION MONITORING SYSTEM l.E.D. SHEET 3" 45.64CI-OCB-003-1 V14.0 "RECOMBINER BUILDING VENT RADIATION MONITORING" 46. H16528 V12.0 "OFF GAS RECOMBINER BUILDING VENTILATION SYSTEM P & ID AND PROCESS FLOW DIAGRAM" 47. 64CI-OCB-001

-0 Vi13.0 "MAIN STACK RADIATION MONITORING" 48.5$56256 V1.0 "GEl4 FUEL BUNDLE INTERFACE CONTROL" 49.$S54974 V0.3 "BWR SPENT FUEL STORAGE RACKS -RACK LAYOUT MPL. F16" 50. S54975 V0.1 "BWR SPENT FUEL STORAGE RACKS -CONSTRUCTION (ELEVATIONS)

MPL. F16" 51 .H 15602 V1 .0 "REAC BLDG FUEL TRANS POUR NL PLAN SECT&DET"

52. H15336 V10.0 "REAC BLDG SPENT FUEL POOL-PLAN

@ EL 228&EL SECT&DET POUR NL" 53. Deleted 54. H25963 Vl0.0 "ARCHITECTURAL

-EXTERIOR WALL AND ROOF DETAILS" 55. H25621 V2.0 "REACTOR BUILDING-STRUCTURAL STEEL-ROOF FRAMING AT EL.280'- 0" -PURLINS AND TOPCHORD" 56. H25694 V0.0 "REACTOR BUILDING-STRUCTURAL STEEL 280'-0"- VERTICAL SECTIONS" 57. H25953 V3.0 "ARCHITECTURAL-REACTOR BUILDING-SECTION A-A" 58. H15880 V4.0 "ACCESS CONT FLOOR PLAN ELEVATION 203 228&244-1 0" 59. H45515 V1.0 "I/O LIST SPDS RTP NODE 2 (SHEET 1)" 60. H16562 V7.0 "AREA RADIATION MONITORING SYSTEM I.E.D" 61. H26010 V10.0 "AREA RADIATION MONITORING SYSTEM I.E.D" 62. H27805 V22.0 "AREA RADIATION MONITORING SYS. 2D21 ELEMENTARY DIAGRAM SHT. I OF 8" 63. H25990 V3.0 "ACCESS CONTROL -FLOOR PLAN -EL. 203' AND 228"'64. S55932 V1.0 "DOSE RATES FROM THE HI-STORM 100 SYSTEM F(OR THE HATCH ISFSI MPL F18" 65. HNP Dry Storage FSAR "Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System" Revision 7, http://nuclear.southernco.com/regqulatory-affairs/H NP-Dry-Cask.html

66. H15878 V14.0 "ACCESS CONT FLOOR PLAN ELEVATION 158" 67. HNP ODCM V23.0 "Offsite Dose Calculation Manual for Hatch Nuclear Plant" 68. FGR12 Sep 1993 "External Exposure to Radionuclides in Air, Water, and Soil" 69. FGR1 1 Sep 1988 "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." 70.5SX18761 V2.0 "P & I DIAGRAM KMG -HRH" 71.S$41454 V2.0 "P & I DIAGRAM KMG-HRH" 72. DoclD: RE203727981 VI .0 "GASEOUS EFFLUENT REPORTS" for procedure 64CH-R PT-007-0.73. DoclD: RE203186522 VI.0 "Form HPX-0893 to procedure 64CI-OCB-003-1" 74. S25213 Rev G "NUC BOILER SYS-DESIGN SPEC" 75. H16565 V5.0 "PROCESS RADIATION MONITORING SYSTEM P. & I.D. SHT. 3" 76. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2" 77. H16014 V39.0 "REACTOR BUILDING REFUELING FLOOR VENTILATION SYSTEM P. & I. D."

Southern Nuclear 0perating Cornpany Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMvNH-13-021 COMPANY Unit: 1&2 SHEET 17 78. H27158 V6.0 "PROCESS RADIATION MONITORING SYSTEM 2D11 ELEMENTARY DIAGRAM SHEET 16 OF 22" 79. Lamarsh, "Introduction to Nuclear Engineering," 2nd edition, 1983 80. H25942, V14.O "ARCH REAC BLDG-FL PLN-EL 130" 81.ASME International Steam Tables for Industrial use. 2 nd edition.82.64CI-OCB-005-1 V15.0 "UNIT ONE FISSION PRODUCT MONITOR" 83.,64CI-OCB-005-2 V15.0 "UNIT ONE FISSION PRODUCT MONITOR" 84. SS6902005 V3.0 "REACTOR DRYWELL & SUPPRESSION CHAMBER VESSELS &APPURTENANCE" 85. BH1-CS-33-P33-06 VI1.0 "CONTAINMENT 02 ANALYZER" 86. SMNH-93-029 VO.0 "LIQUID EFFLUENT DOSES NORMAL OPERATION" 87. Shultis and Faw "Fundamentals of Science and Engineering" 2002 88. BH2-M-V999-0048 V0,0 "SPENT FUEL POOL BOILING DOSE RATE" 89. Gieck "Engineering Formulas" 7 th edition.90. SI15213 V0.2 "RPV VESSEL ASSEMBLY-PF-l1983-63-8" 91. Courtney "A Handbook of Radiation Shielding Data" July 1976 92. 64CI-OCB-006-1 VI14.0 "UNIT ONE PRETREATMENT RADIATION MONITORING" 93. 64CI-OCB-006-2 V12.0 "UNIT TWO PRETREATMENT RADIATION MONITORING" 94.31E0-EOP-014-1 V12.0 "SC -SECONDARY CONTAINMENT CONTROL RR -RADIOACTIVITY RELEASE CONTROL" 95. 31 EO-EOP-014-2 V1 1.0 "SC -SECONDARY CONTAINMENT CONTROL RR -RADIOACTIVITY RELEASE CONTROL" 96. SMNH-05-009 V2.0 "NEI 99-01 EAL Calculations" 97. SNC024-CALC-007 Rev. 0 "HNP Determination of Emergency Action Level for Initiating Condition E-HUI" (Attachment L)98. SNC024-CALC-008 Rev. 0 "Hatch EALs RA1 Threshold to Address NEI 99-01 Revision 6" (Attachment M)

Method of Solutions:

NEI 99-01 Revision 6 Methods conform to the guidance of NEI 99-01 Revision 6. Detailed descriptions of the methods are included in the individual EAL threshold calculations in the Analysis section of this calculation.

Use of Regulatory Guide 1.183, Alternate Source Term Method The NEt 99-01 Revision 6 Recognition Category A (Abnormal Rad Levels/Radiological Effluent)Initiating Conditions (ICs) for declaring an Alert, a Site Area Emergency, and a General Emergency (Emergency Action Levels RA1, RS1 and RG1, respectively) are expressed in terms of Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (CDE).Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," is not the current license basis for performing dose calculations for Hatch. However, it expresses doses in terms of Whole Body and Thyroid.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," does express doses in terms of TEDE and CDE, is the current licensing basis for performing dose calculations for Hatch. However, per section 1.1.4 on page 1.183-6, "This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with emergency dose projections, protective measures, and severe accident management guides." Per section 4.1.1 of RG 1.183, TEDE is defined as the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure.Per section 4.1.2 of RG 1.183, Table 2.1 of Federal Guidance Report 11 provides tables of conversion factors acceptable to the NRC staff. The dose conversion factors (DCFs) factors in the column headed "effective" yield doses corresponding to the CEDE.Per sections 4.1.4 and 4.1.5 of Reg Guide 1.183, the DDE should be calculated assuming submergence in a semi-infinite cloud for the most limiting person at the EAB. The effective dose equivalent (EDE) from external exposure is nominally equivalent to the DDE, thus EDE may be used in lieu of DDE in determining the external dose contribution to the TEDE. Table II1.1 of Federal Guidance Report 12 provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

Southern Nuclear 0peratinl Cornpany ISOUTH=ERNA.

Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMH1-2 COMPANY Unit: 1&2 SET1 TEDE & Thyroid ODE Calculations The basic process for calculating an offsite dose consists of first determining the concentrations of radionuclides in the release stream, be it air, steam, or water. The release stream concentration is determined by dividing the release rates of the radionuclides of interest, expressed as microcuries (pCi) per unit time, by the release fluid's volumetric flow rate, expressed as cubic centimeters (cc) per unit time: pJCi/cc = [pCi/unit time]/[cc/unit time]As we are back-calculating release concentrations based on pre-established dose limits (100 mREM TEDE and 500 mREM Thyroid ODE), the upstream modeling of the specific release paths is not necessary.

The gaseous effluent noble gas radiation monitors care not a whit how those radionuclides arrive at them.Step 1: Identify the radionuclides of interest.

Select the same radionuclides used to calculate doses for the design basis accidents in FSAR chapter 15: the fission product noble gases and iodines. The other fission products and activated corrosion products are particulates and will not contribute significantly to the offsite dose.Step 2: Determine the RCS coolant radionuclide activity for each radionuclide (Xrcs-i pCi/g). This is assumed to be the sum of core fission product inventory released during a LOCA divided by RCS coolant mass (Mrcs g) and the equilibrium RCS coolant activity (Xeq-i) for that radion ucl ide.Xrcs-i = Xeq-i + (1 .0E+06 pCi/1 Ci) x [Core Inventory (Ci)] x [Release Fraction]/(Mrcs g)For no fuel damage, the release fraction is 0 and the RCS activity is the equilibrium RCS coolant activity.

If fuel damage is assumed (release fraction > 0), the quotient of core inventory and RCS coolant mass will be orders of magnitude greater than the contribution from the coolant equilibrium activity.Step 3: Convert coolant activity (Xrcs-i pCi/g) to release stream activity (Xris-i pCi/cc). This conversion is accomplished by multiplying the RCS coolant activity by a dimensionless partition factor (PF 1) and an arbitrarily selected density, pris g/cc: Xrls-i (pCi/cc) = (Xrcs-i pCi/g) x PFi x (pris g/cc)The partition factor will depend on the radionuclide and the release path. The partition factors used in this calculation are discussed in Assumption

1. 4 of this calculation.

Arbitrarily set pris = 1.0 g/cc to make the math easy. The justification for this will be provided in Step 9.Step 4: Determine radionuclide concentration at Exclusion Area Boundary (XEAB-i pCi/cc). This is done using standard dose assessment methods. The release concentration is multiplied by the release volumetric flow rate (Q~s m 3/sec) and the diffusion coefficient

[(X/Q) m 3/sec]: XEAB-I (pCi/cc) = Xris-1 (pCi/cc) x [Qris (m 3/sec)] x [(X/Q) (m 3/sec)]Step 5: Calculate the TEDE for each radionuclide for one hour exposure time at EAB. This is done using the appropriate FGR-11 and FGR-12 dose conversion factors (DCFs), as discussed in the previous subsection.

Southern Nuclear 0peratinl Cornpany SOUTHERNZ[

lnt N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET3202 COMPANY Unit: 1&2 SET2 TEDEi (mREM) = External Exposure + Internal Exposure TEDE 1 (mREM) = XEAB-I (pCi/cc) X texp (hours) x DDEDCF-I [(mREM/hr)/(IJCi/cc)]

+XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CEDEDOF-i (mREM/pCi)

TEDEi (mREM) = XEAB (pCi/cc) X texp (hours) x TEDEDCF-i

[(mREM/hr)/(pCi/cc)]

where TEDEDCF-i

[(mREM/hr)/(pCi/cc)]

= DDEDCF-I [(mREM/hr)/(pCi/cc)]

+BR (cclhr) x CEDEDCF-i (mREM/pCi)

BR (cclhr) = breathing rate Step 6: Add the individual TEDEs to obtain the TEDE for the release (TEDEr~s):

TEDEris = Z [TEDEi]TEDEris = [Xris-i X (X/Q) X Qris x texp x TEDEDCF-I]

TEDEris = [(X/Q) x Qris x texp ] x $" [Xris-i x TEDEDcF-i]

Step 7: Calculate Thyroid CDE for each Iodine isotope for one hour exposure time at EAB. This is done using the appropriate FGR-11I dose conversion factors (DC Es), as discussed in the previous subsection.

CDETHY-i (mREM) = XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CDETHY-DCF-i (mREM/pCi)

Step 8: Add the individual Thyroid CDEs to obtain the Thyroid CDE for the release (CDEris): COEris = E' [CDETHY-i]

CDEris = Z [Xrls-i X (X/Q) x Qris x BR X CDETHY-DCF-i]

CDEris = [(XIQ) x Qris x texp x texp] x X [Xris-i X TEDEDcF-I]

Step 9: Determine the RS1 EALI 100 mREM TEDE threshold release concentrations for each noble gas (Xi00-1 pCi/cc). This is done by multiplying each noble gas' release concentration (Xris-i IJCi/cc) determined in Step 3 by the quotient of 100 mREM and the sum of the TEDEs for all of the radionuclides considered (TEDEr~s mREM). Only noble gas concentrations are adjusted because the gaseous effluent monitors are noble gas detectors.(Xi0o-i pCi/cc)/(Xris-i pCi/cc) = (100 mREM)/(TEDEris mREM)Xi00-i (pCilcc) = (Xrls-i pCi/cc) x (100 mREM)/(TEDEris mREM)The following demonstrates that the arbitrarily assumed release stream density has no effect on the final result.Xris-i x (100 mREM)Xl00-i =[(X/Q) x Qris x texp ] x $" [Xrls-i x TEDEDcF-i]

Xrcs-i x (1.0) X pris x (100)Xl0 0-i =[(X/Q) x Qris x texp X [Xrcs-i x PFi x pris x TEDEDcF-i]

Xrc-i x pris x (100)Xloo-J =pris X (X/Q) x Qris x texp x T' [Xrcs-i X PFi x TEDEDCF-i]

Xrcs-i X Prls X (100)Xl 0 0-i =pig x (X/Q) x Qris x Z [Xrcs-i x PFi x texp x TEDEDcF-i]

Xrcs-i X (100)X 1 o 0-i-=r(X/Q) X Qris x texp X E [Xrcs-i X PFi x TEDEDcF-i]

The assumed release stream density has no effect on the final result: it cancels out. Thanks to the power of Excel, it is easier to calculate a postulated dose rate and adjust release concentrations than to set up the above equations.

Now to perform a dimensional check: Xris-i X (100 mREM)Xloo-i =[Xris-i X (X/Q) x Qris x PFi x texp x TEDEDcF-i]

? ~(pCi/cc) x mREM IJCi/cc =(pCi/cc) x (sec/rn 3) x (m 3/sec) x (hour) x [(mREM/hour)/(IJCi/cc)]

9 X tRRE-M p Ci/cc=i,-,., ,c,. x x (ms/see) x (heufr) x [(mRE-M/heu-f)/(p Ci/cc)]pCi/cc 1 1/(p Ci/cc)]pCi/cc = pCilcc Step 10: Determine the RS1 EAL1 500 mREM Thyroid CDE threshold release concentrations for each noble gas (X50oTr-i pCi/cc). This is done using the same method as in Step 9. Again, the arbitrarily assumed release stream density cancels out and has no effect on the final result.Several general trends can be inferred from the equation derived in Step 9 above. Holding other factors constant:* Increasing the diffusion coefficient (X/Q m 3/sec) will reduce the 100 mREM release concentration.

• Increasing the release flow rate (Q CFM) will reduce the 100 mREM release concentration.

• Increasing the exposure time (t hours) will reduce the 100 mREM release concentration.

Southern Nuclear 0peratin9 Cornpany SOUTHERNA4 Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations I SMNH-13-021 COMPANY Unit: 1&2 I SHEET 22*Increasing the total release (£ [Xrcs-J x PFi x pris x TEDEDGF-i])

will reduce the 100 mREM release concentration.

Body of Calculation:

RU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2 OR 3)1. Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer.Liquid Effluent Monitors (HNP ODCM Table 2-4) 2 X Setpoints Liquid Radwaste Effluent Line (Detector D1 1-N007, Indicator D1 1-K604)(Ref. 64C1-0CB-009-0, H 16564, H26012)Service Water System Effluent Line (Detector D1 1-N008, Indicator D1 1-K605)(Ref. 64C1-0CB-008-0, H 16564, H26012)Gaseous Effluent Monitors (HNP ODCM Table 3-4) 2 X Setpoints Reactor Building Vent Stack (Detector 1D1 1-N020, Indicator Dl11K619, Detector 2D1 1-NO26A/B, Indicator 2D1 1K4636)(Ref 64CI-OCB-002-1, 64C1-0CB-002-2, H16564, H260 13, H260 12)Recombiner Building Vent (Detector D1 1-N078 Indicator Dl11R763A Detector D1 1-N079, Indicator D 11R763B)(Ref 64C1-0CB-003-1, H 16528)Main Stack (Detector D1 1-N071 Indicator D11IK6OOA, Detector D1 1N072 Indicator D 11K600B)(Ref 64C1-0CB-001-O, H 16564)2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the 0DCM limits for 60 minutes or longer.

Southern Nuclear 0perating Cornpany SOUTHERNE Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SM -102 COMPANYr Unit: 1&2 SHEET 24 RU2 UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability:

All Emergency Action Levels: (1)1.a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

LPersonnel report of iow water level.SEP Low level alarm (1/2G41N372)

Per PDMS 1/2G41N372 (LS-N3 72) spent fuel pool low level alarm is set at EL225'9" and personnel report of low water level.AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors 1 D21 -K601 A -Rx Head Laydown Area 1ID1I-K601 D -Refuel Floor 1 D21-K601 E -Drywell Shield Plug 1D21-K601 M -Spent Fuel Pool and New Fuel Storage area 2D21-K601 A -Rx Head Laydown Area 2D21-K601 M -Spent Fuel/Fuel Pool Areas 2D21-K601 E -Dryer/Separator Pool 2D21-K611 K -RPV Refuel Floor 228'2D21-K61 1 L -RPV Refuel Floor 228'The Drywe// radius is 18'10" -19' (H255 70) this approximation is reasonable for radiation calculations.

To calculate distance to drywel/ edge from radiation detector the radius of drywell is subtracted from distance from radiation detector to drywell center.

Southern Nuclear 0perating Cornpany SOT N4L Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY Unit: 1&2 SHEET 25 Distance to Monitor Detector Name Elevation edge of Range Ref drywell H 15880 1D21K601A 1D21NOO2A 233Head7"111548'H45515 Laydown Area 230 67-9=4'mR/'hr H66 H 15880 ID2Ik601B 1D21NOO2B Refueling Floor 233'0" 95"-19' = 76' 1-1E4 H45515 Stairway mR/hrH162 H 15880 1D21K601D 1D21NOO2D Spent Fuel Pool 23'" 75'"19' = 56' 1-1E4 H45515 Demin Equip mR/hr H15880 1D21K601E 1D2lNOO2E Drywell Shield 233'0" 21"-19' = 2' 1-1E4 H45515 Plug mR/hr H 16562 H15880 1D21K601M 1D21NOO2M 233'P 2"N9'W7'H45515 Fuel Storage 23'" 24'=7 mR/hr H66 H26010 2D21K601A 2D21NOO2A RHed 228'0" 60"-19' = 41' 1E4H27805 Laydown Area mR/hr___________H25990 H26010 2D21K601E 2D21NOO2E ryrSpo raol 228'0" 76"-19' = 57 1-1E4 H27805__________H25990 1E4H26010 2D21K601M 2D21NOO2M SFP Area 228'0" 75"-19' = 56' mhrH27805 H25990 H26010 2D21K611K 2D21NO12K Reco esl 228'0" 71'-19' = 52' 114H27809 Refueling mR/hr H59 H260 10 2D21K611IL 2D21NO12L Reactor Vessel 228'0" 62"19' = 43 1-1E4 H27809______ Refueling mR/hr H25990 Southern Nuclear 0peratinl Cornpany ISOUIHERNZ._

Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-026 COMPANY Unit: 1&2 SET2 RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2 OR 3 OR 4)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer.This calculation is performed in Attachment M.Reactor Building Vent 2.6E-02 pCi/cc Range 1E-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D11P005 Detector 1/2D11N048 and 1/2D11N049 (Ref SX18761, S41454)Main Stack 8.1E+01 pCi/cc Range IE-3 to 1E5 uC~icc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1D11P006 Detector 1D11N0055 and 1D11N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the Site Boundary.3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the Site Boundary for one hour of exposure.4. Field survey results indicate EITHER of the following at or beyond the Site Boundary:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid ODE greater than 50 mrem for one hour of inhalation.

RA2 Significant lowering of water level above, or damage to, irradiated fuel.Operability Mode Applicability:

All Emergency Action Levels: (Table0R1 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on ANY Table R1 radiation monitors.Table Ri Refuel Floor Area Radiation Monitors Unit i Unit 2 1D21-K601 A -Rx Head Laydown Area 2D2l-K601 A -Rx Head Laydown Area 1D21 -K60 1 D -Refuel Floor 2D2 1-K601I M -Spent Fuel/Fuel Pool Areas 1D21-K601 E -Drywell Shield Plug 2D21-K601 E -Dryer/Separator Pool 1D21-K601 M -Spent Fuel Pool and New 2D21-K61 1 K -RPV Refuel Floor 228'Fuel Storage area____......

_________________,__________

2D21-K611 L -RPV Refuel Floor 228'Refuel Floor Ventilation Monitors Unit 1 Unit 2 ID11l-K609 A-D -Rx Bldg. Potential 2D1 1-K(609 A-D -Rx Bldg. Potential Contaminated Area Vent Exhaust Rad Contaminated Area Vent Exhaust Rad Monitor Monitor 1D 1l-K(611 A-D -Refuel Floor Vent Exhaust 2D1 1l-K61 1 A-D -Refuel Floor Vent Exhaust', :' ": .....2D1 1-K(634 A-D -Refuel Floor Rx Well Vent.______...._______________Exhaust

-. * ** ,2D1 1-K(635 A-D -Refuel Floor DW/Sep.________,_______,_______________________

Vent. Exhaust Southern Nuclear 0perating Cornpany ISOUTHERN E1, Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations j SM NH-13-021 COMPANY Unit: 1&2 SHEET 28 Monitor Detector Name Elevation Distance Range Ref to edge of_____ ____ _ ___ ____ drywell _ _ _ __ _ _ _Refuel Floor Area Radiation Monitors: 1D2IK601A ID21NOO2A Rx Head 233'0" 67'-19' = 1-1E4 H15880 Laydown Area 48' mR/hr H45515_________H16562 1D21K601B 1D21NOO2B Refueling Floor 233'0" 95"-19' = 1-1E4 H15880 Stairway 76' mR/hr H45515____ ____ ___ ____ __ _H 16562 1D21K601D 1D21NOO2D Spent Fuel Pool 233'0" 75"-19' = 1-1E4 H 15880 Demin Equip 56' mR/hr H45515 H 16562 1D21K601E 1D2INOO2E Dirywell Shield 233'0" 21'-19' = 1-1E4 H15880 Plug 2' mR/hr H45515_______H16562 ID21K601M ID2INOO2M SFP & New 233'0" 26"-19' = 1-1E4 H15880 Fuel Storage 7' mR/hr H45515 H16562 2D21K601A 2D21NOO2A Rx Head 228'0" 60'-19' = 1-154 H26010 Laydown Area 41' mR/hr H27805____________H25990 2D21K601E 2D21NOO2E Dryer/Separator 228'0" 76'-19' = 1-1E4 H26010 Pool 57' mR/hr H27805 H25990 2D21K601M 2D21NOO2M SFP Area 228'0" 75"-19' = 1-1E4 H26010 56' mR/hr H27805_________

_______ H25990 2D21K611lK 2D21NO12K Reactor Vessel 228'0" 71"-19' = 1-1E4 H26010 Refueling 52' mR/hr H27809 H25990 2D21K611L 2D21NO12L Reactor Vessel 228'0" 62"-19' = 1-1E4 H26010 Refueling 43' mR/hr H27809____ ___ __ ____ ___ ___ __ ____ _ _ ___ ___ H25990 Southern Nuclear Operating Cornpany Pan: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-029 COMPANY Unit: 1&2 SET2 Monitor Detector Name Range Ref Refuel Floor Ventilation Monitors: 1/2D11K609 1/2D11N010 Rx Bldg H16565 A-D A-D Ventilation H26013 Exhaust 0.01- FSAR Radiation lOOmR/hr Table Monitor 11.4-1 PDMS 112D11K611 1/2D11N012 Refuel Floor 0.01- H26012 A-D A-D Vent Exhaust lOOmR/hr H16014 FSAR Table 11.4-1___ ___ __ ___ ___ ___ ___ PDMS 2D11K634 2D11N024 Refuel Floor 0.01- H26012 A-D A-D Exhaust lOOmR/hr H27158 2D11K635 2D11N025 Refuel Floor 0.01- H26012 A-D A-D Exhaust lOOmR/hr H27 158 3. Lowering of spent fuel pool level to (site-specific Level 2 value)*.*The level 2 value was not available at the time of this calculation.

RS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2OR 3)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: This calculation is performed in Attachment I.Reactor Building Vent 2.6E-1 pJCi/cc Range 1E-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D1 1P005 Detector 1/2D1 1N048 and 1/2D1 1N049 (Ref SX18761, S41454)Main Stack 8.1E2 pJCi/cc Range IE-3 to lE5 uCi/cc (ESAR U2 Table 11.4-1 sheet 1)Monitor 1DI1PO06 Detector 1D11N0055 and ID1 1N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid ODE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the Site Bounday.* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Southern Nuclear Operating CornpanyPlant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY Unit: 1&2 SHEET 31 RG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR20OR 3)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Reactor Building Vent 2.6E0 tpCi/cc Range 15-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D11P005 Detector 1/2D11N048 and 1/2D11N049 (Ref SX18761, 341454)Main Stack 8.1E3 pJCi/cc Range 1E-3 to 155 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1D11PO06 Detector 1D11NO055 and 1D11N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the Site Boundary:* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

CA1 Loss of RPV Inventory.

Operability Mode Applicability:

Cold Shutdown, Refueling Emergency Actuation Levels (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).

According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point. According to H16 145 and H26 189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35".2.a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level increase in any of the following due to a loss of RPV inventory.

Drywell Floor Drain Sumps Drywell Equipment drain Sumps Torus Torus room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks Southern Nuclear Operatingl Company SOUTHE=RNZ4 lnt N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET3302 COMPANY Unit: 1&2 SET3 CS1 Loss of RPV inventory affecting core decay heat removal capability.

Operability Mode Applicability:

Cold Shutdown, Refueling Emergency Action Levels: (1 OR 2 OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.

AND b. RPV level less than -41" (6" below the Level 2 actuation setpoint).

According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point.According to H16145 and H26 189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35". Therefore

-35"-6" =-41" 2.a. Secondary CONTAI NMENT INTEGRITY established.

AND b. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad Baffier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:

• UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery: Dryeli Floor Drain Sumps Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.

Operating Mode Applicability:

Emergency Action Level: Cold Shutdown Refueling (1 OR 2)t.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".This can be rounded to -158" AND b. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by the following:

• UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery: Drywell Floor Drain Sumps Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps Radiation monitor readings indicative of core unco very are investigated in Attachments J and K resulting in no monitors able to provide on-scale indications of core unco very.AND c. ANY indication from the Containment Challenge Table C1 Containment Challenge Table C1 Containment H2 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:

greater than 56 psig Secondary CONTAINMENT INTEGRITY NOT established*

Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Damage to a loaded cask CONFINEMENT BOUNDARY E-HU1 Southern Nuclear 0perating Cornpany SOUTHERNR lnt N Title: NEI 99-01 Rev 6 EAL Calculations SNHEET3502 COMPANY Unit: 1&2 SET3 Operating Mode Applicability:

Emergency Action Level: ALL (1)1. Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table El.The calculation is performed in Attachment L.Table El Location of Dose Rate [ Total Dose Rate____________________(Neutron

+ Gamma mRlhr)HI-TRAC 125 Side -Mid- height [450 Top 11l0 HI-STAR 100 or HI-STORM 100 Side -60 inches below mid- 80 height Side -Mid- height 80 Side -60 inches above mid- 30 height Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 Southern Nuclear 0perating Cornpany latHN Title: NEI 99-01 Rev 6 EAL Calculations SMH3-2 COMPANYW Unit: 1&2 SHEET 36 Fission Product Barrier Emergency Action Levels Fuel Clad Barrier: Emergency Action Levels Fuel Clad Barrier Potential Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be determined.(Ret. H16 145 and H26189)According to COLR for HNP the currently used fuel is GEl4. 'According to NED C-32868P Rev 5 Appendix A (Reference of the COLR) the fuel length for GEl4 fuel was increased from 148" to 150" inches. The Appendix A is attached below. Thus the top of the fuel per TS Bases 2.1.1.3 is 158.44 inches below instrument zero.According to 31EO-EOP-015-1 and 31EO-EOP-015-2 "CPI Flow Chart" operators are instructed to maximize water injection rates from alternate injection subsystems when reactor water level drops below -155 inches of instrument zero. This value is more conservative than the actual TOAF level.

NEDC-32868P

-Revision 5 GNF Proprietary Information

-Class ill Appendix A Minor Modifications to the GEl4 Design The nominal pellet density for thle GEI4 design has been increased from 95_5% to 97.0% of theoretical density (TD). The upper specification limit, however, has not been changed. In addition, for the full length U0 2 fuel rods, barrier and non-barrier tubing designs, the active fuel length has been increased from 3759 mm (148 in.) to 3810 mm (150 in.).All GESTAR II criteria identified in Sections 2.1 through 2.13 have been re-evaluated and it has been demonstrated that all criteria are satisfied.

The results of the evaluations are contained in DRF J11-03667-00 and DRF J1 1-03467-01.

As part of the evaluation, a new NRC approved method was used for the ATWS analysis (Section 2.14). The ODYN one-dimensional transient analysis code was used to perform the required ATWS analyses.

All results are below the specified limits for compliance and are documented in DRF B21-00676-00.

The NRC approved version is documented in the lollowing:

MFN-99-029, Letter, S. Richards (NRC) to J, Kiapproth (GE), "~Safety Evaluation by the Office of Nuclear Reactor Regulation on NEDC-24154P, Supplement 1", November 18, 1999.MFN-052-98, Letter, T. Essig (NRC) to J. Quirk (GE), "Safety Evaluation by the Office of Nuclear Reactor Regulation on NEDC-24154P, Supplement 1 (TAC No.MA3478)", November 17, 1998.

Fuel Clad Barrier Loss Threshold 4.A DWRRM greater than 1,400 R/hr.In Attachment C the detector radiation level of 1.4E3 R/hr was calculated.

The calculation used core inventory from NL-06-1637 to calculate isotopes concentrations.

The calculation for DEI131 was performed to find a ratio to DEl 300uCi/gm.

GRODEC was used to calculate the fluence within the drywell.Cylinder geometry was used to calculate the geometric fraction.Fuel Clad Barrier Loss Threshold 5.A Offgas Pre- and Post-Treatment Monitors Offscale High AND Fission Product Monitor Offscale High.Attachment D performed an evaluation for Offgas Pre- and Post-treatment monitors D1 1-K615 (section A of Attachment D) and Containment fission product monitors D11IP010 (section B of Attachment D). It was found that these instruments will be off scale.RCS Barrier: Emergency Action Levels RCS Barrier Loss Threshold 1I.A Primary containment pressure greater than 1.85 psig due to RCS leakage.LIS 1C71N650A-D 1.85 psig Ref. PDMS LIS 2C71N650A-D 1.85 psig Ref. PDMS References (H 16568, POMS)RCS Barrier Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be determined (Ref. H16145 and H26189)The reactor vessel top of active fuel was calculated in Fuel Cla'd Barrier Potential Loss 2.A.RCS Barrier Loss Threshold 4.A DWRRM greater than 40 R/hr.In Attachment E the detector radiation level of 40 R/hr was calculated.

The calculation used core inventory from NL-06-1 637 to calculate isotopes concentrations.

The calculation for DEI13I was performed to find a ratio to DEl Southern Nuclear 0peratinl Cornpany SOUTHERNE4i Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations I SMNH-13-021 COMPANY Unit: 1&2 I SHEET 39 2uCi/gm. GRODEC was used to calculate the f/uence within the drywell. Cylinder geometry was used to calculate the geometric fraction.RCS Barrier Loss Threshold 5.A Drywell Fission Product Monitor reading 5.0 x 105 cpm.This EAL is to cover drywve// fission product monitor indications that may indicate loss or potential loss of the RCS barrier. Attachment F determined that the reading on drywe// fission product monitor Dl 1K630 of 1E6 cpm will indicate potential loss of RCS barrier. Per SX18062 page 34 the monitor K630 range is 10 to lO^6 cpm.Primary Containment Barrier: Emergency Action Level: Primary Containment Barrier Potential Loss Threshold 1l.A Primary containment pressure greater than 56 psig.Containment Design Pressure:

56 psig (SS2 102005 section 304)(SS6902005 section 304)Primary Containment Barrier Potential Loss Threshold I .B Greater than or equal to 6% H2 AND 5% 02 exists inside primary containment.

Explosive mixture inside containment

>_6% Hydrogen (Ref. RG1.7 pgl. 7-6)> 5% Oxygen (Ref. CALC BH2-CS-52-2P33-2 pg 4 and 9)(Ref. CALC BH1-CS-33-P33-06 pg 8 & A-I)Primary Containment Barrier Potential Loss Threshold 4.A DWRRM greater than 26,000 R/hr.The evaluation of expected radiation readings on DWRRM (Dl11K621) was performed in Attachment G of this calculation.

The detector is expected to read 2.6E4Rlhr.

The range of this instrument is 1-10^7 R/hr (established in attachment c).

SU3 Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode: Power Operation Startup Hot Standby Hot Shutdown Emergency Action Levels: (1 OR 2)1. Pretreatment Radiation Monitor reading greater than 240,000 pCi/sec for greater than 60 minutes.According to Tech. Spec. Bases 3.7.6 page B3. 7-31 and Tech. Spec. section 3.7.6 page 3. 7-16, the gross gamma activity rate of the noble gases measured at the main condenser evacuation system pretreatment monitor station shall be <240mCi/second or <240, O00pCi/second.

According to 64C1-0CB-006-1/2 procedures the offgas pretreatment radiation monitors are 1/2D1 1-K601 and 1/2D1 1-K602 OR 2. Sample analysis indicates that the reactor coolant specific activity is EITHER:* Greater than 0.2 pCi/gm and less than or equal to 2.0 pCi/gm dose equivalent 1131 for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.* Greater than 2.0 pCi/gm dose equivalent h131.According to Tech. Spec. Bases 3.4.6 page B3. 4-25 and Tech. Spec. section 3.4.6 page 3. 4-11, the specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity 0.2 pCi/gm. A condition

>0.2 IJCi/gm but 2.0 IJCi/gm must be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A condition

>2pC~igm requires immediate action.

Southern Nuclear Design Calculation IPlant: HNP Unit: 1&2 Icalculation Number: SMNH'13-021 ISheet: A-I Attachment A -SNC Emergency Planning Concurrence Calculation Number: SMNH-13-021 Calculation Version: 1 Calculation Title: NEI 99-01 Rev 6 EAL Calculations I the undersigned have reviewed the subject calculation and concur that:* Its Methods of Analysis conform to the guidance of NEI 99-01 Revision 6* Its Assumptions are consistent with the guidance of NEI 99-01 Revision 6* Its conclusions are consistent with the Methods of Analysis, Assumptions, and Design Inputs..L 1 J' / t, Emergency PlanningSN Name / Signature 0 /Date / Organization Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 1 of 19 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors (Fuel Clad Barrier Loss Threshold 4.A).Defining Units that MathCAD will understand MWt :=r 106W Ci: *- Bq and 2.7.10[tCi := l0-6 Ci Rad := 0.01IGy Per FGR-11 page 219 Remn Remn:= 0.01Sv Per FGR-11 page 219 mRem := -1000 The source terms for noble gases and iodine's are provided in NL-06-1 637 Enclosure 1 page 16 and are shown below.I Isotopes : "Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe- 1381-13 1""I-132""1-133""I-134""1-135" Corelnventory

3.30E+03' 3 .78E+02 6.92E+03 1 .32E+04 1 .86E+04 2.26E+04 3.03E+02 5.27E+04 1 .58E+03 1.89E+04 1.09E+04 4.81IE+04 4.52E+04 2.72E+-04 3 .93E+04 5.52E+04 6.05E+04 5.16E+04 Ci MWt)

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 2 of 19 CorePower

=z 2818.MWt RCS inv :=9965ft 3 FRelease:=

0.05 DW_Volu 1 := 146010ft3 SPVol~u 1:= 112900ft 3 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.The dry well free volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.VolContainmentu1

= DWVo1lu 1+ SPVolu1 = 2.589 x 105-ft3 DW_Vo1u 2 :=146266ft 3 SP_Volu 2:= 109800ft The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol Containmentu 2:= DW_Volu2 + SPVolu 2= 2.561 x( 105.ft3 ContainmentVol Avg : VolContainm entul + VolContainmentu 2 53= 2.575 x 105.ft3 2 To determine the RCS inventory mass the water specific volume needs to found as follows: Ul: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 °F + 532.0 °F TAvgu 1 := 2= 462.2.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 °F TAvgu 2:= 2= 480.35.°F SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEl 99-01 Rev 6 EAL Calculations 3 of 19 The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is l060psia.ft 3 VRCS water :=0.0199161

--According to http://wwwv.spiraxsarco.

corn/resources/steam-tab les.asp the specific volume for 1060psia and 480.35 F conditions is 0.0199161ft^3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv8 RCS mass .- -2.27 x 108-gm'URCS water The concentration of isotopes in RCS is calculated as follows: CoreInventory.

CorePower.

F_Release COnCRCS isotopes : RCS mass augment(Isotopes, COnCRCS isotopes)

=0 1 0 "Kr-83m" 2.049"10-3 1 "Kr-85" 2.347" 0-4 2 "Kr-85m" 4.296"10-3 3 "Kr-87" 8.195" 0-3 4 "Kr-88" 0.012 5 "Kr-89" 0.014 6 "Xe-131lm" 1.881"10-7 "Xe-133" 0.033 C 8 "Xe-133m" 9.809"10-4' gin 9 "Xe-135" 0.012 10] "Xe-135m" 6.767"10-3 111 "Xe-137" 0.03 123 "Xe-138" 0.028 13 "I-131" 0.017 141 "I-132" 0.024 15 "I-133" 0.034 161 "I-134" 0.038 17 "I-135" 0.032 SPlant: HNP U1 & U2 SNC CALCULATION MN-301Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 4 of 19 The DEl for the RCS is calculated in the following work sheet: The Coolant Iodine activities are provided in NL-06-1637 Table 28 Enclosure 1.The dose conversion factors are provided in FGR 11. The RCS mass was calculated just above in this attachment.

The Iodine activities were taken from the calculated above matrix.gConCRCS isoto pes 1]3 Co"I-13_*t (0.017 " rI-132"15 COnCR S-i~ps4 /0.024 Iodines:=/'-3' Activity := ConcRCS isotopesl 5 /l0.034 C L0.038 gm k."-15"ConCRCS-isotopesl 6 .0.032 ConCRCS-isotopes 1 7 2.92E-07 1 .74E-09 BqFRlIhlaintbe DCFFGR11l:

4.86E-08 --~ The dose conversion factors are taken from K8.46E-09J The total exposure to iodines is calculated as follows: (1.826 x 104`157.23 5> Rem ExPiodines

= (DCFFGR1 1'Activity)

=6.169 x 103 gm 40.064 1.004 x 1O 3 SumExPiodines

=VEx..dne 4.6 x14Remn L~Podies 2.53 x10gm The DEI131 equivalent of the exposure from iodines calculated by dividing the Sum of iodines exposure by 1131 DCF factor: Sum-ExPiodines 14 ~ICi DEl131_eqv

= = 2.37 x 10 -DCFFGR110 gm Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-0 1 Rev 6 EAL Calculations 5 of 19 FDEJ300 := gm -0.0 13-~DE1131 _eqv This is the fraction that will be applied to iodine's and noble gases concentrations to adjust to RCS DEl 300 uCi/gm equivalent as described in NEI 99-01.Amount of isotopes in the drywell is found as follows: In this calculation the core inventory is multiplied by core power, containment release fraction, and the DEl of 300uCi/gm fraction.

The amount of total curies released are distributed throughout the drywell and suppression pool. The concentration then is multiplied by the U1 DryWell volume. This way we have the total curies inside the DryWell only. The DryWell for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.Corelnventory.

CorePower.F_Release.

FDE1300.DWVoI 1 1 2 Isotopes in DW: augment(Isotopes, Isotopes in DW) =ContainmentVol Avg*0 "Kr-83m" 3.343E+003

1. "Kr-85" 3.83E+002"2 "Kr-85m" 7,011tE+003 3 "Kr-87" 1.337E+004 4 "Kr-88" 1.884E+004 5 "r-9 2.29E+004 6 "Xe-131lm" 3.07E+002 7 "Xe-133" 5.339E+004 181 "Xe- 133m" 1.601E+003 9, "Xe-135" 1.915E+004

.10 "Xe-135mn" 1.104E+004 11: "Xe-137" 4.873E+004 12 "Xe-138" 4.579E+004 13 "I-131" 2.756E+004 14 "I-132" 3.981E+004 15 "1-133" 5.592E+004

16 "I-134" 6.129E+004 17 "1-135" 5.228E+004.Ci

[Plant: HNP Ul & U2 SNO CALCULATION SMNH-13-021 Attachment CI Title: NEI 99-01 Rev 6 EAL Calculations 6 of 19 I The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air mixture density for pressure 44.7psia and temperature of 343F (NL-06-1 637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

GRODEC does not have an input for absorption of coefficients.

DW_VoIlu 2 = 4.142 x 109.mL 1 -3 gm Psteam- -1.53xx 10 .ft 3 mL 10.4690 lb According to http://www, spiraxsarco.

com/resources/steam-tab les.asp the specific volume for 44.7psia and 343 F conditions is 10.4690ft^3/Ibm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to this attachment below. The drywell pressure and temperature were obtained from NL-06-1637 Enclosure 1 table 11 and table 12.According to PDMS the Drywell Wide Range detector is MPL 1D11N003A&B and 2D11N003A&B.

The detectors are connected to D11K621 Recorders.

As described in the following table: DrawingInstrument Tag Range MPL Sheet No R/hr Model N 1D11N003A A16481 DuIB QD11-RE-N003A lE0- 1E7 877-1 1D11N003B A16481 D11C QD11-RE-N003B lE0- 1E7 877-1 2D11N003A A26481 DlIA Q2D1l-RE-N003A lE0- 1E7 877-1 2D11N003B A26481 DluB Q2D11-RE-N003B lEO- 1E7 877-1 DWG H16566 DWG H26017 I I I I I I!I I I I 1 t!L.DWG H16566 GROSS Q~IM~ GUt~SftS l~A~.-i These drawings show that the detectors are inside the drywell.

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 7 of 19 DWG H16032 Per H16241 and H26417 the instrument location is at elevation of 156ft and approximately 27 ft from the center of RPV. This model will approximate the volume seen by the detector through a cylinder of 41.5 ft high. This number corresponds to the elevation above drywell floor. The radius of the cylinder will adjusted to the drywell free volume that is in the line of site of the detector.

Through the visual inspection of drawings H26417 and H16241 (shown below) it can be seen that about 2/3 of the drywell volume can be seen by the detector.E1llAllwl SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 8 of 19 Drawing H26417 Shows that approximately 2/3 of the containment volume is seen by the detector.Drawing H16241 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 9 of 19 As the result the geometry of detector in this volume will be modeled as a cylinder with the height of 41.5ft. The cylinder radius will be adjusted to match 2/3 of free volume of the drywell. The geometry factor will be evaluated for the point P1.Fig. 6.4.-7. Geometry for non-absorbing cylindrical volume source; interior, surface and exterior points.The uncolided fluxes at interior and exterior points of a non-absorbing cylindrical volume source are provided in Engineering Compendium on radiation shielding Volume 1 pages 381 and 382.(6.4.-19)3 H16032 H16241 and H26417 At P 1 : Sv (.& ;Heightcy1

= 156ft -114.5ft = 41.5.ft DWVolu 2 = 1.463 x 105.ft3 Volumecylinder

=fi r.R2*height table 34 Enclosure 1.Standard equation for cylinder volume.This is fraction of the containment line of site for the detector.FLine :-3 F FLine. DWVolu 2 RadiUSdetector

= = 27.348.ft 7r.Heightcy 1 Heightcyl

-1.5 17 RadiUSdetector the function (aR is determined via page 382 table 6.4-3 of Engineering Compendium of Radiation Shielding Volume 1. Linear interpolation will be used.

[ Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment~il:

E 90 e A acltos1 Cf1 xI:= 1.0 Heightcy -1 .1 x.-RadiUSdetector Y :=0.628 x3:= 2.0 3:=0.390 (x 2 -Xl)(Y 3 -Yl)+ 050 Y := (x 3_ Xl) +Y .0 Pa := Y'Pa = 0.505 Sv-Heightcy 1= 2 .'a ---= 3.1929198406362356406.Sv~m or Heightcy 1 = 1.265 x 103.cm Heq : 1.265 x 103 cm= 2ve 319.31217771913149332.Sv.c4 The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.1.6 ergs*.. o-MV-cm 2" Ia 4 3600S 10O0 ergs/(g .rad)

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 11 of 19 1 -.16 ergs cm 3 6 0 s MeV gm h Rad-hr or Geometricfactor.

Sv 10ergs g~rad 1.6.10- x 3600 -5= 5.76 x 10 100 Defining Units that MathCAD can process.q:=1.60219.10-1 9.coul eV := qe-volt Fundamental charge of electron page 6 of "Fundamental of Nuclear Science and Engineering'.

defines the unit of charge.MeV := 106.eV cc :=mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

(3.00E-01 4.OOE-0 I 5.00E-01 6.00E-0 1 8.00E-01 I1.00E+00 1 .50E+00 2.00E+00 3.00E+i00 4.00E+00.MeVSv:

1.00E+05 3 .62E+04 2.77E+05 4.76E+05 4.96E+05 7.3 8E+05 4.00E+05 3.35E+05 0.00E+00 MeV CO" Sec tI absair : r0.0288" 0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.0238 0.02112 cm gm

[Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 12 of 19 HeightcyI

= 1.265 x 103.cm oa=0.505 Calculated above Calculated above Radiationfield

= ,,Hegt~ absair =0 1 2 3 4 5 6 7 8 9 0 48.637 54.513 19.8 151 253.343 255.767 347.937 175.324 130. 176 0 Rad hr The unit conversion was handled by MathCAD internally.

This unit conversion was explained above in this attachment.

SM Radiationf ied :=

=1.436 x 103.a-Ic , hr~ccording to the S43177 Operator Manual Figure 1-4 the model number 877-1 detector has 1 to 1I atio of radiation present to radiation measured or shown. Therefore, the detector would read i.436E3R/hr.

This can be rounded to 1 .4E3R/hr.

The range of this instrument is I to 1 0A7 R/hr k stablished above).GRODE Inpu 1 18 60,3.343e+3,0 61 ,3.830e+2,0 62,7.011 e+3, 0 63,1 .337e+4,0 64,1.884e-'-4,0 65,2.290e+4, 0 141 ,2.756e+4,0 142, 3. 981e+4,0 143,5.592e+4,0 144,6. 129e+4,0 145,5.228e+4.0 FPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEL 99-0 1 Rev 6 EAL Calculations 13 of 19 146,3.070e+i2,0 147,5.339e+4,0 148,1.601e+3,0 149,1. 915e+'4,0 150,1.104e+4,0 151,4.873e+4,0 152,4.579e+4, 0 OUTPUT OF GRODEC CALCULATION DATA FILE NAME:FCEAL3 INPUT DATA LISTING 1 K<I-I-I-I-ISOTOPE INITIAL ADDITION ACTIVITY(CI)

RATE(CI/HR)(R-83M 3.343E+03 0.000E+i0 (R-85 3.830E+02 0.000E+00 (R-85M 7.011E+03 0.000E+0 (R-87 1.337E+04 0.000E+00 (R-88 1.884E+04 0.000E+00 (R-89 2.290E+04 0.000E+00-131 2.756E+04 0.000E+00-132 3.981E+04 0.000E+00-133 5.592E+04 0.000E+00-134 6.129E+'04 0.000E+00-135 5.228E+04 0.000E+00 (E-131M 3.070E+02 0.000E+(.E-133 5.339E+04 0.000E+0(.E-133M 1.601E+03 0.000E+(.E-135 1.915E+04 0.000E+0((E-135M 1.104E+'04 0.000E+(.E-137 4.873E+-04 0.000E+0OC (E-138 4.579E+04 0.000E+0()0 0 00 0 00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

= 1000 OR 1.000000 HOURS Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 14 of 19 TOTALACTIVITY

=323379.000000 CURIES ISOTOPE ISOTOPI NUMBER NAME E INITIAL ADDITION ACTIVITY ACT(CI) RATE(CI/HR) (Cl)3.343E+03 0.000E+00 2.303E+03 60 61 62 63 64 65 75 76 77 84 141 142 143 144 145 146 147 148 149 150 151 152 161 163 164 171 KR-83M KR-85 KR-85M KR-87 KR-88 KR-89 RB-87 RB-88 RB-89 SR-89 1-131 1-132 1-133 1-134 1-135 XE-131M XE-I133 XE-I133M XE-I135 XE-I135M XE-I137 XE-138 CS-I135 CS-1 37 CS-i138 BA-137M 3. 830E+02 7.011 E+03 1 .337E+04 1 .884E+04 2. 290E+ 04 0.000OE+00 0.000E+00 0.000E+00 0.000E+00 2.756E+'04 3.981E+04 5. 592E+04 6. 129E+04 5.228E+04 0.000E+'00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000OE+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+'00

3. 830E+i02 5.989E+03 7.737E+03 1.471E+04 4.800 E-02 I1.696E-11I 1.441 E+04 4.006E+02 8.779E-0I 2.746E+04 2. 930E+04 5.404E+04 2.755E+04 4.71 3E+04 3.070E+02 0.000E+00 3.735E+02 5.339E+04 0.000E+00 5.341E+'04 1.601E+03 0.000E+00 1.581E+03 1.915E+04 0.000E+00 2.165E+04 1.104E+04 0.000E+00 7.681E+02 4.873E+04 0.000E+00 1.142E+00 4.579E+04 0.000E+00 4.256E+-03 0.000E+00 0.000E+00 5.404E-07 0.000E+00 0.000E+00 1.204E-02 0.000E+00 0.000E+00 9.922E+03 0.000E+00 0.000E+00 1.204E-02 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME RB -87 CS-135 ACTIVITY (CI)I1.695969E-11I 5.403755E-07 WHAT IS THE SOURCE VOLUME (CC)4. 142000E+09 WHAT IS THE SOURCE DENSITY (GMICC)1 .530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.

SPiant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 15 of 19 I HE NUMBER. OF ENERGY GROUPS INPU I 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 2 3 4 5 6 7 8 9 10 3.000000E-01 4.000000E-01 5.000000E-01 6.000000OE-01 8.000000E-01 1 .000000E+00 1 .500000E+00

2. 000000E+00 3.000000E+00

4. 000000E+00 MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO.UC/GM SOURCE STRENGTH UC/CC Cl KR 83M KR 85 KR 85M KR 87 KR 88 KR 89 KR 89 RB 88 RB 89 SR 89 45 47 46 48 49 50 51 67 68 21 4 5 6 7 8 I I I I I 131 132 133 134 135 3.634336E+0 6.044329E+01 9.450985E+0 1.220921 E+03 2.321087E+03 7.574059E-03 7.574059E-03 2.274495E+03 6.320660E+01 1 .385227E-01 4.333311E+03 4.622957E+03 8.527853E+03 4.347433E+03 7.436683E+03 5.89351 9E+C 2.494256E+IC 1 .212008E+C 1.801554E-01 1 .900384E-03

1.

1 .900300E-0:

2 5.560535E-01 9.247824E-02 2 1.446001E+0C 1 .868009E+00 3.551264E+00 1.158831E-05 1.158831E-05 3.479977E+00 9.670610OE-02 2.11 9398E-04 6.629966E+00 7.073123E+00 1.304762E+01 6.651 572E+00 1.137812E+01 I1 9.017084E-02 I 1.289410E+01 132 3.816212E-01 I 5.227550E+00 132 1.854373E-01

2. 756378E-04 S1.027530E+00

2. 907587E-06 I 2.395366E+00 3 2.907459E-06 2.303174E+03

3. 830449E+02

)5.989335E+03 7.737294E+03 1 .470933E+04 4.799878E-02 4.799878E-02 1 .441406E+04 4.005567E+02 8.778545E-01 2.7461 32E+04 2. 929688E+04 5.404322E+04 2.755081 E+04 4.7128 19E+04 5.340735E+04 1 .580675E+03

2. 165251 E+04 7.68081 3E+02 1. 141692E+00 4.256031 E+03 1 .204322E-02 9.921605E+03 1 .204269E-02 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 137 35 CS 138 36 BAI137M 78 Piant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Ttle: NEI 99-01 Rev 6 EAL Calculations l6 of 19 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)I 2 3 4 5 6 7 8 9 10 3.000000E-01 4.000000E-01 5.000000E-01 6.000000E-01 8.000000OE-01 1 .000000E+00 1 .500000E+00 2.000000E+00 3.000000E+00 4.000000E+-00

9. 166370E+04 1.001 597E+I05 3.621505E+04 2.766909E+05 4.763061 E+05 4.956754E+05

7. 378778E+05 3.999907E+05 3.3536 18E+05 0.000000E+00 1 .265570E+

15 1. 037153E+ 15 3.000054E+

14 1. 91 0089E+15 2.466075E+-15

2. 053088E+ 15 2. 037527E+ 15 8.283807E+

14 4.630229E+14 0.000000E+00 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C 'Title: NEI 99-01 Rev 6 EAL Calculations 17 of 19 I Heere Abeet Us Products & Service. -teduetnes

& AapOce4less.

-TrelsIna flhesuras Yes, wrettere:

tltn I > Ta p. Sub Saturae Water Regie.Sub Saturated Water Region -Steam Table Atany pressare water belowissaturation tem atre issaid tobelsna sub saturated state Fur example. wate, at a iressure aft1 atmosphere aod temperature below Ohe saturated tempsteratur'e at iS sub saturated Water at a pressare sf 10 atutospheres has a saturation temperature of 1l0"¢, antd so mater betow Ohis terepeatare is also sub saturated Learn mere about steam et our tutorial -Wf.atItoomLL Set your ofioots fur" these steam tables Note: -You cannot use commas (,) as decimal points.Please use periods (.)Example: 1.02 not 1,02!Feature* Tratutt Out lt en~*tcfsaeta 0 lnputs Pressure Temperature Pressure and Temperature El~ StogIe Value Table put abratute El El E~fl Z~Z Z~Z Vapout Pressure Saturation Teomperature Specdflc Enthralpy of Watet (ho)D~ensiy of WaterVolume of Water (o)Specdl Envopy of Water (sr)trith JflcgK El El I, Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Atcmn Title: NEI 99-0 1 Rev 6 EAL Calculations 18tchofnt19, itl n lsite for Spir you ace ttere(: t Etem5 .I S~perthnated Steam Resists Superheated Steam Region -Steam Table The superheated steam regon depics steam at a temperature trigter than its saturation temprature Should saturated steam he heated at constant pressure, its temperature wit rise. producing sarperheated steam Learn mere about steam In our tutorial -Wa sSem Set your toogt for these steam tablesI Note: -YOU caritot use commas (,} as decimal points.Please use periods (0 Example: 1.02 not 1,02 Output t Sigl Vatue Table Supertheat Temperature F r-Pv Feature.L Treater 0 atofe Saturation Temperature Degrees Superheat SpctcEnstralpy at Water (Irs)Specific Enthalpy ottEnaporation (N)Specitic Enttalpy of Superhseated Steam (h)Density of Steam Specific Vohene of Steam (s)Jag JAro jag 555 Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 19 of 19 Table 6.1.-I. Units used to express S tabulated in relation to the units used to express________ _______S__________

Point source Line source Surface. Volume o0 as a factor of to. S SOUrce source S 0 S. S S y-photons em 2 s?-photons cm~s MeV fim 2 s MeV cm~s MeV cm 2 s rads h rads h R h R R R?'-photons mCi (milhicuries) m$mCi V-photons MeV V-photons MeV mCi rag-equiv, Ra y-photons cms mCi cm MeV oms mci cm mCi cm V-photons¢ms MeV cins V-photons cms MeV cms mCi cm y,-photons cm 2 s mCi cm 2 MeV cm2S mCi cm 2 mCi cm 2 y-photons cm's MeV cm 2 s v-photons cm 2 s MeV cm~s mCi y-photons cm 8 s mCi MeV cm8$mCi cm 8 mCi V-photons M~eY cmas y-photons cm3 s MleV mCi.70 disintegrations v-photons 3.1~smCi ndisintegration 3,7" 10'disintegrations V -photons MeV smCi disintegration ,gphoton 4?MeV 4.Jcmn~smCi Me V 1.6 ~ ergs _ cm 2 s 100 ergsl(g~rad) 1 610 ergs cm 2 s 100 ergs/(g, rad)eV 1.6io ergs _cm2 60 87.7 ergs/(gR)c rs m 2 s M.. e- -V g .-g- Iio B"7.7 ergsl (gR)Rcm 2 43r. Kr hmCi Rcm 2 4~84hmg-equiv.Ra mg-equiv.

Ra mg-equiv.

Ra Img-equiv.

Rn cm cm 1 cm 3 Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment Dl Title: NEI 99-01 Rev 6 EAL Calculations 1 of 11j Purpose: The purpose of this appendix is to calculate radiation readings for Other Indications (Fuel Clad Barrier Potential Loss 5.A).a. Off-Gas Post Treatment Radiation Monitor D11K615 Procedure 64C1-OCB-004-1/2.

b. Fission Product Monitor D11K630 Procedure 64C1-OCB-005-1/2.

Defining Units that MathCAD will understand MWt := 10 6-W I Ci:.- *Bq and 2.7.10 Rad := 0.01iGy, Per FGR-11 page 219 IItCi := 10-6 Ci Rem := 0.01Sv Rem mRem :=-1000 Per FGR-11 page 219 cc := mL 1 eps := -sec 1 cpm :=-min The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used with the post treatment radiation monitors.

These isotopes are listed in 64CI-OCB-004-1/2 Page 56.Isotopes: "Ar-41I""Kr-85m""Kr-85""Kr-87""Kr-88""Kr-89""Kr-90""Xe-13 lm"'"Xe-133m""Xe- 133""Xe-135m""Xe-135""Xe-137" S"Xe-138" Core._Inventory

0 3.30E+03 3 .78E+02 6.92E+03 1 .32E+04 1 .86E+04 2.26E+04 3.03E+02 5.27E+04 1.58E+03 1 .89E+04 1 .09E+04 4.81E+04 4.52E+04 Ci MWt Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment 0 Title: NEI 99-01 Rev 6 EAL Calculations 2 of 11 CorePower

= 2818.MWt RCS inv := 9965 ft 3 F Release := 0.05 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.To determine the RCS inventory mass the water specific volume needs to found as follows: Ul: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4°F+/-+

532.0 °F TAvgu 1 := 2 462.2.OF U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 °F + 535.0 °F T Avgu 2 := 2= 480.35.°F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 URCS water := 0.0199161-

-Ibm According to http :/l/www. s piraxs arco. c om/res ources /st eam-tab les.asp the specific volume for 106Opsia and 480.35 F conditions is 0.0199161ft^3/Ibm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv 18.g RCS mass .- -2.27 x 1 i VRCS water Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 3 of 11 The concentration of isotopes in RCS is calculated as follows: FDEI300 .- gm -0.0 13-~2.38-104 gm This fraction was determined in Attachment C.This is the fraction that will be applied isotopes concentrations to adjust to RCS DEl 300 uCi/gm equivalent as described in NEI 99-01.CoreInventory-CorePower.

F_Release.

F_DEI3O00 Conc_DEI 3 0 0 RCS isotopes : RCS mass augment(Isotopes, Conc_DEJ300 RCS_isotopes)=

o: "Ar-41" 0E+000 i. "Kr-85m" 2.582E+001 2 "Kr-85" 2.958E+000

-3 "Kr-87" 5.415E+001 14 "Kr-88" 1.033E+002 S51 "Kr-89" 1.456E+002 6 "'Kr-90" 1.769E+002 7 "Xe-131m" 2.371E+000 8 "Xe-133m" 4.124E+002

9. "Xe-133" 1.236E+001 10, "Xe-135m" 1.479E+002 1i1 "Xe-135" 8.53E+001 12 "Xe-137" 3.764E+002

'13 "Xe-138" 3.537E+002 I-Ci gm

[Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 4 of 11 RCS_mass.

Conc_DEI 3 0 0 RCS isotopes Rateoff Gas*: 1lhr The gaseous activity corresponding to 300uCi/gm DEl would evolve from the RCS water to main steam. Assuming dilution in one hour of main steam flow and release through the offgas.augment(Isotopes, Rateoff Gas)o 1 0! "Ar-41" OE+000 1 "Kr-85m" 1.628E+006 2 "Kr-85" 1.865E+005 3 "Kr-87" 3.414E+006 4 "Kr-88" 6.512E+006 5 "Kr-89" 9.176E+006 6 "Kr-90" 1.115E+007 7 "Xe-131lm" 1.495E+005 8 "Xe-133m" 2.6E+007 9 "Xe-133" 7.795E+005 10 "Xe-135m" 9.324E+006 11 "Xe-135" 5.377E+006 12 "Xe-137" 2.373E+007 13 "Xe-138" 2.23E+007 IltCi sec Total-RateoffGas

= Rateoff -a =1.197 >< 108. tC-- sec Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 5 of 11 RadM~nDose Factors : (7.34E-4'8.25E-5 1 .26E-6 4.40E-4 1 .09E-3 9.44E-4 7.00E-4 1 .68E-6 I1.29E-5 1.3 8E-5 2.42E-4 1.33E-4 9.55E-5 ,6.16E-4, The radiation monitor Dose Factors were obtained from 64CI-OCB-004-1/2 Page 56.toRemo sec plCi~yr DoseRate := (RadMOnDoseFactors.RateoffGas))

augment(Isotopes, Dose Rate) =0 1 0 "Ar-41" 0E+000 1 "Kr-85m" 1.343E+002 2 "Kr-85" 2.35E-001 3 "Kr-87" 1.502E+003 4 "Kr-~88" 7.098E+003 5 "Kr-89" 8.662E+003 6 "Kr-90" 7.805E+i003 7 "Xe-131m" 2.511E-001 8 "Xe-133m" 3.354E+002 9 "Xe-133" 1.076E+001 10 "Xe-135m" 2.256E+003 11 "Xe- 135" 7.152E+002 12 "Xe-137" 2.266E+003 13 "Xe-138" 1.374E+004 mRem yr Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 6 of 11 Dose_RatetotaI

= DoseRate Summing all rows of the Dose_Rate matrix.Dose atetotal ; 4.452 x 0 ' em Calculating the ratio of this release to ODCM allowed release: LimitODCM

=500 m-- The ODOM limit is listed in ODCM Section 3.1.2 yr and in 64CI-OCB-004-1/2 Page 56.Factorallocation

= 0.225 The factor is listed in 64Cl-OCB-004-1/2 Page 56.Factorsafety

= 0.5 The factor is listed in 64C1-OCB-004-1/2 Page 56.DoseRate total RatioReleaseODCM LiiOC.Fatrioato=sft

-LimtODCM FactralloationFactorsft LRatioReleaseODCM

= 791.516J Calculating the detector response as follows: Total-RateoffGas

=1.197 x 108 C-- sec 14 cc Flowofa := 200cfm =9.439 x 10 -Ofgssee Factorefficiency

= 3.8.105 p liCi cc calculated above.The efficiency factor is approximately 3.ESEcps/uCi/cc per attached below e-mail.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 7 of 11 TotalRateoff Gas" Factoreicny Reading~cs DEI300 : Flowoffgas Reading~cs DEI300 =4.82 x 108.cp]According to S$16039 pg 21 and SX27520 pg 14 the range of this instrument is 1E-1 to 1E6 cps. The reading will exceed the range of this monitor, but a reading at the high end of the monitor range will still be indicative of fuel damage.Part B of FCEAL4 The fission product monitor for the containment air is DilP010 per SX18062 pg 32 and SX27520 pg 32. The iodine-noble gas sample panel is designed to extract a representative sample of gaseous and particulate effluent, transport the sample to: the iodine monitor where the iodine is retained on a filter for 1-131 gamma radiation detection and the gas sampler where it is continuously monitored for gross gamma and beta radioactivity (SX29455 section 6, S19207 section 7, and $30523 section 1-5). Since Iodine's and Noble-Gases are separately measured, only noble-gases will be evaluated in this attachment.

Also, noble-gases provide real time indication.

gal Leakage~wHP

= 3900---day gal Leakage~wLP

= 700 --day The High Purity Waste Stream leakage inside the DryWell is obtained from SMNH-93-029 page 6.The Low Purity Waste Stream leakage inside the DryWell is obtained from SMNH-93-029 page 6.Leakage~wtot

=Leakage~wHP

+ Leakage 0 w LP =4.6 x .0 3 ga-_ day DWVolu1 := 146010ft 3 SPVOlu 1:= 112900ft 3 The dry well free volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol_Containmentu1

= DW_Voluj 1 + SPVolu1 = 2.589 x 105.ft3 3The dry well free volume is listed in NL-06-1 637 Enclosure DW-V°Iu 2:= 146266ft3 1 page 47 table 34.

SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13--021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 8 of 11 SP_VoIu 2:= 109800ft3 Sne suppression pooi voiume Is iistea in INL-UO-Ilbi(Enclosure 1 page 47 table 34.Vol_Containmentu 2:= DWVolu 2+ SPVolu 2=2.561 >( 105.ft3 ContainmentVol Avg : VolContainm entul + Vol Containmentu 2 53= 2.575 x 105.ft3 2 ft 3 a~water STP :=0.016714-

-Ibm Specific volume at STP per ASME international Steam Tables page 174 1 0.958. gm PwaterSTP

"- teTPm The isotopes concentration in the containment (DryWell and Suppression Pool) after the 1st hour is calculated as follows: C_IsotopeScontainment

1 hr Leakage~w -tot" Pwaterona met STp 'Conc -DEI 3 0 0 RCSvo v isotopes i augment(Isotopes, CIsotopeScontainment)

=0 0 "Ar-41" 0 1 "Kr-85m"'

2.463"10-3 2 "Kr-85" 2.821"10-4 3 "Kr-87" 5.164"10-3 4 "Kr-88" 9.851"10-3 5 "Kr-89" 0.014 6 "Kr-90" 0.017 7 "Xe-131m" 2.261"10-4 8 "Xe-133m" 0.039 9 "Xe- 133" 1.179" 0-3 10 "Xe-135m" 0.014 12 "Xe-135" 0.13"0-3 13 "Xe-137" 0.034 mL Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 9 of 11 F Responsexel 3 3 :=~ 2.7.10 cpm mL F ResponseKr 8 5 := 2.29-10o8 cpm liCi mL The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number P0l0.The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number Polo.Since the D11P010 gas detector is G-M counter ($30523 pdf page 76 section 1-4) the Xe133 response factor will be applied to all isotopes of Xenon, while the Kr85 response factor will be applied to all isotopes of Krypton. Because GM will detect all gammas as long as they are above the threshold for the detector.Isotopes =K "Ar-4 1""Kr-85Sm""Kr-8 5""Kr-87""Kr-8 8""Kr-89""Kr-90""Xe-131lm""Xe-133m""Xe-133""Xe-135m""Xe- 135""Xe-137" S"Xe-138" 0 FResponseKr 8 5 F_ResponseKr 8 5 F_ResponseKr 8 5 F_ResponseKr 8 5 F_ReSPOnseCKr 8 5 F_ResponseKr 8 5 N FResponse

F_Responsexe1 3 3 FResponsexe1 3 3 FResponsexe 1 3 3 F_Responsexel 3 3 F_Responsexe1 3 3 I, FResponsexel 3 3 F Responsexe1 3 3)

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment ID Title: NEI 99-01 Rev 6 EAL Calculations 1l of 11 The response of the detector D11P010 is calculated as follows: the response factor for the detector is multiplied by the concentration of isotopes in the containment.

Responsep 0 1 0:= (FResponse.CIsotopeScontainment) augment(Isotopes, Responsep 0 1 0)=0 1 0 "'Ar-41" 0 1 "Kr-85m" 5.64"105 2 "Kr-85" 6.46"104 3 "Kr-87" 1.183.106 4 "Kr-88" 2.256" 106 5 "Kr-89" 3.179"106 6 "Kr-90" 3.862.106 -cpm 7 "Xe-131lm" 6.106"103 8 "Xe-133m" 1.062" 106 9 "Xe-133" 3.184"104 10 "Xe-135m" 3.808"105 11 "Xe-135" 2.196'105 12 "Xe-137" 9.692'105 13 "Xe-138" 9.108'105 Tot_Responsep 0 1 0:= 2Responsep 0 1 0=1.469 x This exceeds the range of the monitor 10 to 10^6 cpm. The reading at the high end of the monitor range will still be indicative of fuel damage.U2: Per SX29455 Table 6-1 the recorder for P010 is D11-R630.

Per SX27520 page 31 the recorder range is 10 -1 0^6 cpm.UI: Per $19207 Table 7-1 the recorder is MPL D11-R630.

Per SX18062 page 31 the recorded data sheet is A16462 sheet D11B. Per this data sheet the recorder range 10-10^6 cpm.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 11 of 11 From: Hostetter, Dwight A.Sent: Thursday, March 31, 2005 8:50 AM To: Wehrenberg, James A.

Subject:

Efficiency Factors Andy, Hatch Chemistry gave me the following efficiency factors for the instruments listed: Plant Service Water Monitor (Procedure 64C1-OCB-008)

U-i: 1.78 E+7 cps/uCi/ml U-2:1I.53E+7 cps/uCi/mI Liquid Radwaste Monitor (Procedure 64CI-OCB-009)

U-i: 3.136 E+6 cpsiuClIml U-2: 4.84 E+6 cps/uClfml Dwight Offgas Post-treatment .Monitor Response Factor From: Hodgins, S. R.Sent: Tuesday, April 26, 2005 2:01 PM To: Wehrenberg, James A.

Subject:

RE: Calibration set points Sorry for the incomplete information.

The number is alilthat is listed on our set point record. In 1994 we contracted Ultrapure Watel: Technology Inc to determine efficiency factors for our process radiation monitors.

We.do not verify these subsequent to the initial mock up determinations that they did then. We do 24 month source checks to verify that monitor response has not degraded appreciably., went to the Ultrapure report and the unit for the efficiency factor Is cps/(uci/cc).

---

Original Message-From: Hostetter, Dwight A.Sent: Friday, April 22, 2006 2:19 PM To: Hodgins, S. R.; Wehrenberg, James A.

Subject:

RE: Calibration set points Thanks Steve.Dwight From: Hodgins, S. R.Sent: Friday, April 22, 2005 2:17 PM To: Hostetter, Dwight A.

Subject:

RE: Calibration set points To my knowledge we do not determine an~eff fac for FPM NG monitors.

For most of our monitors, we correlate release rates to monitor response by comparing isotopic analysis of a process sample to monitor response.

The FPM noble gas samples we pull typically are clean (no activity).

Unit I Post Treat A and B =3.85E5: Unit 2 Post Treat A and B = 3.75E5.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 1 of 16 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors (ROS Barrier Loss Threshold 4.A). This attachment is some what repetitive of FC EAL3 (Attachment C) except of the DEI131 here is equal to TS instead of DEII31 of 300uCi/gin.Defining Units that MathCAD will understand MWt := 106-W Mega Watt Thermal is same as watts.Ci: .-*Bq 2.7.10 1 and Rad := 0.01Gy Per FGR-11 page 219:= 10-6 Ci The source terms for noble gases and iodine's are provided in NL-06-1637 Enclosure1 page 16 and are shown below.Isotopes : "Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 1m""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-13 8""I-131""1-132""1-13 3""1-13 4""I-135" Corelnventory

(3.30OE+03

" 3 .78E+02 6.92E+03 I1.32E+04 1 .86E+04 2.26E+04 3.03E+02 5 .27E+04 1 .58E+03 I1.89E+04 1.09E+04 4.81IE+04 4.5 2E+04 2.72E+04 3 .93E+04 5 .52E+04 6.05E+04 5.16E+04 Ci MWt 9 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 2 of 16 CorePower

= 2818-MWt RCS inv := 9965ft3 FRelease := 0.05 DWVolu1 := 146010ft3 The core power is provided in NL-06-1 637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1 637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.SP_Vo1u 1 := 112900ft3 The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol_Containmentui

= DWVOl~uI + SPVolu1 2.589 x 10 5.ft 3 3 The dry well free volume is listed in NL-06-1 637 DW-V°Iu 2 := 146266ft3 Enclosure 1 page 47 table 34.SP_Volu 2 := 109800ft3 The suppression pool volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.Vol_Containmentu 2 := DW_Volu 2 + SPVolu 2 =2.56 1 x< 105.ft 3 VolContainmentul

+ Vol-Containmentu 2_3 ContainmentVol Avg := 2.575 x lO-.ft-2 To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 0 F TAvguI : = 462.2.OF U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 0 F T Avgu 2:= = 480.35.°F Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 3 of 16 The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 V°RCS water := 0.0199161-

-Ibm According to htt p://www.s piraxs arco. com/resources/steam-tab les.asp the specific volume for 1060psia and.480.35 F conditions is 0.0199161ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv8 RCS mass >- --2.27 x 10 8gin"-RCS water The concentration of isotopes in RCS is calculated as follows: CoreInventory.

CorePower.

F_Release COnCRCS isotopes : RCS mass augment(lsotopes, COnCRCS isotopes)0 1 0 "Kr-83m" 2.049'10-3 1 "Kr-85" 2.347" 0-4 2 "Kr-85m" 4.296"10-3 3 "Kr-87" 8.195"10-3 4 "Kr-88" 0.012 5 " T Kr-89" 0.014 6 "Xe-131lm" 1.881"10-7 "Xe-133" 0.033 8 'Xe-133m" 9.809"10-9 "Xe-135" 0.012 10 "Xe-135m" 6,767'10-3 11 "Xe-137" 0.03 12 "Xe-138" 0.028 13 "I-131" 0.017 14 "1-132" 0.024 15 "1-133" 0.034 16 "I-134" 0.038 17 "I-135" 0.032 Ci gin tPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEl 99-0 1 Rev 6 FAL Calculations 4 of 16 The DEl for the RCS is calculated in Attachment C and was found to be 2.37E4uCi/gm.

gm___ -5 F DEJ TS.- g -8.439xx10

_ _ 2"714 I-tCi gm Per NEI 99-01 RCS Barrier EAL4 RCS should be of TS limit. According to HNP Tech Specs section 3.4.6 the DEIl31 limit is 2uCi/gm Amount of isotopes in the drywell is found as follows: In this calculation the core inventory is multiplied by core power, containment release fraction, and the DEl of 300uCi/gm fraction.

The amount of total curies released are distributed throughout the drywell and suppression pool. The concentration then is multiplied by the U1 DryWell volume. This way we have the total curies inside the DryWell only. The DryWell for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.Corelnventory.

CorePower.

F_Release.

F_DEl_TS-D WVolu 2 Isotopes_in_DW

augment(Isotopes, Isotopes inDW) =ContainmentVol Avg 0 h 1_ _"Kr-83m" 2.229E+001"Kr-85" 2,553E+000"Kr-85m" 4.674E+001"Kr-87" 8.916E+001"Kr-88" 1.256E+002"Kr-89" 1.526E+002"Xe-131m" 2.047E+000"Xe-133" 3.56E+002"Xe- 133mi" 1,067E+001"Xe-135" 1.277E+002"Xe-135m" 7.362E+001.Ci"Xe-137" 3.249E+002"Xe-138" 3.053E+002 "1-131" 1,837E+002 "1-132" 2.654E+002 "1-133" 3.728E+002"I-134" 4.086E+002"I-135" 3.485E+002 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 5 of 16 The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air density for pressure 44.7psia and temperature of 343F (NL-06-1637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

DW_Volu 2 = 4.142 x 109-mL Psteam'.-

-1.53 x10 , ft 3 mL 10.4690 --lb According to http://www.spiraxsarco.

corn/resources/steam-tab les.asp the specific volume for 44.7psia and 343 F conditions is 10.4690ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below. The drywell pressure and temperature were obtained from NL-06-1637 Enclosure 1 table 11 and table 12.The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.ergs -cm 2 3 0 S eVg h 100 ergs/(g .rad)2-6 ergs cm s 1.6.10 --.-~ 30 Rad_hr or~GeometriCfactor-Sv 100 ergs g. rad-6 1.6-10 x 3600 -5= 5.76 x 10 100 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEl 99-01 Rev 6 EAL Calculations 6 of 16 Defining Units that MathCAD can process.q:=1.60219.10-19coul eV := qe.volt MeV := 106.eV Fundamental charge of electron page 6 of Fundamental of Nuclear Science and Engineering.

defines the unit of charge.cc := mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

3 .00E-0 1 4.00E-0 1 5.00E-01 6.00E-0I1 8.00E-01 1.00E+00 1 .50E+00 2.00E+00 3.00E+00 4.00E+00 SMeVSv: r6.11E+02" 6.6 8E+02 2.41IE+02 1 .84E+03 3.18E+03 3 .30E+03 4.92E+03 2.67E+03 2.24E+03 0.00E+00 MeV cc.SeC Iabsair : 0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.0238 0.02 11194 2 cm gm Heightcyl

= 1.265.10J cm Calculated in Attachment C Calculated in Attachment C Radiationfield

SHighty, 1 (P.labsairlj 0 1 2 3 4 5 6 7 8 9 0 0.324 0.364 0.132 1.003 1.693 1.702 2.32 1.171 0.871 0 Rad hr The unit conversion was handled by MathCAD internally.

This unit conversion was explained above in this attachment.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NE! 99-01 Rev 6 EAL Calculations 7 of 16 SMRadiation field := £Radiationfil

= Rad-iedhr Radiationleak

= SUMRadiationfield

= 9.581- a-hr According to the $43177 Operator Manual Figure 1-4 the model number 877-1 detector has 1 to 1 ratio of Radiation present to radiation measures or shown. Therefore the detector would read 9.58 1 R/hr. However the detector is located in the area where some radiation is present during normal operations.

The following calculation will determine this contribution.

Per H16241 and H26417 the N003 detector location is at elevation of 156ft and approximately 27 ft from the center of RPV. According to (BH2-M-V999-0047 Table 2 and 3) the Solenoid Valves B21-AOV-F013 are located in approximately at elevation of 153.6 and radius between 17 to 22 feet.The radiation field from RPV at elevation 153.6 for Unit 1 and Unit 2 is listed below (BH2-M-V999-0047 Table 2 and 3).Solenoid Valves Solenoid Valves for U1 Rad/hr for U2 Rdh 1B21-AOV-F013A 6.44 2B21-AOV-F013A 2.46__1B21-AOV-F013B 4.41 2B21-AOV-F0138 9.96__1B21-AOV-FO13C 12.3 2B21-AOV-F013C 9.96__1B21-AOV-F013D 13 2B21-AOV-F013D 24 1B21-AOV-F013E 12.3 2B21-AOV-F013E 2.46 1B21-AOV-F013F 12.3 2B21-AOV-F013F 9.96 1B21-AOV-F013G 12.3 2B21-AOV-F013G 9.96 I1 B21-AOV-F013H 4.41 2B21-AOV-F013H 2.46 1B21-AOV-F013J 4.41 2B21-AOV-F013K 9.95 1821-AOV-F013K 12.3 2B21-AOV-F013L 9.96 I1B21-AOV-F013L 12.3 2B21-AOV-F013M 9.96 Ave rage = 9.7 _ ______ 7.2 9.7 + 7.2 Rad _85.Rad Tkn naeaebtentouis Rdainpv 2 hr hr Plant: H-NP UI & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 8 of 16 The radiation field from Main Steam Lines (BH2-M-V999-0047 Table 6 and 6).Solenoid Valves Solenoid Valves for Ul ,Rad/hr for U2 Rad/hr 1B21-AOV-F013A 17.57 2B21-AOV-F013A 17.39 1B21-AOV-F013B 17.28' 2B21-AOV-F013B 17.28 1B21-AOV-F013C 17.08 2B21-AOV-F013C 17.08 1B21-AOV-F013D 17.01 2B21-AOV-F013D 17.39 1B21-AOV-F013E 17.16 2B21-AOV-F013E 17.28 1B21-AOV-F013F 16.13 2B21-AOV-F013F 17.08 1B21-AOV-F013G 16.13 2B21-AOV-F013G 15.43 1B21-AOV-F013H 16.62 2B21-AOV-F013H 17.28 1B21-AOV-FO13J 16.33 2B21-AOV-FO13K 15.43 1B21-AOV-F013K 16 2B21-AOV-F013L 14.78 1B21-AOV-F013L 13.4 2B21-AOV-F013M 14.78 Average = 16.4 _________

16.5__6.4__+ 6. Rad 16.45- a Taking an average between two units.RadiationMsL 2.h-h The radiation field from Recirculation Lines ((BH2-M-V999-0047 Table 9 and 10).Solenoid Valves Solenoid Valves for Ut Rad/hr for U2 Rad/hr 1B21-AOV-F013A 7.5 2B21-AOV-F013A 11.97 1B21-AOV-F013B 8.23 2B21-AOV-FO13B 6.23 1B21-AOV-F013C 6.19 2B21-AOV-F013C 6.85 1B21-AOV-F013D 5.2 2B21-AOV-F013D 10.08 1B21-AOV-F013E 5.89 2B21-AOV-F013E 15.22 1B21-AOV-F013F 5.15 2821-AOV-F013F 6.19 1B21-AOV-F013G 5.89 2B21-AOV-F013G 6.85 1B21-AOV-F013H 7.5 2B21-AOV-F013H 15.96 1B21-AOV-F013J 7.5 2B21-AOV-FO13 K 6.85 1B21-AOV-F013K 6.22 2B21-AOV-F013L 7.8 1B21-AOV-F013L 6.22 2821-AOV-F013M 7.8 Average = 6.5 9.3 6.5 + 9.3 Rad Rad RadiationRL, .- -7..2 hr hr Taking an average between two units.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NE! 99-01 Rev 6 EAL Calculations 9 of 16 2 RadiationtotaI

= Radiationleak

+ Radiation~ppV

+ RadiationMsL

+ RadiationRL, = 0.763 --.-R ea K* s2 hr Rad Radiationtotai 42.381 hr This can be rounded to 40R/hr. The range of this instrument is 1-10^7 R/hr (established in attachment C).GRODEC Input 1 18 60,2.229e+

1,0 61 ,2.553e+0,0 62,4.674e+1

,0 63,8.916e+1

,0 64,1 .256e+2,0 65,1 .526e+2,0 141,1.837e+2,0 142,2.654e+2,0 143, 3.728e+2, 0 144,4.086e+2, 0 145,3.485e+2,0 146,2.047e+-0,0 147,3.560e+2,0 148, 1. 067e+- 1,0 149,1 .277e+2,0 150, 7. 362e+ 1,0 151,3.249e+2,0 152, 3. 053e+2,0 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E l Title: NE! 99-01 Rev 6 EAL 10oaf 16 Calculations OUTPUT OF GRODEC CALCULATION OPERATING IN MODE DATA FILE NAME:RCSEAL4 1 INPUT DATA LISTING ISOTOPE INITIAL ADDITION ACTIVITY(CI)

RATE(CI/HR)

KR-83M KR-85 KR-85M KR-87 KR-88 KR-89 1-13 1 1-132 1-133 I-134 1-135 XE-131M XE-133 XE-1 33M XE-135 XE-i135M XE-137 XE-i138 2.229E+01 2.553E+-00 4.674E+01 8.916E+01 1 .256E+02 1.526E+02 1 .837E+02 2. 654E+02 3.728E+02 4.086E+-02 3.485E+02 2. 047E+00 3.560E+02 1 .067E+01 1 .277E+02 7.362E+01 3. 249 E +02 3.053E+02 0.000E+00 0. 000E+00 0.000E+00 0.O00E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0. 000E+00 0.000E+00 0.000OE+00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

=1.000000 HOURS TOTAL ACTIVITY = 2155.917000 CURI ES ISOTOPE ISOTOPE NUMBER NAME INITIAL ADDITION ACT(CI) RATE(CI/HR)

ACTIVITY (CI)60 61 62 KR-83M KR-85 KR-85M 2.229E+01 2.553E+00 4. 674E+01 0.000E+00 1.536E+01 0.O0OE+00 2.553E+00 0.000E+00 3.993E+01 Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 11 of 16 63 KR-87 8.916E+01 0.OOOE+00 5.160E+01 64 KR-88 1.256E+02 0.000E+00 9.806E+01 65 KR-89 1.526E+02 0.000E+00 3.199E-04 75 RB-87 0.000E+00 0.000E+i00 1.131E-13 76 RB-88 0.000E+00 0.O00E+00 9.609E+01 77 RB-89 0.000E+00 0.O00E+00 2.669E+00 84 SR-89 0.000E+00 0.000E+00 5.850E-03 141 1-131 1.837E+i02 0.000E+00 1.830E+02 142 1-132 2.654E+02 0.000E+00 1.953E+02 143 1-133 3.728E+02 0.OO0E+00 3.603E+02 144 1-134 4.086E+02 0.000E+00 1.837E+02 145 1-135 3.485E+02 0.OOOE+00 3.142E+02 146 XE-131M 2.047E+00 O.000E+00 2.490E+O0 147 XE-133 3.560E+02 0.000E+00 3.561E+02 148 XE-133M 1.067E+01 0.00OE+00 1.053E+01 149 XE-135 1.277E+02 0.000E+00 1.444E+-02 150 XE-135M 7.362E+01 0.000E+00 5.122E+00 151 XE-137 3.249E+02 0.O00E+00 7.612E-03 152 XE-138 3.053E+02 0.000E+00 2.838E+01 161 CS-135 0.O00E+00 0.O00E+00 3.603E-09 163 CS-137 0.000E+00 0.000E+00 8.030E-05 164 CS-i138 0.000E+00 0.000E+00 6.615E+-01 171 BA-137M 0.000E+00 0.000E+00 8.029E-05 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME ACTIVITY (CI)RB-87 1.130985E-13 CS-I135 3.603327E-09 WHAT IS THE SOURCE VOLUME (CC)4.142000E+-09 WHAT IS THE SOURCE DENSITY (GM/CC)1 .530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.THE NUMBER OF ENERGY GROUPS INPUT: 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 3.0O0000E-01 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 12 of 16 Calculations 2 3 4 5 6 7 8 9 10 4.000000E-01 5.000000E-01 6.000000E-01 8.000000E-01 1 .000000E+00 1 .500000E+00

2. 000000E+00 3.000000E+00 4.000000E+00 MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO.UC/GM SOURCE STRENGTH UC/CC Cl KR 83M 45 KR 85 47 KR 85M 46 KR 87 48 KR 88 49 KR 89 50 KR 89 51 RB 88 67 RB 89 68 SR 89 21 I 131 4 I 132 5 I 133 6 I 134 7 I 135 8 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 137 35 CS 138 36 BAI137M 78 2.423253E+00 3.707578E-03 1 .535679E+01I 4.029027E-01 6.164411 E-04 2. 553299E+-00 6.300656E+00 9.640004E-03

3. 992890E+01I 8.141909E+00 1.245712E-02 5.159739E+01 1 .547392E+01 2.367509E-02

9. 806223E+01 5.047167E-05 7.722166E-08 3.198521 E-04 5.047167E-05 7.722166E-08 3.198521 E-04 1.516330E+01 2.319985E-02 9.609376E+01 4.211933E-01 6.444258E-04 2.669212E+00 9.23081 6E-04 1.41231 5E-06 5. 849808E-03 2.888350E+'01 4.419176E-02 1.830423E+02 3.081971E+01 4.715415E-02 1.953125E+02 5.685235E+01 8.698410OE-02 3.602881 E+02 2.898289E+01 4.434381E-02 1 .836721E+02 4.957314E+01 7.584691E-02 3.141579E+02 3.929409E-01 6.011 996E-04 2.4901 69E+00 5.619389E+01 8.597666E-02 3.561153E+02 1.66231 8E+00 2.543346E-03 1 .053454E+01 2.278258E+01 3.485735E-02 I1.443792E+02 8.082252E-01 I1.236585E-03 5.121 933E+00 1.2011 59E-03 1. 837774E-06 7.61 2058E-03 4.477746E+00 6.850951E-03 2.837664E+O1 1 .267052E-05 I .938590E-08 8.029641E-05 1.043846E+01 1.597085E-02 6.615126E+01 1 .266997E-05 1 .938505E-08 8.029286E-05 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)1 2 3.000000E-01 4.000000E-01

6. 111697E+02 6.676899E+02 8.438217E+12 6.913930E+12 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 13 of 16 SCalculations 3 4 5 6 7 8 9 10 5.000000E-01 6.000000E-01 8.000000E-01 1 .OQ0OO0E+00 1 .500000E+00 2.000000E+00 3.000000OE+00

4. 000000E+-00 2.414372E+02 1.844615E+03 3.175361 E+03 3.304496E+i03 4.91 8952E-+03 2.666496E+03 2.235936E+03

0. 000000E+-00 2.000066E+i12 1 .273399E+

13 1.644043E+

13 1.368722E+13

1. 358287E+ 13 5.5223 14E+ 12 3.087083E+-12 0.000000OE+00 SPlant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 14 of 16 Calculations

[Yes are here: h'4rSe WTs S Srib Saturaed Water Region Sub Saturated Water Region -Steam Table At an~y pressure, water belew its saturatesn temperature is said It be ir a sub Cronwa ea saturated 1t atrsasph~ers base sa turaton temperature at 180°C, and so water Irelowr the temperature is sub saturated em Learn mare abaut steam in aorr tutmriatl-What is Steam?eaueSat youre ptitae tor these steam tale~Note: -You cannot us commas (,) as decimal points.Trannr that,,. Please use perods (.)tartat mstrt Example: 1.02 not 1,02Single Vatue Table Temperature F EJ Temperature

-: ...... +::[Speciltc Votume of Watw+(v) : :::::: s:::: I,

[Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 AttachmentiteNE9-0Rv6EACaclios1 Ef1 I 1 5 o nIsefo prf 0.5 AI~set Us ., Products & tamtes *, teatmoles

& Awkatons -TraegM You me here: I, th I I t Sseerp ae SlIeaseReOel.

Superheated Steam Region -Steam Table The supeseied sleam region diepicts steam at a tempr~eature higher than Its satturataons temperatu re Shtould saturated steam be heated at cntn presue tu its tnperature wI rise. preoducktg superheater steam Learn more about steam is our tuora -Whi m tem Set your for these steam tabltes Note: -You cantnot sue commas (,} as dlecimal points.Please use petlods (.)Example: 1.02 not 1,02 Feaur Er L e t hat 0 Restore and S.serteat Teerperatore

• StirgIe Value Table yu absolute Pressur~e Supedtsat Temperature liii Saturation Temperaturesuperheat Spedific Enhls of Water (ho)SpcfcEnthalpy of Fvaporatlorr (hl Specific Enthalpy of Superheated Steam (h)O~eomst of Steam f[I JArg Specifi Volume ot Steam (v)

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021AtahetE Title: NEI 99-01 Rev 6 EAL Calculations AtaC6of161 Table 6.1.-I. Units used to express S tabulated in relation to the units used to express _________S Point source Line source Surface. Volume co as a factor of ~ o" S source source S 0 t.S Sv7 p-photons 7-photons V-photons y-photons y-photons cm 2 s s cms cm~s cm~s 7-photons mCi mCi mCi mCi ,70 disintegrations

.p-photons cm, s (millicuries) 7m cm 2 cm 2 7O s mCi disintegration MeY. MeV MeV MeV MeV cm~s s cms cm~s cmS s MeY ~ mCi mCi mCi 37 0 cm~s cm cmz cm* disintegrations, n--photons MeV smCi "disintegration photon MeV mCi mCi mCi MeV cms mCi cm mncm 2 z J msmCi EMeV -eg m3600-rads 7-photons 7-photons y-photons 7-photons 7-hoo M.eV0 2hg h s cms cm~s cm8 s 100 ergs/(g.rad)

Me MY .6l0~ ergs cm,3600 s rads MeV MeV MeVMV 1..1- ha h s cmns cm~s cm 3 s 100 ergsftg ,rad)EMeV _erscm"- s R 7-photons v-photons 7-photons 7-photons y-pho.ton

-.1a ' h60-hs cms cmas cm~s 87.7 ergs/(gR)1 6 06ergs -cm 2.00s R MeV MeV MeV MeV 1.e.1 &-eee.a---

h hs cmns cross cm~s 87.7 ergs! (gR)R mCi mCi mCi Rcm 2-- mCi cmct cm 3 4 ' R mg-equiv.

Ra mg-equiv.

Ra mg-eqluiv.

Ra .Rcm 2"--h- rag-equiv.

Ra cm cm 1 cm 3 h 8. ra g-equiv. Ra Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations I of 6 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Fission Product Monitoring sensors (RCS Barrier Loss Threshold 5.A). The normal RCS liquid activity is shown in FSAR Table 11.3-1. Noble gases are not typically retained in the RCS water, but are continuously released via offgas system. Offgas releases are shown in FSAR Table 11.3-1. Assuming 2% of the halogens are volatile per NUREG-0016 Table 2-4. According to the table the ratio of concentration in reactor steam to the concentration in reactor water for halogens is 0.015 or 1.5% we assumed a higher value of 2%. It is assumed that 100% of noble gases leaves the solution.

All other nuclides 0.1% are assumed to be airborne per table 2-4 NUREG-0016.

Fission Product Monitor D11K630 Procedure 64CI-OCB-005-1/2.

Defining Units that MathCAD will understand MWt:= 106.W Ci:= .Bq 2.7.10[LCi:= I0-6 Ci Per FGR-11 page 219 I cpm :=rai RCS inv := 9965ff3 The RCS inventory is page 47 Table 34.provided in NL-06-1637 Enclosure 1 DW Volu1: 146010ft 3 SP_Vo1u 1:= 112900ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.Vol Containmentu1

= DW_Volu1 + SPVolu1i 2.589 x 105.ft3 DWVolu 2 := 146266ft 3 SP_Vo1u 2:=109800ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 2 of6 Vol_Containmentu 2:=DW Volu 2+ SP_Volu 2= 2.56 1 x 105.ft 3 Containment Vol Avg : Vol_Containmentul

+ VolContainmentu 2 3= 2.575 x 105.ft3 2 To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 0 F TAvgu 1 := = 462.2.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 °F TAvgu 2 := -480.35. F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 Acc"-'RCS water := 0.0199161

--Ibm http480 web of tU atta The RCS mass is calculated as follows: RCS inv 108.i RCS mass .- -2.27 x 0 i'0 RCS water*ording to:Ilwww. spiraxsarco.

com/resources/steam-tab asp the specific volume for 1060psia and.35 F conditions is 0.0199161ftA3/Ibm.

This)site was validated in Attachment H. The copy he webpage and the provided information is ched to this attachment below.RCS mass = 5.003 x 105.Ibm Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 3 of 6 The concentration of isotopes in the containment is calculated as follows and using MathCAD internal conversion converted into cpm via fission response factor : Now we need to figure out noble gas concentration in the containment as follows: The noble gas isotopes Release is listed in FSAR Table 11.1-1: "Kr-83m""Kr-85m""Kr-8 5""Kr-87""Kr-8 8""Kr-89""Kr-9 0""Kr-9 1""Kr-92""Kr-93""Kr-94""Kr-95""Kr-97""Xe-13 lm""Xe-133m""Xe-133""Xe-135m""Xe-135""Xe-137""Xe- 138""Xe-139""Xe-140""Xe-141""Xe- 142""Xe- 143""Xe- 144" ReleaSeNoble

(3.40E+03" 6.10E+03 10 2.00E+04 2.00E+04 1 .30E+05 2.80E+05 3 .30E+05 3.30E+05 9.90E+04 2.30E+04 2.10E+03 1 .40E+01 1.50E+01 2.90E+02 8.20E+03 2.60E+04 2.20E+04 1 .50E+05 8.90E+04 2.80E+05 3.00E+05 2.40E+05 7.30E+04 1 .20E+04 5.60OE+02 IlCi s I The detector responses listed below were applied as follows: the Xe133 response was applied to all isotopes of xenon, while the Kr85 response factor was applied to all isotopes of krypton.Because GM will detect all gammas as long as they are above the threshold for the detector.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment 6Fl Title: NEI 99-01 Rev 6 EAL Calculations 4 of 6 FResponsexe1 3 3 := 2.7-107 -p mL FResponseKr 8 5 : 2.29. 108.p hmL NolIsotop 0"Kr-83m""Kr-85m""Kr-8 5""Kr-87""Kr-88""Kr-89""Kr-90""Kr-91""Kr-92""Kr-93""Kr-94""Kr-95""Kr-97""Xe-131mi""Xe-133m""Xe- 133'"Xe-135m""Xe- 135""Xe- 137""Xe- 138""Xe- 139""Xe- 140""Xe- 141""Xe- 142""Xe- 143""Xe- 144" The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number P0l0.The response factor was obtained from SX1 8062 pg32 and SX27520 pg 32 for equipment tag number P010.FResponseKr 8 5 FResponse Kr85 FResponseKr 8 5 FResponseKr85 FResponseKr 8 5 F ResponseKr 8 5 FResponseKr 8 5 FResponseKr 8 5 FResponseKr 8 5 F ResponseKr 8 5 F ResponseKr 8 5 F ResponseKr 8 5 FResponseKr 8 5 Responsefactor

F-Responsexel 3 3 F-Rep~nexe33 F-Responsexei 33 F-Rep~nsxe 33 FResponsexe1 3 3 F-Responsexe1 3 3 FRspos ee 33 F-Responsexe 133 F-Rep~nexe133 FResponsexel133 FResponsexe1 3 3 F-Responsexe l 3 F-Responsexe 1 3 3 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 5 of 6 FlOWcoreUl

= 78.5.106 hrm 7716 Ibm Fl°WcoreU2

= "7 " hr-The flows can be found in GE-NE-O000-0003-0634-01 Figures Ia & lb.Flowore~ + FowcoeU07 Ibm FlOWav :=F°cr~ lw~e2=7.775 x 10 ag2 hr The concentration of noble gases isotopes in the containment and activity is found as follows. In this equation RCS mass is divided by RCS flow rate gives time that can be multiplied by the release rate of noble gases. The resulting amount of isotopes is divided by the containment volume and multiplied by the detector response factor.iFe v( RCmssl,owavg .ReleaSeNobleRePnfao]

A~tV~tnole LContainmentVolAvg 0 "Kr-83m" 2.474"103 1 "Kr-85m" 4.439'103 2 "Kr-85" 7.276 3 "Kr-87" 1.455"104 14 "Kr-88" 1.455'104 5: "Kr-89" 9.459"104 61 "Kr-90" 2.037'105~

7 "Kr-91" 2.401"105 i8 "Kr-92" 2.401"105~

9 "Kr-93" 7.204" 104.10 "Kr-94" 1.674"104 12 "Kr-97" 10.187 1i3 "Xe-131m" 1.287 14 "Xe-133m" 24.879 15 "Xe-133" 703.479 augment(Noblelsotop, ActivitYnoble)

=*cpm 16"Xe-135m" 2.231"103 1=1 _____

Piant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 6 of 6 17"Xe- 135" l.887" 10: 18 "Xe-137" 1.287' 04 19 "Xe-138" 7.635"103 20 "Xe- 139" 2.402"104 21 "Xe-140" 2.574" 104 22 "Xe-141" 2.059"104 23~ "Xe-142" 6.263" 103"Xe-143" 1.029"103 25 "Xe- 144" 48.042 ActivitYnoble~tot

= Ac'tivitYnoble

=1.008 x 10 6.cpm~er SX18062 page 34 the monitor K630 range is 10 to 10A6 cpm. This reading of 1E6 cpm ~tthe upper limit of the detector scale. This reading may be difficult to distinguish from!he RCS Barrier Potential Loss 5.A in attachment D.RP5: No radiation monitors capable of indicating a potential loss of the RCS dentified.

we/

lPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G]Title: NEI 99-01 Rev 6 EAL Calculations 1 of 121 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors ( Primary Containment Barrier Potential Loss Threshold 4.A).DWRRM is D11K621 NMP-EP-110-GL02 pg 66.Defining Units that MathCAD will understand MWt := 106.W Mega Watt Thermal is same as watts.Ci :- *Bq 2.7.10 and Rad := 0.01Gy Per FGR-11 page 219 IJ.Ci := l0-6 Ci The source terms for noble gases and iodine's are provided in NL-06-1 637 and are shown below. We will only consider the isotopes that have release fraction above 0 as described in RG1.183 Table 1 and Table 5."1-129""I-130""I-131""I-132""I-133""I-134""1-135""I- 136""I-137""1-138""Kr-83m""Kr-8 5""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m" 1.23 E-03 I1.05E+03 2.72E+04 3 .93E+04 5,.52E+04 6.05E+04 5.1 6E+04 2.45E+04 2.39E+04 1.18E+04 3 .30E+03 3 .78E+02 6.92E+03 I .32E+04 1 .86E+04 2.26E+04 3 .03E+02 5 .27E+04 1 .58E+03 Isotopes : Corelnventory

Ci M tt Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 2 of 12"Xe-135m""Xe-137""Xe- 138""Rb-86""Rb-88""Rb-89""Rb-90""Cs- 134""Cs-134m""Cs-135""Cs- 136""Cs-137""Cs- 13 8""Cs- 139" 1 .09E+04 4.81 E+04 4.52E+I04 7.06E+01 1 .89E+04 2.42E+04 2.34E+04 6.83E+03 1 .65E+03 2.3 5E-02 2.18E+03 4.14E+03 5.02E+04 4.75E+04)I CorePower

= 2818. MWt The core power is Enclosure 1.provided in NL-06-1637 page 14 RCS inv := 9965ft 3 FRelease := 0.05 DWVoluI := 146010ft 3 SP_Vo1u 1:= 112900ft 3 The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases, halogens (i.e. Iodine's), and Alkali Metals is 0.05 per RG-1.183 page 13. GAP Release.The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.VolContainmentul

= DWVoIuI + SPVolu1 2.589 x 105.ft DW Volu 2 := 146266ft3 SP_Volu 2:= 109800ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.

Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 3 of 12 VolContainmentu 2:= DW Volu 2+ SPVolu 2=2.561 x 105.ft3 Containment_VolAvg

VolContainmentul

+ VolContainmentu 2 2.7 10ft 2 We find the amount of curies in the drywell as follows: FCore := 20%The 20% is the core fraction gap release as described in NEI 99-01 for the Containment Barrier.FRelease = 0.05 CorePower

= 2.8 18 x 10 3.MWt The concentration then is multiplied by the UI DryWell volume. This way we have the total curies inside the DryWell only. The DryVA~ll for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.DW_Volu 2 = 1.463 x 105.ft3 Containment Vol Av .7 0.t Coreinventory.

CorePower.F_Core.F_Release.

DWVolu 2 Isotopes in DW : Containment Vol Avg 0 1 0 "1-129' 1.969E-002 1 "I-130" 1.681E+004 2 "1-131" 4.354E+005 3 "1-132" 6.291E+005 4 "I-133" 8.836E+005 5 "I-134" 9.685E+005 6 "1-135" 8.26E+005 7 "1-136" 3.922E+005 8"I-137" 3.826E+005 B "1-137" 3.826E+005 I

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment~l:NI9-1Rv6ELCluain G] 1 FF r 9"I-138" 1.889E+005 augment(Isotopes, Isotopes in DW)10 "Kr-83m" 5.283E+004 11 "Kr-85" 6.051E+003 12: "Kr-85m" 1.108E+005 13 "Kr-87" 2.113E+005 P14 "Kr-88" 2.977E+005 15 "Kr-89" 3.618E+005 16 "Xe-131m" 4.85E+003 17, "Xe-133" 8.436E+005 18 "Xe-133m" 2.529E+004 19 "Xe-135" 3,025E+005 20 "Xe-135m" 1,745E+005 21 "Xe-137" 7.7E+005 22 "Xe-138" 7.235E+005 23 "Rb-86" 1.13E+003.24 "Rb-88" 3.025E+005 25 "Rb-89" 3.874E+005 26 "Rb-90" 3.746E+005 27- "Cs-134" 1.093E+005 28 "Cs-134m" 2.641E+004 29 "Cs-135" 3.762E-001 30 "Cs-136" 3.49E+004 31 "Cs-137" 6.627E+004 32 "Cs-138" 8.036E+005 33 "Cs-139" 7.604E+005.Ci The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air density for pressure 44.7psia and temperature of 343F (NL-06-1637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

The isotopes of 1-129, 1-136, 1-137, and 1-138 are excluded from GRODEC because they are not supported, i.e. these isotopes are not part of CALC F-86-03 (pdf pages 105-112) and thus have no way of making appropriate entrees.

Plant: HNP U1 & U2 SNC CALCULATION SMNH.-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 5 of 12 The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.1.6.0-6ergs 1.5.-V-cm 2 s* -3600 gT" " 100 ergs/(g .rad)2 161- 6 ergs cm ...s 1-.6.1-m'0°°° Rad-hr Geometricfactor.

Sv 100 ergs gr ad or-6 1.6.10- x3600 -5=5.76 x 10 100 Defining Units that MathCAD can process.q =1.60219.10-1 9-coul eV := qe-volt MeV := 106.eV Fundamental charge of electron page 6 of Fundamental of Nuclear Science and Engineering.

defines the unit of charge.co := mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

"3.00E-01

" 4.00E-0 1 5.00E-01I 6.00E-01 8.00E-01 1.00E+00 1.50E+00 2.00E+00 3 .00E+00 4.00E+O0**MeV Sv: r1.5 1 .64E+06 7.81 E+05 4.46E+06 9.75E+06 8. 12E+06 I1.50E+i07 6.42E+06 6.72E+06 5 .22E+00 MeV cc sec lx~absair

0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.023 8 0.02 11 0,.0194)2 cm gm Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 6 of 12 3 HeightcyI

= 1.265-103cm qo =.505 Calculated in Attachment C Calculated in Attachment C Radiationfield

/Sv~,.Hehtcy 1~P4 saii)0 1 2 3 4 5 6 7 8 9 0 801. 194 894.343 427.342 2.432"103 5.191"103 4.189" 103I 7.075" 103 2.815"103 2.612"103 1.866" 10-3 The unit conversion was handled by Rad MathCAD internally.

hr This unit conversion was explained above in this attachment.

SMRadiationf ied := ~'-"Radiationfil

= 2.644 x10Ra-c ie dhr~ccording to the $43 177 Operator Manual Figure 1-4 the model number 877-1 detector~is 1 to 1 ratio of Radiation present to radiation measures or shown. Therefore the~itector would read 2.644E4 R/hr. The range of this instrument is 1-10^7 R/hr established in attachment C).

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 7 of 12 GRODEC Input 1 30 60,5 .283e+4, 0 61 ,6.051e+3,0 62,1 .108e+5,0 63,2.11 3e+5,0 64,2.977e+-5,0 65,3.618e+5,0 73,1.130e+3,0 76,3.025e+5,0 77,3.874e+i5,0 78,3.746e+5,0 140,1.681e+4,0 141 ,4.354e+-5,0 142,6.291e+-5,0 143,8.836e+5,0 144,9.685e+5,0 145,8.260e+5,0 146,4.850e+3,0 147,8.436e+5,0 148,2.529e+I4,0 149,3.025e+5,0 150,1 .745e+5, 0 151 ,7.700e+5,0 152,7.235e+5,0 159,1.093e+5,0 160,2.641e+4,0 161 ,3.762e-1,0 162,3.490e+'4,0 163,6.627e+4,0 164,8. 036e+-5, 0 165,7.604e+5,0 OUTPUT OF GRODEC CALCULATION OPERATING IN MODE 1 DATA FILE NAME:BCEAL4 INPUT DATA LISTING I.qATAPI::

INIITIAI 13rllTI NNI SPlant: HNP Ul & U2 SNC CALCULATION SN-301Atcmn Title: NEI 99-01 Rev 6 EAL Calculations 8 of 12 ACTIVITY(CI)

RATE(CI/HR)

KR-83M 5.283E+04 0.000E+00 KR-85 6.051E+03 0.000E+'00 KR-85M 1.108E+05 0.000E+00 KR-87 2.113E+-05 0.000E+00 KR-88 2.977E+05 0.000E+00 KR-89 3.618E+05 0.000E+00 RB-86 1.130E+03 0.000E+00 RB-88 3.025E+05 0.000E+00 RB-89 3.874E+05 0.000E+00 RB-90 3.746E+05 0.000E+-00 1-130 1.681E+-04 0.O00E+00 1-131 4.354E+05 0.000E+00 1-132 6.291E+05 O.000E+-00 1-133 8.836E+05 0.000E+00 1-134 9.685E+05 0.000E+00 1-135 8.260E+05 0.000E+-00 XE-131M 4.850E+03 0.000E+00 XE-i133 8.436E+05 0.000E+00 XE-133M 2.529E+04 0.000E+00 XE-135 3.025E+05 0.000E+00 XE-135M 1.745E+05 0.000E+00 XE-i137 7.700E+05 0.000E+00 XE-138 7.235E+'05 0.000E+00 CS-134 1.093E+05 0.000E+00 CS-134M 2.641E+04 0.000E+00 CS-135 3.762E-01 0.000E+'00 CS-136 3.490E+04 0.000E+00 CS-137 6.627E+04 0.000E+00 CS-138 8.036E+05 0.000E+00 CS-139 7.604E+-05 0.000E+00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

= 1.000000 HOURS TOTALACTIVITY

= 5803093.000000 CURIES ISOTOPE ISOTOPE INITIAL ADDITION ACTIVITY NUMBER NAME ACT(CI) RATE(CI/HR) (CI)60 KR-83M 60 R-3M 5.283E+04 0.000E+00 3.640E+04 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 9 of 12 b1 1K1-t-i5 b5.Ubll:+UJ U.UUUL=+UU b5.Ub2ZI:+U3 62 KR-85M 1,108E+05 0,000E+00 9,465E+04 63 KR-87 2.113E+05 0.000E+00 1.223E+05 64 KR-88 2.977E+05 0.000E+00 2.324E+'05 65 KR-89 3.618E+05 0.000E+00 7,583E-01 73 RB-86 1.130E+03 0.000E+00 1.128E+03 75 RB-87 0.O00E+00 0.O00E+00 2.680E-10 76 RB-88 3.025E+05 0.000E+00 2.570E+05 77 RB-89 3.874E+05 0.000E+00 3.237E+04 78 RB-90 3.746E+05 0.000E+00 2.334E-01 84 SR-89 0.000E+00 0.000E+00 8.717E+i01 86 SR-90 0.000E-+00 0.O00E+00 7.483E-02 92 Y-90 0.000E+00 0.000E+00 7.497E-04 140 1-130 1.681E+04 0.000E+00 1.589E+04 141 1-131 4,354E+05 0,000E+0O 4,338E+'05 142 1-132 6.291E+05 0.000E+00 4.630E+05 143 1-133 8.836E+05 0.000E+00 8.539E+05 144 1-134 9.685E+05 0.000E+00 4.354E+05 145 1-135 8.260E+05 0.000E+00 7.446E+05 146 XE-131M 4.850E+'03 0.000E+00 5,900E+03 147 XE-133 8.436E+05 0.000E+00 8.439E+05 148 XE-133M 2.529E+-04 0.000E+00 2.497E+04 149 XE-135 3.025E+05 0.O00E+00 3.420E+-05 150 XE-135M 1.745E+05 0.000E+00 1.214E+04 151 XE-137 7.700E+05 0.000E+00 1.804E+01 152 XE-138 7.235E+05 0.O00E+00 6.725E+04 159 CS-134 1.093E+05 0,000E+00 1.093E+i05 160 CS-134M 2.641E+04 0.000E+00 2.079E+04 161 CS-135 3.762E-01 0.000E+00 3.762E-01 162 CS-136 3.490E+04 0.000E+00 3.483E+04 163 CS-137 6.627E+04 0.000E+00 6.627E+04 164 CS-138 8.036E+05 0.O00E+00 3.777E+05 165 CS-139 7.604E+05 0.000E+00 9.546E+03 170 BA-136M 0.000E+00 0,000E+00 3.483E+04 171 BA-137M 0.O00E+00 0,000E+00 6.627E+04 172 BA-139 0.000E+00 0.000E+00 5.834E+04 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME ACTIVITY (CI)RB-86 1128.253000 RB-87 2.680317E-10 S R-90 7.48271 9E-02 Y-90 7.497340E-04 I-130 15889.090000 CS- 134M 20787.760000 CS-i135 3.762085E-01 BA-I136M 34826.530000 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 10 of 12 WHAT IS THE SOURCE VOLUME (CC)4. 142000E+09 WHAT IS THE SOURCE DENSITY (GM/CC)1.530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.THE NUMBER OF ENERGY GROUPS INPUT: 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 3.000000E-01 2 4.000000E-01 3 5.000000E-01 4 6.000000E-01 5 8.000000E-01 6 1.000000E+00 7 1.500000E+00 8 2.000000E+00 9 3.000000E+00 10 4.000000E+00 MAX GAMMA ENERGY OF BR 90 IS GREATER THAN THE MAXIMUM GROUP ENERGY STANDARD FIXUP: MAXIMUM GROUP ENERGY IS SET TO THE HIGHEST ENERGY YET ENCOUNTERED+'0.01 MEV NEW MAX ENERGY = 5.210000 MEV MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO, SOURCE STRENGTH UC/GM UC/CC CI KR 83M 45 5.743404E+03 8.787408E+00 3.639745E+'04 KR 85 47 9.549409E+02 1.461060E+00 6.051709E+03 KR 85M 46 1.493609E+04 2.285221E+01 9.465387E+04 KR 87 48 1.929548E+04 2.952209E+01 1.222805E+05 KR 88 49 3.667663E+04 5.611524E+01 2.324293E+05 KR 89 50 1.196635E-01 1.830852E-04 7.583388E-01 KR 89 51 1.196635E-01 I.830852E-04 7.583388E-01 RB 88 67 4.055830E+04 6.205420E+01 2.570285E+05 RB 89 68 5.108358E+03 7.815787E+00 3.237299E+04 BR 90 69 3.682816E-02 5.634709E-05 2.333896E-01 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEi 99-01 Rev 6 EAL Calculations 11 of 12 1 131 I 132 I 133 I 134 I 135 4 5 6 7 8 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 134 33 CS 136 34 CS 137 35 CS 138 36 CS 139 76 BA 137M 78 BA 139 37 6. 845877E+04 7.305456E+'04 1 .347498E+05 6.869780E+04

1. 174962E+05 9.310631E+0 1 .331608E+05 3.940021 E+0 5.397361 E+04 1 .915720E+0J 2.846699E+00 1 .061136E+04 1 .724668E+04 5.495517E+03 1.045720E+04 5.960024E+04 1 .506267E+03 I1.045720E+0 9.206075E+03 C... l-trtij OL_.1 .-u 1.047419E+02 1.117735E+02 2.061673E+02 1.05 1076E+02 1. 79769 1E+02)2 1.424527E+00 S2.037359E+02

'3 6.028232E+00

• 8.257963E+01

'3 2.931051E+00

• 4.355450E-03 1 .623539E+01

• , 2.638742E+01 S8.408140E+00

• , 1.599952E+01

• , 9.118837E+01

• 2.304588E+00 4 1.599952E+01 1 .408530E+01 U,., I i IJ1 l ,JU -, UL. IJ 4.338410E+05 4.629658E+05 8.539448E+05 4.353559E+05 7,446038E+05 5.900389E+-03 8.438743E+05 2.496894E+-04 3.420448E+05 1 .214042E+04 1 .804027E+01 6.724697E+04 1 .092967E+-05 3.482652E+

04 6.627002E+04 3.777023E+05 9.545605E+-03 6.627002E+04 5.8341 30E+04 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)1 2 3 4 5 6 7 8 9 10 3.000000QE-01 4.000000E-01 5.000000OE-01 6.000000E-01 8.000000E-01 I1.000000E+00

1. 500000E+00 2.000000E+00 3.000000OE+00 5.21 0000E+00 1.51 0080E+06 1 .638563E+06 7.8 10211E+05 4.458929E+06 9.754498E+06 8.118191E+06 I1.504968E+07 6.422278E+06 6.717616E+06 5.222530E+00

• 2.084916E+16 1.696731E+16 6.469979E+15 3.078147E+16

5. 05039IE+ 16 3. 362555E+ 16 4.155719E+16 1 .330054E+16 9.274789E+

15 4.151962E+09 Table 6.1,-I. Units used to express S tabulated in relation to the units used to express q q ~Point source Line source Surface

• Volume cv as a factor of W o S SOUrCe source So S *S S y-photons cmn 2 s 7-photons cm 2 s MeV-cm 2 s_ MeV~y-photons 5 m~i (mfilicuries)

MeV 5 7-photons oms mCi cm MeV oms mCi y,-photons cm 2 s mCi cm 2 MeV mQi y-photons cm 8 sI mCi cm 8 MfeV cm~s m~i 3.7. 1Ov disintegrationis y-photons smCi *ndisintegration Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 12 of 12 cm~s cm cm 2 cms disintegrations V-photons E MeV smCi ndisintegration photon-'-MeY mCi mCi mCi M~eV cm 2 s mCi cm cm 2 cm 3 4" cm 2 smCi_ MeV ...Beg cm 2 s rads V-photons y-photons y-photons y'-photons 7-photon ' /a 3 0 0 h s cms cmzs cma s 100 ergs/(g.rad) ergs _cera 6-s rads MeV MeV MeV MeV 1.6' 10-8 eeV"a "g'-'0 h-h1 s cms cmts cross 100 ergs](g rad)Me 1 6 0 ergs -cm2 360 R V-photons V-photons V-photons V-photons E-hoo M.e10- ht-. 60-h s cms cross cmas 87.7 ergs/(gR)ergs *cm--2 a 36 R MeV MeV MeY MeV 1..0aM--e"V g s cms cm~s cm~s 87.7 ergs/(gR)R mCi mCi mCi 2"h- mCi cm cm 2 cm 2 2VK hinCi R mg-equiv.R mg-equiv.

Ra mg-equiv.

Ra 4 , Rcm 2-- rag-equiv.R1a cm cm 2 cm 8 a 8 h mag-equiv.

Ra Southern Nuclear Design Calculation

[Plant: HNP Unit: 1&2 [Calculation Number: SMNH-13-021

[Sheet: H-I ATTACHMENT H -VALIDATION OF SPIRAX SARCO ON-LINE STEAM TABLES Rather than interpolate from the ASME steam tables, an on-line set of steam tables was used to determine the specific volume of the reactor coolant at normal operating conditions.

Spirax Sarco, a global provider of products for the control and efficient use of steam, provides on-line steam tables at their company website, http://www.spiraxsarco .com/resources/steam-tables. asp.Reactor Coolant Normal Operating Conditions To verify that the Spirax Sarco steam tables provide accurate results, the specific volumes of subcooled water at 1000 and 1100 psia and 470 and 490 F (see sheets below) are compared below to the ASME steam table values.Units ASME Spirax Sarco Delta*ASME Spirax Sarco Delta*P psia 1000 1000 1100 1100 ____T F 470 470 470 470 ____SV cu ft/Ibm 0.019722 0.0197175

-0.02% 0.01 9704 0.01 96999 -0.02%T F 490 490 490 490 SV cuft/lbm 0.020140 0.0201348

-0.03% 0.020119 0.0201141

-0.02%* Delta =[(Spirax Sarco -ASME)/ASME]

X 100%The Spirax Sarco steam tables agree extremely well with the ASME steam tables.The linearly interpolated results from the ASME Steam Tables would likely be less accurate than using the on-line steam tables because specific volume is a non-linear function of pressure and temperature.

Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 ICalculation Number: SMNH-13-021 Isheet: H-2 Ho.me Aboat U. Pradect & 50w. -tmdsflaek

& r'r.nag Rammo -Contsct Sub Saturated Water Region -Steam Table Al any txeosse, water bedow tis saturalkon temperature ts said to be in a sub saturated state For exaopte' water at a reesure of I atmospfhere and a temperature betow I fIte saturated temperature of Is sub saturated.

Water at a pressure of 10 atroosptteres has a saturaion temp4erature of 180"C, and so water below5 fits temrperature Is also sub saturated.

Learn more about steam to ottutr ioal -jters.Set your for thseso steam tables Note: -You cannot use commas 4,) as decimal points.Pleas, use periods (.)Example: 1.02 not 1,02 Feature* Tra~let reat Swets tlamt 0 Preosora aid tamrtratere 4~ Sto~e Value '~ Table urn tie atoojot.Pressure TemeratEl Vapou PressuJre Saturation Temperature Specifc Entlialpy of Waler (hi)Oerelt of Water Specific Voftame of Water (s)Specifi Entrop of Water (si)so jAra k~n9 J~5 5 K Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13-021 Isheet: H-3 I-- e onr Spia Sxa.o In Hsqre AttstUs is Preoferts&

Seyi.

• tidtslbes

&Aepij~oans

• Trattme is Contest V Yes a-e her: Homee I) Seantt Sten Thl Suil Saturated Water Region Sub Saturated Water Region -Steam Table AJ1 any p~ewawe, water below its saturation teenperature is said to be to a su saturated state orexample water at a pressure of 1 arnospttere and a temlperature below tesaturated temperature of lt000C Is sub saturated.

Water at a pressure of 1atmosphers has a saturation temperature of 180"C, arnd so water below Urstemperature Is also sub saturated.

i Learn more about steam is our tutortat -Set your io..tS for theKse steam tables.Note: -You cannot use commas (,) as decimal points.Please use periods (.)Example: 1.02 not 1,02 Feature L Tratng ba~ltos 0* Single Value ©Tabl~e El Pressure pat atiselat."temperature Vapour Pressure ear gauge Saturation Temperature El Specifc Enthalpy of Water (hi)Denstit of Water Specifc Volume of Water (v)Specifi Entropy of Water (is)Jigs JAg K El El Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13-021 Isheet: HA4 r intiffii~nai site for Spirax Sarco Ve.t yew eltonat Seea~c Sane Cebsita tieme Abe..t Us Predart. & Serewes

• todasbios

£ Apptmabee.

-Trarnere Rewarm Coetad -Yoet tw a : Htom I> Resomy Sit Sarta Wate Regai Sub Saturated Water Region -Steam Table MIany prossee water below its saturation temperature ts said to be itt a sub saatated state For example, water at a preemie of I atmnosphere and a temperature below Ite saturated temperature of lt00'C is tub saturated, Water at a pressure of 10 atmospheres has a satura-=iton temperature of and so water below hits temperature Is also sub satuated.eant merm abou Seam le ourtutolal Set your for these steam tables Note: -You cannot us. commas I,) as decimal points.Please us. periods (.)Example: 1,02 not 1,02 Feature* Treein ma 0 Pressure aind Tanlperature

• Sn~ge Veto. Table Pressure EJ Temp-=ertture Vapour Pressure Saturation Tenmterattae Specific Enthatpy of Water (he)Density of Water Specific Volume of Water (v)Specifi Entrop of Water (Si)Jag eWe'JAg K i~J Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13"021 ISheet: H-5 I~%ar-Intrntonal site for Spirax Sarco Homee Absst U. -P Proa..ct & Seswcea -lndealie.

& Apetir~atees

• ;ionineg Raelweca

• Coetast ,-Yeu m ere:ar Home ), k T ite Sue Set~ratad Water Regis..Sub Saturated Water Region -Steam Table Al any pressure, water below fits satuion ateperature is said to be to a suab saturated state Fer examine', water at a pressure of I atmosphere and a temperature below3 Ihe saturated temperature of 1l0"C is sub saturated Water at a pressure ofE 10 atmospheres has a saturation temperature of 180°C and so water below thds tenperature is also sot, saturated Learn more about steam is our tutorial -istSean?Set yoer for these steam tale Note: -You cannot use commas (,) as decimal points.Please us. periods (.)Example: 1.02 not 1,02 Feature LI Tratleile that s:ea °"p Eo we r sse" a er t 0 Pressure TemperatureStogie Value © Table jI.. P.absd E]Vapour Pressure Saturation Temperature Spcii Erthlp of Water (he)Denslt of Water Specifi volume of Water (v)Specifc Entropy of Wate (se)Oar eases'C kg/o9 JAg K Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 1 of 25 Purpose: The purpose of this appendix is to calculate radiation readings for AS1 (RS1) EALl.Defining Units that MathCAD will understand MWt := 10 6-W 1 Ci: .-*Bq 2.7.10 Mega Watt Thermal is same as watts.Per FGR-11 page 219 tlCi := 10-6 Ci Rem := 0.01lSv Rem mRem:= -1000 Per FGR-11 page 219 cc := mL 1 cps := --sec 1 cpm := -" mm The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used in HNP FSAR Chapter 15 evaluations.

Such as Iodine's and noble gases Table 15.3-4. It is assumed that only iodine's and noble gases needs to be considered, because particulates will be retained in the primary containment water.Isotopes : "1-131""1-132""1-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe-138" Core__Inventory

(2.72E+04" 3.93E+04 5.52E+04 6.05E+04 5.16E+04 3 .30E+03 3.78E+02 6.92E+403 1.32E+04 1 .86E+04 2.26E+-04 3.03 E+02 5.27E+04 1.58E+03 1 .89E+04 1 .09E+04 4.8 1E+04 4.52E+04 Ci MWt'I Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEl 99-01 Rev 6 EAL Calculations 2 of 25 CorePower

= 2818.MWt RCS inv := 9965 ft3 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 °F T Avgu 1 := = 4.622E+002.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 0 F T Avgu 2 := 2= 4.804E+002.

0 F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 VRCS water : 0.0199161-

-Ibm* According to http://www.spiraxsarco.

com/resources/steam-tab les.asp the specific volume for 1060psia and 480.35 F conditions is 0.0199161ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv RCS mass .- -2.27E+008-gm

'URCS water Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 3 of 25 The Core Release Fractions are taken from RG1.183 Table 1. This fraction describes release of isotopes to RCS water.Isotopes 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83mi""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-138" F_Core_Release

(0.3 10.3 0.3 0.3 0.3 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEl 99-01 Rev 6 EAL Calculations 4 of 25 The concentration of isotopes in RCS is calculated as follows: Conc_IsotopeRCS

Core Inventory.Core

_Power-F FCore Release RCS mass Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""1-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe-138" Cone_IsotopeRcS 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.013E+005 1.464E+005 2.056E+005 2.254E+005 1.922E+005 4.097E+004 4.693E+003 8.592E+004 1.639E+005 2.309E+005 2.806E+005 3.762E+003 6.544E+005 1 .962E+004 2.347E+005 1 .353E+005 5.972E+005 5.612E+005 gm The RCS will have some equilibrium noble gases and some iodine in the RCS water from normally operating reactor. According to HNP FSAR U2 Table 11.1-2 the levels of iodine's is on the order of 1 E-1 uCi/g which is negligible to calculated above order of magnitude of 1 E5 uCi/g.The reason these isotopes are in low concentrations because they are continually removed in the power plant steam condensers.

Thus initial equilibrium noble gases and iodine's will be neglected.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 5 of 25 The Partition Coefficient is the ratio of the concentration of a nuclide in the gas phase to the concentration of that nuclide in the liquid phase when the liquid and gas are at equilibrium.

It is assumed that 100% Noble gases are released into the steam. According to NUREG-0016 Table 2-7 page 2-13 the Iodine's partition coefficient is 0.004.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe-138" FPartCoefficient

0.004 0.004 0.004 0.004 0.004 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 6 of 25 The release concentration of isotopes is calculated by multiplying the concentration of isotopes in RCS by the partition factor and an arbitrary density of release fluid. The arbitrary density of fluid will cancel out. For ease of math calculations it was chosen to be 1lgm/cc.Prls c=c.g XL =(ConcIsotopeRCS-FPartCoeffcient-Pris)>Isotopes =0 1 2 3 4 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-8 5""Kr-85m""Kr-87""Kr-88""Kr-89""Xe- 13 lin""Xe- 133""Xe-133rn""Xe-135""Xe-135mn""Xe-137""Xe-138" XRL~S 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 4.053E+002 5.856E+002 8.225E+002 9.014E+002 7.688E+002 4.097E+004 4.693E+003 8.592E+004 1.639E+005 2.309E+005 2.806E+005 3.762E+003 6.544E+005 1.962E+004 2.347E+005 1.353E+005 5.972E+005 5.612E+005 IICi cc Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 7 of 25 Reactor Building Vent TEDE part of the calculation:

According to HNP ODCM page 3-14 the release from reactor building vent is considered to be a ground level release.-6 sec__X~gnd := 8.37.10--3 m Table 3-4 of HNP ODCM page 3-17 Q~rlS~x1d

= 1.42.10 mL = 3.009E+005.cfm Table 3-4 of HNP ODCM page 3-17 sec The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEABRxBLD

= (Q~rlspxld'XQgnd-XRj~s)

Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""1-133""1-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe- 131mi""Xe-.133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe-138" XEAB RxBLD 7-0 1 2 3 4 5 6 8 9 10 11 12 13 14 15 16 17 a 4.817E-001 6.96E-001 9.775E-001 1.071E+000 9.138E-001 4.87E+001 5.578E+000 1.021E+002 1 .948E+002 2.745E+002 3.335E+002 4.472E+000 7.777E+002 2.332E+00 1 2.789E+002 1.609E+002 7.098E+002 6.67E+002 tlCi cc Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 AttachmentI Titie: NEl 99-01 Rev 6 EAL Calculations 8 of 25 The Dose Conversion Factor (DCF) for Effective Dose Equivalent (EDE) was taken from FGR12'Effective Column" of Table 111.1. The unit conversion was performed with MathCAD internal features.Isotopes 0"I-13 1""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe- 135m""Xe- 137""Xe-138" DCFEDE :

1.12E-13 2.94E-14 1.30OE-13 7.98E-14 1.50E-18 1.19E-16 7.48E-15 4.12E-14 1.02E-13 0 3.89E-16 1.56E-15 1.37E-15 1.1 9E- 14 2.04E-14 0 5.77E- 14)Sv" Bq. sec 0 2.427E+005 1.493E+006 3.92E+005 1.733E+006 1.064E+006 2E+001 1.587E+003 9.973E+004 5.493E+005 1.36E+006 OE+O000 5. 187E+003 2.08E+004 1.827E+004 1.587E+005 2.72E+005 OE+O000 7.693E+005 mRem. cc hr. 3 Sv. m mRem.cc-1.333E+019.

B q. sec hr. MathCAD internal conversion feature.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 EAL Rev 6 Calculations 9 of 25 The Dose Conversion Factor (DCF) for Committed Effective Dose Equivalent (CEDE) was obtained from FGRI1 Table 2.1 column labeled "Effective".

The unit conversion was performed with MatchCAD internal features.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"1-131""1-132""I-133""I-134""1-135""Kr-83m"~"Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe- 138" DCFcEDE: S8.89E_09" 1.03E-10 1 .58E-09 3.55E-11 3.32E-10 0 0 0 0 0 0 0 0 0 0 0 0 0 Sv 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.293E+001 3.815E-00 1 5.852E+000 1.3 15E-001 1.23E+000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O OE+O000 OE+O000 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 mRem p.Ci Sv mRem= 3.704E+009.-

Bq liCi MathCAD internal conversion feature. Same conversion factor is available on page 121 of FGR-11.

SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 AttachmentiteNE9-0Rv6EACaclios1 If2 The EDE, CEDE, and TEDE are calculated as follows: The exposure time is provided in NEI 99-01.texp := lhr 3-4m BR := 3.5.10 -sec The breathing rate for persons offsite is listed in section 4.1.3 page 1.183-16 of RG-1.183.EDE~xBLD :=(DCFEDE'XEABRxLD'texp)f Effective dose Equivalent during a release from Reactor Building Vent.,

Effective Dose Equivalent from CEDERxBLD

= (DCFcEDE'XEAB_RxBLD'*texp BR)nhalation during a release from Reactor Building Vent.0 1 13 i5 16.7 i *0' 1. 169E+005 1.039E+006 3.832E+005 1.857E+006 9.723 E+005 9.74E+002 8.851E+003 1.019E+007 1.07E+008 3,733E+008 0E+000 2.319E+004 1.618E+007 4.259E+005 4.426E+007 4.375E+007 0E+000 5. 132E+008*mRem CEDE~xBLD

-41:7 9'9'10 12 13 14 15;17 1.998E+007 3.345E+005 7.208E+006 1.775E+005 1.416E+006 0E+000 0E+000 OE+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+i000 0E+000 0E+000-mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

TEDERxB3LD

- ZEDEpxBLD

+ ECEDEPxBLD

= 1.142E+009.mRem Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 11 of 25 The reactor building monitor indication related to the exclusion area boundary TEDE is calculated as follows: X100_RxBLD

S100 mRem X The 100 mRem comes from NEI 99-01.Isotopes=2'4 6'8 11 12 14 15 17"I-131""1-132""I-133Y"I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Xe- 13 im""Xe- 133""Xe-133m""e15""Xe-135m""Xe- 137""Xe-138" Xlo0 RxBLD =(0f 2-5,¸6¸: 7 8BI 9 10 11 12 15 16 17* ,.. 0 " 3.549E-005

5. 128E-005 7. 203 E-005 7.895E-005 6.733E-005 3.589 E-003 4.111E-004 7.525E-003 1.435E-002 2.023E-002 2.458E-002~

3.295E-004 5.73 1E-002 1.718E-003 2.055E-002

1. 185E-002 5.23 1E-002 4.915E-002]

IlCi cc Since the detector only respond to noble gases according to HNP FSAR Table 7.5-1 (Sheet 31 of 34) note 14 and pg 3 of Doc ID RE203727981, thus only noble gas isotopes will be summed.17 XTEDE_100_RXBLD

=i=5 This is summation of rows that correspond to Xl00_RxBLDinoble gases only. These rows are 5 through 17 of the matrix above.XTEDE_100_RXBLD

= 2.639E-001.

c Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NE! 99-01 Rev 6 EAL Calculations 12 of 25 CDE part of the calculation:

The Thyroid Committed Dose Equivalent (CDE) from inhalation was obtained from FGR 11 Table 2.1 column labeled "Thyroid".

The unit conversion was done via MathCAD internal unit conversion feature.Isotopes =0.2 3 5.6 7 8 10 11.12 13 15 17 0"I-131""I-132""1-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe- 135m""Xe-137""Xe- 138" DCFcDE : r2.92E_07"N 1 .74E-09 4.86E-08 2.88E-10 8.46E-09 0 0 0 0 0 0 0 0 0 0 0 0 0 Sv.0 2.3 4 6 7 8 9.10 11 12 13!5 16 17 0 1.081E+003 6.444E+0001.067E+000

3. 133E+001 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 mRem liCi 1 Sv- 3.704E+009.

• e Bq p~tCi MathCAD internal conversion.

Corresponds with FGR-11 page 121.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations l3 of 25 The CDE exposure due to the Reactor Building Vent release is calculated as follows: 3 BR = 3.5E-004.m sec texp = 1E+000.hr Provided above In tis quaton he DE dse onvrsinforvise mulipledbistpscneraont exclusion area boundary, and breathing rate, and time of exposure.IsotopoCSDERxBLD

= (DCFCDE.XEAB_RxLD'BR'texp))

IsotopesCDE~xLD=

2,"5 6 17 81 i-i 12 14'5 16 17',i "0. ..6.564E+008 5.651E+006 2.217E+008 1.44E+006 3.608E+007 OE+O00 OE+O000 OE+O000 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 OE+O000 OE+O000*mRem CDE~xLD := ZlsotopesCDERxB1LD

= 9.213E+008.mRem Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 14 of 25 The reactor building monitor indication related to the exclusion area boundary CDE is calculated as follows: ( 500mRem~5 OO RxBLD :=-y CDERXBLD) .XRLS The 500 mRem comes from NEI 99-01.Isotopes=0 1 2 3 4 5 6.7 8.9 10 11 12 13 14 15 16 17 0"1-131""I-132""I-133""1-134""I-135""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-138" X 5 00 RxBLD=0 2 3 6 7 8 9 10 111 12 14 15 16 17 0 2.2E-004 3. 178E-004 4.464E-004 4.892E-004

4. 173E-004 2.224E-002 2.547E-003 4.663E-002 8.895E-002 1.253E-001 1.523E-001 2.042E-003 3.551E-001 1.065E-002 1.274E-001 7.345E-002 3.241E-001 3.046E-001 cc 17 XCDE_500_RXBLD

= E X500-RxBLDi i= 5 XCDE_500 RxBLD =_l.635E+000-l~This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.

Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 15 of 25 XTEDE_100 RxBLD = 2.639E-001.Ixi cc This was calculated above and since it is more limiting should be used as the indication for this EAL.Main Stack TEDE Part of the calculation:

According to HNP ODCM page 3-14 the release from Main Stack is considered to be a Elevated release.-8 sec X Qelev := 4.10-10 .-3 m Q~rlSMain

= 9.44.10 mL- 2E+004.cfm sec Table 3-4 of HNP ODOM page 3-17 Table 3-4 of HNP ODCM page 3-17 The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEABMain :=(QrlSMain"XQelev'XRWs))

Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe-138" XEABMain 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.569E-004 2.266E-004 3.183E-004 3.489E-004 2.976E-004 1.586E-002 1.817E-003 3.326E-002 6.344E-002 8.939E-002 1.086E-001 1 .456E-003 2.533E-001 7.593E-003 9.083E-002 5.238E-002 2.312E-001

2. 172E-001 liCi CC Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 16 of 25 The dose conversion factors were provided above in this calculation and are used in the same manner. The EDE, CEDE, and TEDE are calculated as follows: tex = 1E+000.hr 3 m BR = 3.5E-004.-

sec Provided above Provided above EDEMain := (DCFEDE'XEAB_Main'texp))

Effective dose Equivalent during a release from Main Stack.Committed Effective Dose Equivalent from inhalation during a release from Main Stack.CEDEMain := (DCFcEDE-XEABMain~texp.BR)

EDEMain=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.806E+001 3.384E+002 1.248E+002 6.048E+002

3. 166E+002 3.172E-001 2.882E+000 3.3 17E+003 3.485E+004 1.216E+005 0E+000 7.552E+000 5.268E+003 1.387E+002 1.441E+004 1.425E+004 0E+000 1.671E+005.mRem CEDEMain =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 6.508E+003 1 .089E+002 2.347E+003 5.78E+001 4.61E+002 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000*mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 17 of 25 TEDEMain := EEDEMain + Main =3.718E+005.mRem The Main Stack monitor indication related to the exclusion area boundary TEDE is calculated as follows: Xl0Main := 100T~en)'XRLS The 100 mRem comes from NEI 99-01.Isotopes =o 1 2 3 4 7..8 9'10 11l 1i2 13 16 17"I-131""I-132""I-133""1-134""I-135""Kr-83m""Kr-85""Kr-87""Kr-88"*"Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe- 138" X100 Main 0 1.09E-001 1.575E-001 2.212E-001 2.424E-001 2.068E-001 1.102E+001 1.262E+000 2.311E+001 4.408E+001 6.211E+001 7.547E+001 1.012E+000 1.76E+002 5.276E+000 6.311E+001 3.64E+001 1 .606E+002 1.509E+002 IlCi ee Since the detector can discriminate noble gases according to pg 3 RE203727981, thus only noble gas isotopes are summed. The P DMS connects 1D11N055 and 1D11N056 to 1D11P006 and manual $57925.17 XTEDE_100-Main

= E Xl0Main.i= 5 This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 18 of 25 XTEDE 100 Main = 8.104E+002*-~

--cc CDE part of the calculation:

The ODE exposure due to the Main Stack release is calculated as follows: 3 m BR = 3.5E-004.-

sec texp =1E+OOO0hr Provided above In tis euatin th ODEdoseconvrsinforvise mulipledbistpscneraont exclusion area boundary, and breathing rate, and time of exposure.IsotopesCDEMain

= (DcCcDE.XEAB_Main.BR~texp))

IsotopesCDEMain

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 2. 137E+005 1.84E+003 7.22E+004 4.689E+002

1. 175E+004 OE+O00 OE+O00 OE+O000 OE+O000 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00.toRero Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 19 of 25 CDE Main := lsotopesCDEMain

= 3E+005.mRem The Main Stack monitor indication related to the exclusion area boundary CDE is calculated as follows: X50Main := *XDEai)RLS The 500 mRem comes from NEI 99-01.Isotopes = 1, 2.5 16 7 9 i10 11 12 13 14:15 16 ,17"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe- 138" X500_RxBLD

='2" 7 8 9 10 11 12 13 14 17 0 2.2E-004 3. 178E-004 4.464E-004 4.892E-004

4. 173E-004 2.224E-002

2. 547E-003 4.663E-002 8.895E-002 1.253E-001 1.523E-001 2.042E-003 3.551E-001 1 .065E-002 1.274E-001 7.345E-002 3.241E-001 3.046E-001 cc XCDE_500 Main:=i= 5 X50Maini This is summation of only noble gases, rows 5 through 17.I pCi I XCDE 500 Main 5.022E+003 I -cc Plant: HNP U1 & U2 SNC CALCULATION SMNH-13--021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 20 of 25 IXTEDE_100 Main = 8"I04E+002" liCi[cc This was calculated above and since it is more limiting should be used as the indication for this EALU Recombiner Building Vent TEDE part of the calculation:

According to HNP 0DCM page 3-14 the release from recombiner building vent is considered to be a ground level release.Sec XQgnd =8.37E-006.--

35m m Provided above

d. := 2.36.105 ..- = 5.001E+002.cfm Table 3-4 of HNP ODCM page 3-17 The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEAB~aeB~d

-- (OrIsReCBld.X gfnd-Xjus))

Isotopes ="I-131"1"I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 im""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe- 138" XEAB RecBld 0 I.2.3 4 5..6.7.8.9 12 14 15 16 17 8.006E-004 1.157E-003 1.625E-003 1.781E-003 1.519E-003 8.094E-002 9.,271E-003 1.697E-001 3.238E-001 4.562E-001 5.543E-001 7.432E-003 1.293E+000 3.875E-002 4.636E-001 2.673E-001

1. 18E+O000 1.109E+000 litCi cc Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment II Title: NEI 99-01 Rev 6 EAL Calculations 21 of 25 The EDE, CEDE, and TEDE are calculated as follows: texp = IE+000.hr 3 m BR = 3.5E-004.--

see Provided above Provided above EDERecBld

= (DCFEDE'XEAB_RecBld'texp)r Effective dose Equivalent during a release from Reactor Building Vent./ ,>Committed Effective Dose Equivalent from CEDERcCBId

= I[DCFcEDE'XEABRecBld*texp'.BR) inhalation during a release from Reactor Building Vent.EDERecBId=

0 1.943E+002 1.727E+003 6.369E+002 3.086E+003 1.616E+003 1.619E+000 1.471 E+00 1 1.693E+004 1.778E+005 6.204E+005 OE+O000 3.855E+001 2.689E+004 7.079E+002 7.355E+004 7.272E+004 0E+000 8.529E+005.mRem CEDERecBId

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.321E+004 5.56E+002 1. 198E+004 2.95E+002 2.353E+003 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 OE+O000 OE+O00 0E+000 0E+000-mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 22 of 25 TEDERecBId

= ZEDERecB~d

+ ZCEDERecBId

= 1. 898E+006-mRem The recombiner building monitor indication related to the exclusion area boundary TEDE is calculated as follows: Xl00 RecBld : C 00rnRern STEDERecBldI "XRLs The 100 mRem comes from NEI 99-01.Isotopes =0 0 "I-131" 1 ."1-132" 2 "I-133" 3 "I-134" 4 "1-135" 5 "Kr-83m" 6 "Kr-85" 7 "Kr-85m" 8 "Kr-87" 9 "Kr-88" 10 "Kr-89" 11 "Xe-131lm" 12 1"Xe-133" 13 "Xe-133m" 14 "Xe-135" 15 "Xe-135m" 16 "Xe-137" 17 "Xe-138" XlO0_RecBlid-0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 2. 136E-002 3.086E-002 4.334E-002 4.75E-002 4.051E-002

2. 159E+000 2.473E-001 4.528E+000

8. 637E+000 1.217E+001 1.479E+001 1.983E-001 3 .448E+00 1 1.034E+000 1,237E+001 7, 132E+000 3,147E+001 2.957E+001 iiCi CC Since the detector only respond to noble gases according to pg 3 of DoclD: RE203727981, thus only noble gas isotopes will be summed.17 XTEDE_100_RecB~d

= X100-RecBldi i= 5 This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.XTEDE_100_RecBld

=1'588E+002-lCi cc i Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 23 of 25 CDE part of the calculation:

The ODE exposure due to the Main Stack release is calculated as follows: BR = 3.5E-004.m sec texp = 1E+000.hr Provided above Provided above In this equation the CDE dose conversion factor is multiplied by isotopes concentration at exclusion area boundary, and breathing rate, and time of exposure.IsotopesCDERcCBld

= (DCFcDE' XEAB_RecBld.BR-texp))

IsotopesCDERecBld

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.091E+006 9.392E+003 3.685E+005 2.393E+003 5.996E+004 OE+O00 0E+000 OE+O00 OE+O000 OE+O000 OE+000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000.tRero Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 24 of 25 CDERcCB~d

= LlsotopesCDERecBld

= 1.53 1E+006-mRem The Recombiner Building monitor indication related to the exclusion area boundary CDE is calculated as follows: ( 500mRernm X50RecBld

= ) L The 500 mRem comes from NEI 99-01.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"1-131""1-132""1-133""I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe-138" X500_RecBld 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.323E-001

1. 912E-00 1 2.686E-001 2.944E-001 2.511E-001 1.338E+001 1.533E+000 2.806E+001 5.352E+001 7.542E+001 9.164E+001 1.229E+000 2.137E+002 6.407E+000 7.664E+001 4.42E+00 1 1.95E+f002 1 .833E+002 kiCi cc 17 XCDE_500_RecBld

= X500-RecBld.

i= 5 XCDE_500 Re =l 9.84E+002.

l~-~cc This is summation of only noble gases, rows 5 through 17.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 25 of 25 XTEDE_100_RecBld

= 1.58+/-0 ~_ cci This was calculated above and since it is more limiting should be used as the indication for this EAL.FCaIR 7 6 3 := 4.80-107 cpm cceD The calibration factor for 1011R763A/B was obtained from DoclD RE203186522.

XTEDEI 00_RecBld counts :=XTBDE_100_RecBld'F-Ca 1 R 7 6 3= 7.622E+009.cpm XTEDE 100 RecBld counts Purpose: The purpose of this appendix is to calculate expected radiation field at Drywell Radiation Monitoring sensors (CG1).As water level in the RPV lowers, the dose rate above the core will increase.

The dose rate due to this core shine should result in up-scaled Containment High Range Monitor indication and possible alarm. Containment Challenge Table calculations should be performed to conservatively estimate a site specific dose rate set point indicative of core uncovery (ie., level at TOAF). Additionally, post-TM I studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

Defining Units that MathCAD will understand MWt :=106.W Mega Watt Thermal is same as watts.1 Ci .- Bq Per FGR-11 page 219-11 2.7.10:= 10-6 Ci Rem := 0.01Sv Per FGR-11 page 219 Rem 1 1 mRem := -cc := mL cps :=-- cpm := -" 1000 sec mm The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top Of Active Fuel. The calculation pdf page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Sense the RPV core is assumed to be columized source the same radiation level is assumed to be at the edge of the RPV. This is conservative and according to the BH2-M-V999-0048 pdf page 39 column 2 on the edge of the SFP the radiation level is 2.29E5R/hr.

As can be seen the radiation level does not very much from the middle of the core to the edge of the core. The page B-9 of BH2-M-V999-0048 provides the gamma source strength broken down by energy groups. It can be seen that the maximum gamma strength is spread between 0.4 to 1.8 MeV, therefore this calculation will use linear attenuations that are associated with 1MeV gamma rays.RSFP := 2.68-105 R BH2-M-V999-0048 last page of the caic column 3.hr The next step is to calculate the plant elevation level that corresponds to the TOAF for vessel instrument zero.TOAFvz := -158.44in TS Bases 2.1.1.3 for 150 inch long fuel.VIZ := 517in VIZ :=517inthe vessel instrument zero is in reference H26189 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment J Title: NEI 99-01 Rev 6 EAL Calculations 2 of 4 VZEL :=143ft + lin The elevation of the vessel zero is in reference H16032 TOAFEL :=VZEL + VIZ -TOAFvz =199.37.ft According to H16241 and H26417 the Drywell Wide Range detector is located at EL156ft and 27ft from the center of RPV.DetEL :=156ft DetR := 27ft When the water level reaches TOAF some radiation will travel from the edge of the core to through the RPV steel and Concrete Sacrificial shield as shown on the figure below.therefore the first step is to find the angle (line of site) from TOAF to the detector.

This will allow calculating the distance that gammas will travel through the shielding material (steel and concrete).

H16032 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment JI Title: NEI 99-01 Rev 6 EAL Calculations 3 of 4 Per FSAR U2 Table 5.2-6 the RPV material is steel. Per FSAR U2 Table 5.4-1 and S 15213 the RPV inside diameter is 218inches.

Per FSAR U2 Table 5.2-10 the wall thickness is 5.38 inches.DIA~pV:= 218in Thickwall

= 5.38in The next step will determine the angle of the line of sight from the TOAF RPV edge to the detector.AY :=TOAFEL -DetEL = 43.37.ft Difference in elevation.

DIA~pV RPVR 2- -9.083.ft Determining radius from diameter DetR = 27.ft AX := DetR -RPVR =17.9 17-fi ta~)=opposite adj asent Difference in their radiuses to determine deltaX Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.(AX'\Q:= atani -I = 22.446.deg

~ AY,.J This is the angle for a shortest line of site travel for a gamma ray to the detector.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment4J Title: NEI 99-01 Rev 6 EAL Calculations 4 of4 This step will determine linear distance that a gamma ray will have to travel trough steel vessel at the angle determined above.ino)= opposite hypotenuse Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.Thickwl d steel .- wal -14.09l1in sin(ca)or dsteeI = 35.79.cm This step will determine radiation level gamma radiation by the RPV steel.1'steel := 0.460.-cm I(t) = Io-e-t'R RSFP = 2.68E+005.-

hr dsteeI = 3.579E+001.cm on the other side of the RPV wall due to attenuation of Linear attenuation for steel at 1 MeV from page 178 of Engineering Compendium of Radiation Shielding Vol 1.Radiation intensity equation due to material attenuation page 170 of Engineering Compendium of Radiation Shielding Vol 1.determined above determined above-P~steel'dsteel

-IRpV := RSFp.e hr.19-The resulting value is much smaller than the lower range of the Drywell Radiation Monitoring sensor of I to 10A7 R/hr (Attachment C). There is no point to further account for the attenuation that the concrete sacrificial shielding would introduce, since it will lower the value even further.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 1 of 15 Purpose: The purpose of this appendix is to calculate radiation readings expected at Area Radiation Monitors due to core shine. This is done to satisfy initiating condition CG1 EAL2b "Other indications".

Defining Units that MathCAD will understand MWt := 106-W Ci .- .Bq Per FGR-11 page 219 2.7-10 ltCi := 10-6 Ci Remn:= 0.01Sv Per FGR-11 page 219 Rem 1 1 mRem .- cc := mL cps := -- cpm :=-1000 sec mai The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top Of Active Fuel. The calculation pdg page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Since the RPV core is assumed to be a columnized source, the same radiation level is assumed to be at the edge of the RPV.DRsFP := 2.68.105.Ri BH2-M-V999-0048 last page of the caic column 3.hr The main body of this calculation (Section AU2 EALIlb) provides a list of Area Radiation Monitors and their distance to the edge of the drywell. The furthest monitor is Refueling Floor Stairway and is approximately 95 feet away from the center of the drywell. None of the listed monitors have a direct view of the reactor core. However, they do "see" gammas that reflect off the refueling area ceiling. The operating deck dose rate due to these reflected gammas is given by the following from Davisson "Gamma Ray Dose Albedos" (copy provided below in this attachment).

Area DRmon = DRsfp.COS(9).

.Rmonitor 2 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 2 of 16 The first step is to find the angle of the building roof reflection to the Refueling Floor Stairway monitor and the distance from the roof to the detector.AX := 95ft EL roof := 281ft + 9in =281.75.ft ELdet := 233ft AY :=ELroof -ELdet =4 8.7 5-ft tan(0)-= opposite adjacent This is the horizontal distance from the drywell center to the detector.Bottom elevation of the roof DWG H25963 Section 3.The detector elevation is given in the table for AU2 initiating condition within the main body of this calculation.

Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.Rmonitor := JAX 2 + AY 2 = 106.778.ft This is hypotenuse from the secondary containment ceiling to the radiation detector.cx := 0.5099%cx= 0.005099 or The dose albedo for 1 MeV gamma for iron with incident angle of 0deg and emerging angle of between 55.2 to 64.6 degrees. This albedo is obtained from the tables listed below in this attachment.

DWG H25694 shows a metal roof deck. 1MeV gamma is chosen because most of the radiation from the SFP falls in the range of 1MeV per BH2-M-V999-048 page B-9.The drywell radius is given in DWG H25570 RDrywell := 18ft+/- +l1in = 18.833.ft Ae:=r.Drywell Area Rem DRmon :=DRSFPD-cos(0).o

--cx = 60.975.-Rmonitor 2 h or DRmon = 6.098E+004.

mRem hr The area radiation monitor range is 1 -1E4 mR/hr as described in main calculation forAU2 initiating condition.

It can be seen that the radiation at the furthest Area Radiation Monitor due to reflection from secondary containment roof is off scale for the detector and thus all other monitors that are closer to the drywell would also be off scale.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 3 of 15 A145/SD-7$/1-4 A HANDBOOK OF RADIATION SHIELDING DATA J. C. COURTNEY, EDITOR Nuclear Science C~enter Louisiana State Universityv Baton Rouge and Shielding and Dosimetry Division American Nuclear Seclety JULY 1 1978 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K]Title: NEI 99-01 Rev 6 EAL Calculations 4 of 15/5-27 Gamma Bay Dose Albedos C. M. Davisson U. S. Naval PReserch Laboratoryi The dose rate reflected from a surface as deduced from Reference 1 through 4 may be represented as: P.R. =D.R,° cos ° a(Eo,98, , r where D.R. -Reflected dose rate D.R.° Dose rate incident on surface at 60 A -Reflecting area r Distance from center of reflecting area to receptor CA and r 1 must be in the saine units)tt(E 6wo, 6, *) = Dose al~bedo le abeod ll(E~ , c,8,>, for gaimas incident on water. concrete.

iron and techniques in an extension of the original work by Theus and Beech 6.The albedos are given for incident game energies of 0.2) 0.662) 1.0, p2.5 and 6.13 MIeV and for incident angles with respect to the normal of 00, 220O* 44°, 660 and 880, as well as for point sources on the surface of the materials.

The emerging polar angles, 8i, as well as the emerging sectors or directions into which the emerging gammas were divided are shown in Fig. 5.13. The values of the polar angles, O., and of the azimuthal angles 4, defining the emerging directions, are given on each page of Table 5.8.Note: The dose albedo values have statistical eirrors that range from 402 or 50Z at very small albedo values to 5% or 10% at large albedo values.Ref erences' Reactor Shielding Design Mdanual, T. Rockwell IllI, editor, rIID-7004 (larch 1956)p. 334.2 D. 3. Raso, "Ilonte Carlo Calculations on the Reflection and Transmission of Scattered Gaimna Rays," Nuzcl. Sol. and Eung. 17, 411 (1963). This report has a good discussion of the meaning of various terms and derived quantities.

The lose albedos given here are those which he described in quotes, as "dose" albedos.s W. E. Seiph, "Neutrons and Ga~mma-lay Albedos," DASA-189 2-2 (May 1967), OBNL-BSIC-21 (February 1968), or Chapter 4 of Weapons Radiation Shielding Handbook (NTIS No.AD-816 092). The dose albedos given here arc those defined as cni in this report.'9 R. L. French. and 1!. B. Wells, "An-Angle-Dapondent Albedo foi Fas 2--Reflactior Calculations," Nuel. Sci. and 1mg. 19, 441 (1964).5C. N. Dayvissom and-L. "Gamma-Ray Albedos of iron," NRL Quarterly on Nud. Scd. and Tech. (January 1, 1960), p. 43; and private communication.

SR. B. Theus and L. A. Beach, "Gamma-Ray Albedo," NRL Quarterly on tued. Sci. and Tech. (July-September 1955).

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment KI Title: NEI 99-01 Rev 6 EAL Calculations 5 of 15 5-28 F~iSure 5.13 Geometry arnd Solid Angle Divisions OAII{A 615 lOS 052120 Po1.r 300001000 0.02 03 0.66 IsV ASile 6,ioldos.

i1 Point

• incident Cl Point A EL 016 11 61 261 11. 66 861 lootcO 61 03 1.4 661 03 Soorce 0.-1.t 1 32.0-160.3 3.7217 5.0279 6.59 r.,g .1sg6 6.7109 6.5177 1,9555 5.O01.0 2.3331 3.1226 2.3036 2.9521 0-5.. 2 0.0- 20.0 t. 0580 6.0560 6.7552 6.8551 2.050o *.0107 ,.050. 2. 1272 0. 3769 3. 1216 2.2009 6.15053 0.2-160.0 5.622 5.655-9 6.17119 7.0712 6.7605 6.2620 1.6923 1.0022 2.5+975 2.95229 2.6011 1.2.6 0.0- 112.0 6. 1556 5.6935 6.1797 6.5515 7.21.53 ,,oo 2~J .0361 2., 222 3.500 1.5959 9.i565 6 5.5/2 6 6.1701 3.0213 1,76.3 0.1723 3.6~5+21.6-51..

8 62.0- 20.2 *_. 065 5. 7320 5.053)7 7.0255 7.0110 -,. 106 *:.0125 1.3255 0.71 1"3 136 3 55.o- 60.0 5.70602 .0628 2.9776 1.00o1. 2.5752 2.1017t 5,555 11 7962 3.0609 6. 1505 5.0675 1.7232 1.6105 2..90 ,; 0.266 32 032 .0-155.0 .4.os. 5.12o7 6.3206.7091.0151,1 2.5255 2.0695 . 6. 15 3.012. 1*1.3 5.5769 5.07 6.1.6 6.55197 .05 6~1 1.7155 5.010 0 .51 .11 202.4.o-11.1 )I.1 60.0- 20.0 s.7.6 6.6722 6.62?+7 *.iss 9.o357 1.2072 5.1120 9.7062 15 50.0. 60.0 1.6 5.16 7.0077 1o.61o 1.2756 5.,1.1 3.7105 6.520 16 0.0. 50.0 5.0563 5..662 7. 3892 31.8576 1.756 2.1395 713 0 .0125*.17 187. 5-010.0 3.6913 1+.6TO9 6.3100 3.3525 1.6551 0.0235 2.562 18 125.0-177.5 1 .19l.11 5.3667 3.4736 5.9877 1.2763. 1.1.815 1.9277 2.2150 09 112.5.1.35.0 3.9385 .25 524 .591166 160 .56 561 1.-5. 1 67.5- 90.0 9:.19 1.5092 4.5054 5.6655 6.8639 6.3765 6.2170 1.5515 1.7202 3.7536 1.5573 2.611 23 ,0.- 25.05D 5.6001. 7.35143 11.16023 1.7505 0.72 0 1:.1570 12.25'0 26 153.0-157.

5 3.5258 5.7106 5.8529 51.718 1.2016 1.7152 57 112.5-1.5.0

5. 552 2. 5657 1.82 2.0o25. 9 1.137 1.1367 6.7217 2.6232 32 15.0- 67.5 5.11.92 1.50o02 6.6516 6.2112 1.31.32 1.7870 5.5Z061 6.1.255 31 20.5- 5.o 5. 3066 712 7.4570 15. 57571.o .L 1076 52 0.2. 20.5 2.07 5IT .5837 0.0211. 15.6 1.7 1. 506, 3.2151 1.00 5, 0.7556.9 35 165.0-020.0 2.631. 2.1966 3.5050 .6611. .0537 1.555 1.2397 34 155.0-163,0 3.251 2.6515 3.51515 5,5015 .6372; .6555 6.1556 1.6204 35 135.2-155.0 2.2259 2..1537 5.5257 3.1516 .711l.1 .6135 6.1151 1.93253 36 122.0-155.0 3.160 5.3.162 3.31.50 5.71..10 36 120 202 27 955.0-122.0

2,0710 2.0599 2.3.214 3.54 0.5769 .7615 .2157 1.17 3.35,O7 6, 38 20.0-103.0 6k.0159 2.2706 2. 7706 5.6s1o1 14.sox .72553 .7550 1.5095 1.5537 2.7010 60 .6-77.6 59 75.0- 90.0 2.1.57 2.6761 1 91 6.1131 6

.0112 .7835 6.1200 6.016 3.723. :k.coTo 10 60.0- 75.0 2.1t719 3z. 606 1. o16 7.5955 1.0125 6.1.012 2.1133 1.5050 41 1~o 062.1670 5.1039 5.025T 9.5597 .9793 1,53-90 2.9164 6.8533 S3.3 , 22.- 1.0 ~ 2.s601 2.6170 6.3645 12.11255 .9302 1.6891 1.030 51.9030 S15,05.0 30.2 2.03 7.05.10 15.0022 .0866 6.658 1/.0537 1751 II 2..- 15.0 2.5059 3.7160 7.5061 1"7.2589 1.1235 5.21.2 6.1 5.2097 1. 10.0-165.0

.6556 .761.3 1.2862 2.2572 .0535 .2.57 .3727 3.5970 08 90.0.105.0

.7116 .6125 .0167 1.5501 5.0618 2.0502 .2536 .1201 .-11935 .6755 1.5195 5.11.900 77.6-90.0

51. 75.3" 20.3 =6. 328 .6558 1.7757 1. 1357 5.6699 2.O0545 6E. 097 .2060 .3933 .725 0.2523 9 .01.4 52 6o.0- 75.0 .6790 I0 1.073 07900 0.*3076 .456 .3926 .3602 5.2155 53 1'5.0- 61.3 .7515 1." 01 2.0595 6.5575 .5167 ,6655 61.675 5.3678 54+ 30.0- 451.0 .7226 ,1620 30. 19219 .509 .0577 1.9128 9.16990 55 55.0- 50.0 1.1135 1,5215 2.5320 12.5765 .109L39 .251.7 2.73.51 56 0.0. 15;.0 ,. 0129 1. 5.0156 .l 3.0756 Intoe:l 1o.0 A1684o0. .36.2097 o632.6255 55.91.70 1.7.7165 6.6155 7.2759 1.65o9 35.7205 35.2072 15.6507%yuSernioa source.. so n1 aeersged Cu 3-.Lu 135".3 J.'C.-I-C)0:3 looldoot Inord8n 0j 1 90.0-190.0 1.3792 1.2130 1.52510 2.3172 2.2958 2.1525 .5272 ,5519 .6587 1.9675 1.5452 1.0565 0.0.-45.4 2 0.0. 90.0 :. 1085 0.4151 1.8977 2.4498 2. o65"7 0.1131 j.0445 .5827j .8154J 1.0088 +.0ons a.-za 4 0.0- 90.9 8.0715 1.52977 I..6nI1 2.62s6 9. 1221 j.5120 *522 .8011 1.0085 9.0179 5 153.9-100.0 i t.276 1.5562 2.0074 .4917 .0020 .7905 1.15O7t 6 100.01-t5o.o 1.0954 1.406g 0.5570 2. 1000 .5598 .9995 1.3971 Ia 7 82.0-120.0 1.1509 1.1671 1.3091 2.7098 2.156 .479?8 .6025 .9819 1.607o 1.1102 09 30.0- 60.o 1.2709 1. 5i46 5.7504 4.298Y0 .5941 .7720 1.r5035 2.9785 19 0.- 30.0 1.5196 1.9399 5.1052 4.7"o32 .62:5a .7001 1.60 2S.3791 11 150.5-190.0 1.1512 1.2303 9.5564 9.077 .324 -590 .6758 .9081 12 121:.Z-152.0

.9574 1.2022 1.0540 2.1020 .4909 .5705 .7769 1.2527 O, 13 9..0-120.0 1.1139 1.0226 1.2921 9.10s 2. 5550 0.1340

• 4954 .45s0 .6202 1.0531 1.405g IS 3'.0" 00.o 1.0659 1.7002 3.0009 9.9080 .508 .7005 1.6281 2.1771 Ic 1.0.- 90.0 2.1599 3,7278 9.69 25 7.0O806 .0260 2.14 4O~l I.8090 19 139S.0-157.5

.9069 1.0317 1.0798 1.6790 .3738 .42ia .5o51 .8154 I) 112.5-125.9 1.1771 1.5228 1.95o45 .4213 .4693 .7959 93 30.9-112.5

.9122 1.0373 1.0435l 1.7373 0.007 .3760 .4318 .,087 .8100 1.3045 51 67.5- 90.0 +/-.0000 1.0146 1.3435 2.009"0 9.4307 8.0494 9.0107 .409 .5950 1. 1142 2.7150 -a 45.0. 61.5 1.1600 2.6902 2.Osoo1 4.4925 ,4459 .7377 1.8735 2.9009-33 2a'. 45-I.O .9707 1.7171 2.5322 7.9551 .4061 .8072 o.22i6 2.1611 01 1.8- 22.5 1.1750 1.9935 3.0007 11.0427 .5145 .0970 2.288 7.8425 1=5 157. 5-190.0 .601i .0137 1.0616 .2598 .2698 .4979 .79007 14 125.0-157.5 0.0910 1.5215 .0009 .2365 ,50o5 .8026 27 112.5-1-05.2

.7570 .9009 1.2150 1.9799 .2067 .3789 .5206 .981l0

.7406 .7099 .8506 1.35g4 2.10502 2.9046 .ssia .3238 .4591 .761o 1.5894 1.4309 59244J..0 09 07.9. -9C.o .77623 1.3058 2.0294 2.1052 9.0092 .3559 .9297 .9141 1 .7).59 8.1290 01 22.5- 45.o .9608 1.o124 9.4070 9.4818 .s'rro .9316 029015 6.a6a2 03 105.0-183.5.

.4; 0. .4000 ."624L 1.12132 .1810* .05543 .07 .3959 1'7 155.0-122.0

.5002 .5283 1940 1.5907 .2119 .2521 .4537 40 6..75.0 .7799 1.0019 2 s6 4oi 20 .835 2.6061 42 2S-0.0- 4.o .70 913 2-5.324 11.1040 .564 .52 .4rs 789 42 19.2- 20.0 .7260 i.G05s 4.1034 17.2134 .2903 .8317 2.8571 18. 59225 45 165.9-180.0

.2197 .1537 .2230 .6492 .0612 .0667? .0687/ .0055 46 190.0-1445, .1502 .i065 .2105 .7564 .0453 .0607 .1044 .3579 47 135.0- 150.0 .1579 .142 .2290 .8177 .2722 .0600 .1175 .0083 48 100.0-11.5.0

.1559 .1359 .1392 .902 .096 .0954 .1572 .3404 49 105.0-100.0

.2255 .1566 .4122 1.8017 .0765 .0757 .1826 .5771 77. 6-90.0 59 75.8- 90.0 9.9*07'9 .2278 .2g14 ,560 1. , o.0420 8.006(1 .0758 .1013 1 ,oo2 =9.0927 52 00.o- 795.0 .2749 .3590 .6221 5.6954 .1012 .1811 .0448 1.7108 52 45.0- .0002 .3933 .0001 4.9210 .1000 .2478 .55192 .1069 lao oner Y ' 0.5977 109.2228 112.570 16.00290 18.7845 7.9942 59.8734 213.9090 69.7339RDo.. Ailoodoc 4.463 6.7/990 12.2559 29.9346 15.630 1.8052 2.1007 s.14o09 6.6100 22.5993 7.026 lyrotttol 0aoce.e 02 11 ooaluesver0std tn 0 0-Va 0+/- 25 (503 co-u Co IC o -x CDC M 0)m r C)03 03 C 0 03 89 0)z C)C):54 r C)C r:54-I 0 z ('7 z I 03 0 N)02 C)03-A -C),;,;

OMI RAY 9001 8021109 (In porcnnt)Etenoglg OEgr~on Water Conoonne p'olar 6.13 eOv 8.2 NoY Anglo looldoon at polno

• looldant St PFent *IIf, 90012. 25 .309 .33 .688 ...688 .6976e 8.2932 4.24 .73 5.3lD 5.287 5ore. ,, 8.-24 8 00 ooB24 36 49 76 .71 e,.o181 g.2 3.8388 6.1261 6.8688 *t. 651 9. 3 90.0-480.0

.2;5"78 .244d" .2894 .6886 .9588 .6237 4,.1898 4.3431 8.3*651 5.0883 5. 5"53 5.2082 5152.80-180.0

.2018 ,5178 .3825 .33024 4.,iUO 4.7218 5.8179 6120.0-150.0

.2237 .2190 .5678 .9718 3.8218 4. 5280 4.s08 5.1188 2183. 60.2- 82.0 o.007'8 .2423 .588 1.2751 4e.0124 _..0022" 2.6716 4'.1859 5.2787 6.1243 -o.1175 89 2o0.2 63.0 .2318 .8763 .7833 1.6718 4.2174 4.25225 6.1098 9,l.112 12 30.0- 30.0 .2057 .5197 .0422 I E2.003 8.0747 8.1884 5.9747 8.1123 it 152.0-180.8

.2181 .o66s .8442 .806-6 3.2770 4.1608 8 .61 5.6075 32 120.0-150.0

.2173 i t :192 .6096 3.8652 2.6887 3.1 8.9097 6.162 5.2 8s 12 20.0-130.0

.220 .2222 02 .6 3Fl .18 h718 67.o. 23;.8 8.°0119 .2020 .2272 .7222 1.7.873 2. b017 t.61 2,42 3. 9545, 5.221 6.7826b 8.2826 1 5 3.2- 67.o .2593 .397,8 .0,252 1.9868 2,2757 83.277 5 5.7968 8.2271 16 1,0- 30.0 .3636 .8196 1.8125 S.ll5 4.0812 4. 5738 6.91829 18.6'92 17 137'.5.180.0

.15G0 .2355 .3ay6 ,.10s. 2.2l05 3.27 4.00 60495 5 .86cc 2A 335.0-157.

5 .1759 .1988 .3241 .0713 2.1533 1.162c 8 .7.68 3 .2382.9 132. 5.132.0 .3036 ,2692 7327 2.28823 8 .3318 8.7437 88.8.-5.2 23E 67.3- 90.0 6.0359 .0220 .2325 .4800 1.3248 2.8082 .03024 3 .15417 2.8667 8.,5(5 651726 8.1902 22 8,0- 67,5 .2187 .2095 .7226 1.2876 3.3082I 8.9238 5.8388 24 0.0- 22.2 .2118 .4789 1.0819 2 .1882 8.9771 6.2274 15.8200 25157.3-180.0 .172 .3115 .61o1 2.2108 3.160 3.9808 8.4252 26 11.3-157.5

.0132 .1202 .2012 .68,31 2.5306 2.9224 2.3702 8.8886 87 112.5-135.0

.1850 .2t22 .77 2'5 .2088 2.9708 3. 4301 8.3944 1 28 0.50-12.5

.1 * .1314 r5 416" ,4u0 .8762 .73721.5E 2.8167197 313 328 s'.16 8.83 5526.8 8 710.0-190.0 8.08 .1785 .2525 ,535 1.3451 202 600 2.4271 3.1553 8..oczc 6.8551 .85'.o-16i.

0 .0 .1431 .2833 .2689 2.71365 35 1,C3 2.821 8,6288 33 20.2- 4s.o .1768 .8152 .1 1.0312 .2478L.4682 3 6,2030 12.827 22 0.0- 92.5 .1057 .3087 1.3012 920486 2.0c5 2.943 15.7236 22 965.0- 180.0 .1o03 .0873 .36559 .5855 1.6988 2.29712 2.8418 34 159.0-105.0

.0813 .1823 1.88302 1.5812 1.918 2. 1 8.4477 12 13.0-150.0

.138 .3172 l.25809 o .8355 1.763 1 .9715J 14.8298 86 120.0-1950.8

.1124 .1620 .208O .5290 1.5987 1.055 2.7871 4.1273 27 105.2-128.0o

.1236 .112 .217 i 2.57 2,282 8.8542 8, 539 05.0-125.20 08 .015 .1617 .362 753 .853 1.22 ,7652 1.982= 2.7899 .727 3.01 5 6867. 3 75.8,-905.2

• .0,227 .1062 .1280 .3381 1.4250 t.0337 ,%-22 1.5o8 2.213 2C.6 574 .8886o- 79.2 .1088 .1487( 1.4822 76023 2.418 3 .8185 6.91 ,11 8143.0- 6o.o .0867 .1214J .5068 2.8067 .5278 2.2082 8.7597 82 30.0- .1039 .a840 .7292 8.9018- 1.766 2.7071 8.6227 12.1898 43 1.0- 30.0 .3018 .2768 1.237 12.382 1 .8128 3.0786 5.5393 18.22223 44y 0.0-a 15.3em .1585aue vead.57 1.52 3.65211 3.48 649 1.02 45o 165.0-192o0

.050 026f.508 .220.371 652 .801 2.8e H 90 0'I.-(2 (on U'La H a 25 rt if 613 a."-A oz o,-x Qo (DC<zN a)m-r--0 On; Fberging Ooeolot to 8 Concrlete' Polar tzeconion

.60 1.00 358 Angie O~tidont sc Point *, pi 9. 1 33. 0-183.O 1.7035 1.7597 2.27r37 o.6uio 3.3631 2.6836 1.i41.o 0. 135 1.4J476 2.1105 2.3946 1.9354 Ga 3 3.-0. 1.16 .70 1.16 .111 2g,4 .56 1.72 Ii6 1.11 2.11096 2.03 2.6906 15.4-21.8

1. 0.0- 900 soo .33O207 .1203 *4151 o.tO44 ~ 4a 112 1..5032 0.1.1.2 3.653a o.631.6792 2.499 2.7653 I.36lO 1.1243 1.5225 1.8918 1o .3908 0.8752, 2.1.3. 9 .061 1.1944 1.7291 , 292 6lu 1.81 2.65 2133 p.662 .232 .0827 12 .01 .8 192 21.8-.24.

8 6o.o- 33.0 8.0230 1.7327 2. 1599 2.9783 5.0533 s.t009 3.1299 1.25662 .244s1, 3.2331 0.069 9 33O.0- 6o.o t.756 2.2596 2.1876 5.3331 1.2380 1.6173 2.3o1.L 1.5625 13 0.0. 33.2 1.6595 3.1156 6. 27r2 I.o96o 3.0e99 5. 1607 11 159.0.1- .1.o50 1.5202 2. 2361 0.417 .9731. 1.029 1. 103?E.I~ 1.79B 12 129.0.5-10.0 1.3, t.-5619yo 2.103 3.4006 .956 1.376 1.18353 1.7 1E0 8o. 13- 337.0-129.0

,.1056 i.o58 .3 2.33 .8231 0.2.82 .6030 1.0161 1.322 1.9s56 23.36 2.15728 12 £.0- 60.0 1.55025 2.1703 3.6205 6 .837°7 1l.1098 i.7611 3.812 16 0.23.0- 1, .60356 * ..5o1. 4.32965 8t.7624 1,o94 1.60,6 337 63.19 70.6.20 171375-8.0-

.1.1838 1.3731 i.54o 0. 1O23 .6224 .933y 1.1733 1,568,S 18/ 333.0.1375.

1.336 1 .3302 1.7560 2.1680 .7003 .8755 1.2353 1.6217 18310.5.0135.0 1td 2.154 2.3601 .77275 1.9433 7.163 1.8533 '*J os 3 0,0-1135 1.75 1.231 1.331 2.128 2.123 2.13 .63 .7826 1.t121 2.71.84 2.53139 .1445. 7 .50.0-29.0 037 1.41.i 1.7656 3.553 101 *34 .o1.s 1.2366 2.0337 3.34 o7o7245. T S, r :O7 1.276 2.1503 2.8730 325.70 .6B .6o54o 1.5070 2.3981 4 .6 2 22.5- 45.O 3.174z3 21.972 4456 2.5.907 1.203 1.79067 3.*4167 72.803 24 0.0- 02.5 4.1.29 2.571 4.67t .73 1106 .20930 1.70 4.3559 196705 25 135.0-129.0

.050. 1.0730 1.576 1.8492 .6~o .6568 1.3170 36 133.8-131.9 1.2201 .o95 1.4789 .6128 .7157 1.1295 1.530 27 112.3-135.O .2900 1 .358-5 1 .5008 2.2072 .13 .6449 .6520 1.1777 1.9185 133 8o 0..12. .9512 1.083 1.330 1.903 2.26. 2.771o.57 .36/7 2.6247 .25 5522.6 2 75,. 33.0 o.30 .9689 2.4243 2.2393 4.2i *.0965y5 .7205 .38333 3.2112 o 7 24 15,0-165.0

.5931 .7933 l.1722 1 .635 .39618 .6524 '.7336 1,27?33 1.5.0-.350.0

.7038 .6536 12.333- 15.820o ,33o .1.23 1.,353 8~t~ go 6 20.0-105.0

.6a04 6.795 a.623 92.76 h.451 .6.31 .2l.3251-3.475 3E.09'3 m b 33.0 .:20 1..1S32.4 C 2'.0.l5.3

.4t .7 .gO7 ..K .7 l. .61g .. 0.623 1693 43 03 Co I-0 Ln rt 0k Is Vi La--t co-'C 0)m C, z C)I r C/)

(D coT Co C 0-~(DC C N)0)m r C)C 0 ci)C', z 0 C-)r C~)C r-I 0 z Co z I 0 N)0)0-' r 02 n CYi7 o.orgtoo Isoargiog lroo It rronaIo Polar llrotLion 0.20 9V o.06c nor Angl Inetdoat on tooldoot at 95 flk V 501 1; 605 or POolrt P* 2 8B roba..Source Source Ga 1 90.2-180.0 1.5372 1.496 .2.0526 2.6779 .5.226 2.4796 1.001 o.on;o 1.4037 2.0759 1.963 0.o-t5.4 2 0.0- 30.2 a. 1032 1.6439 2.0055 2.6235 3.0449 0.0182 o. 201 1.1199 i. 3665 2.3952 2.404 35 8 13 0.0-180.3

.50 1.5593 3.9621 2.3042 3.2706 2472 102 1.9663 1.312 2.1290 2.5932 1.9 5150 .0-10.0 1.469 1 .02.17 2.4oa3 3.5782 1.1427 1.2839 8.52C32 1.250 1 3 20.0-150.3 3.464 2.3335 3.1257 .2763 1.s7 1.87 2.2686 Do 75 20.0-132.

0 1.212.3 2.148 1.914, 2.1633 3.0681 2.661i6 .272.0 1.124 1.2351 1.5sa3 1.920 238-4. 601.0-10.0 j.01 1.5015 1 .9526 2.39"4 1 .3021 0.22 21 .0510 1.3008 2.1106 1.8514 .0 19 12.9-0035.

1.14t4 8.8930 2.5023 3.741 o.10 8.846 3.0375 4.671001A'.0- 30.0 .0 .3.5o4 2. 3 1939.282 3.42 5.068 .(2 0 -0 1 .192 1 .69139 32.132 423.71 91 950.3-380.2 1.4z4 1.9211 2.0594 2.2051 .35s 1.1039 1.30 217 12 20.--150,0 1.2006 1.6594 2.842 3.21,3, 1.8051 1.1654 15.641. 2.2273 8 ~ 1

.32 1.19657 1.9043 3.0.71 259 25 1.19,0 1.,2939 1 .0343 .715. 26. 14 85 1.34'94 i.164 3.0712 a.05g .1112 1.7646 2.7042 1.2700 1O 0.0-130.2 1.632 2.1163 3.440 1.320 1.0732/ 2.3511 1,(.502 17 153.5;-38.0 1.301 3.4394 1.9701 3.2361 .9523 1.8902 9s 30 30.0-112.5; 1.148 1.10o 3.3 ,174 3.50142 2,4700 3 .945 1055 1.28"8 44 8 2. 3407 32 5.o- 67.5 1.0260 1 .2302 3.0289 6.43395 .9194 1.9639 2.1797 230513.5-30-,.3

.69 1.7396 1.7142.. 2.8"374 .ss'r .9323 1.7740 26 33.0-357.50

.960 1.6340 3.0008 .0o21 .8374 1.355 1.6750 27r

.960 1.2937 1.5515 3.6351 .931 .32 32226, 90 12 90.0-21. 1.273 .8955 1.34332 .0747 36.507 2.38.0 .6222 1.955 .2979( 2 .6589, 2 .38 552-4. 2 67.5- 20.20 .3 1.3073 1 2.2097 4.0 .0791 *.213 .65 1.6899 2.084B 317.05 0.74 31 2.5- 45.o 1.270 19.459 .882 3.0520 3.7375 32 22.5 3.9402 2.0640 5.373 11.8493 .972 i.7i6= s.sn6 1.890 33 165.0-280.0

.6655 *t, 1.417 2.2463.377

.4764 .755 1.3213 04 150.0-165.0

.06402 .6122 1.1012 2.2640.50679

..2032 1.440 27 105.0-120.0

.930 .2275 1.6010 3.2051 .5163 .61s .330 2.8647 82 3 90.0-105.3 .Oas 10209 .337 1.2013 2.92 2.03 .35 .646 .7633 3.187 2.9364 .15 6467. 9 75.0-. 90.0 4.06 -.7162 1.0"39 1.52 8q 37.2074 .75 +/-0 .2i94 .09237 1 .6226 3.03720 ~a 60.o- 75.0 .7262 1.1760 9O. 601 .4936 .9097 3.73 6.e 44 ve 0.0- 15.2 13.5542 4a.32078 13.425 13 .35 1,.7711 1.7327 202.10311 1.9 4a 05.-19.0a et~ .200 216 .333 8.79es3o6

.144t222 .00 pn c2 Co 3-.mu IA to)-IT~20 Ion COT CO IC o -x Qo (DC C M a)m r C-)20 5-a 0 n (33 C,)z C)C)r 0 C 5 z C,)z I CA, 0 N):1, 20 03-' n.-09 n-a -C,,?;

Ht 2)CD~Cot Co 0 -~0< N.)0)m I-C~)2)C-)2)0 n CO C,)z C~)C)I-C)C-I 0 z C,)z C~)o:3, 2)C-)p"Cp Eworotte Eoc~io Stan( it

~Polar Sireotion 6.13 lbS 0.20 KaY AOtOe ineldosto an i ncidooc at Source Soarco St 90.0-020.6

.2939 .5583 .5065 .SS4S lt9 1 t .6093 .0756 .3795 .6939 .1.396 .5045 .5l 4 9 1 t 0 0.0- 90.5 *.0573 .2927 .6636 .aii6 1.626 .1136 .1302 .oc1.6 .3079 3- 90.3-153.u

.0.350 .0073 .5.301 .9101 .3562 .0130 .059 .0289 .4003O: .1699 5 100 .0-480.0 .051.0 .5629 .7646 6.22308 .0692 .003"0 .0145 .250-1 6 130.0.1.50.0

.3561 .6067 .9909 i.9146 .0500 .3471 .s047 .0353 6 o 7 53.0-120.0 .036 .5307 .9566 1.2965 ,6363 .0655 .0652 .1290 .0993 .2669 .2135 21.6-34.6 8 6o.o- 00.0 9.6215 .05635 .657s .6530 1. 5"67 t.0i57/ 9.0095 .103 .1039 .00.65 3° 55.0- 6o.o .5002 .59s5 .9703 1.50o .0630 .05 .121 .53.0 35 0.0- 30.0 .0909 .6573 0.0307 .301.0 .099 .2261 .2090 94 13 37.5-151.0

.2763 .46253 .7091 0.00o9 .9059 .06576 .0369 .4004 .6105F 34o4 .4 i 60.0- 55.0 tO5103 .0106 .076i2 .299 1.76(80 +/-.630-4q 079 .0310 .5096 .2930 .650 9 .0139 55 30.0- 6o.0 .006s .955o 1.030-7 .0607 .o66s .0714 .61.43 56i 0.0- 30.0 .030 .533 1.076 2.o767 .0916 .07668 .3160 o7937 17 157.5-08.6O

.3073 .4177 .5337 1.3516 .oao ,.1136 .0710" .0269 26 135.0-157.5 .4537 .5790G 1.2611 .0027 .0506i .17/03 .3099 23 05.0. 97.5 .3470 .5771 .ks.0 2.006? .0757 .1790 .2308 .8026 23 3.5- .0021 .5066 0.0553 3.0951 .05i2 .0956 .6767 .9641.24 o0 03'"2.5 .4100 .6563 1.0356 .131.7 .1511 .253 1.0693 0'5 1075-635°0.0

.1002 .1300 .3008 1.2535 .50639 .0733 .1017 27 112.5-122.6

.31J23 .0771 i.0,6n7 .031 .106 .0500 .3921 As 09 ,0-6111.o5

.2569 .2102 .2553 .5030 1.i6{,9 .6503 .0711. .0060 .0615 .044 .0010 30, 025.- 675. .36415 .7604o~ .1065 .11 .255 31 2.-0. 659 .30 195 3.9179 .6814 .0634 .0307 .7621.32 0.0- 00.5 .5"139 1.4453 y 9.i991 .0310 .2365 .1167 1.3083 33 S5.0-So.0

.1613 .1640 ,o5 .6593 .0061t .0913 ° .1096 34 050.065.0$

.1655 .0063 ,o1.s .9559 .0052 .1031 .6356.35 135.0-050.9

.6606 .1766 .168 .57T00 .0630 .816L7 .0010 .3501 36 120.8-135.0

.00O37 .3120: .3130 .5991 .0792 .0421 .6192 .1.761 37 105.0-100.0

.00.00 .1]359 .0.49 1.0213 .6557 .6034 .01.60 .3000 87 39 so.0-1055.0

.1613 .1355 .1763 .0Z558 0. 1033 .9193 .061 .0605 .10000 .1005 .2131 60.9-7/7.6 35 70.0- 60.01 9.00.3 .1647 .0747 .5303 1.3510 -. 0506 9.0 075 .91.67 .0109 9.6376T l0 60,0- 75.6 .1931. .0642 .6236 1. 76, .3412 .0790h .66s7 .011.20 05.o- 60.0 .30I31 .3951 .7668 2. 4.0500 .1271I .1713 .65 02 00.0- 45.o .2738 .4o.I 0.wO01 01.3537 .0933 .1890 0.00.j1 03 15.0- 30).8 .2339 .066 80.131 .06so .0939 .3112 5.5"00 21. 0.0- 15.9 .90s6 .0562 15.003.5 .080 .3162 .230Z5 o.56a08 45 .165.0- 130-0 .0656 .0395 .0899 .o673 .0570 .0596 .0053 .230q 47 125.0-15.500

.602h~ .0730 .i]O4 5 .5333 .01.76 .3256 .0030 .6390 09 203.0-033.0

.6972 .1630 .1]060 .4479 .0061 .809..f .6156 .1270 Ig 105.0-1.0.0

.3490 .091.6 .1335 .2855 .0002 .031.0 .0507 .1369 0s 3') 3,.0-105.6

.66:23 .0070' .u770 .1206 .639 .795 .06l 5.1 .6326 .317 .0056 ,0075 .2056 77.9-53.o 51 75.0- 90.03 ol.13S8. .0726 .1101 .160 .736-3 0.0581 9.0001 .o0016 .23,3 .O076 .3533 9.30123 52 60.6- .C660 .o193 .060 1.0000 .0230 .06, 009 Ja5 23 05,0" 6o.o .o1.1. .0064 .3647 .0451 .0336 .6580 .9781 53 15.0- 30.0 .3042 .0303 .8509 6.2307 .61.36 .035 .2977( 1.6197 56 0.0- 15.6 ,3032 .s345 1.2309 9 .36,543 .0035 .001.9 .2613 1.3 5375 Ott Otttt 61 00.'500 0.5.10392 09.8102 11.9.0573 49.1275 0.8500 3.2531 2.2971 0].346 06.5300 16.1923 Totals Ooe Albndos 1.6133 1.7693 0.3200 2.3350 16.9i13 5.514.3 .334 .06550 .o161 .6361 3.23418 1.1&00 hSymncrlca.l 8000000. so voloes soeraged Lit GD O rt Lit!a-It~ 23 (Sn cot CD IC o -x Qo CDC< N)03 m r C*)03 5.a 5.0 n 99 C')z C~)C)r C)C r-I 0 z Co z z 03 0 N)02 C)032 gco n-x -

0880PAY 0000 608960 lie patrontS Oomrglo8 Lead [.lsAd Polar Direction 0).060 I 1 .OO 8 AoE0o I olrolen 80 £niridro:

• 16' 8602 6 g Pit__ __ _ _ _ _ __ _ _ _ _ __ _ _ _ _ [ 0 Ocr v.o-15.6 m0 61.1. -55 .3 lo 06.6-U,.S I 93.2-180.0 2 0.0- 90.0 1 9. c- E'.o. r.6 12L,..-15...

78 35.6-127.6 O St.!- 0 2.6)- 6.,5 I2 6..1- 30.0 11 0 10 0.o-165t.13 ..0-126.0 117 06.0- 20.0 15 06.0- 60.0 10 6:.0- 26.0 17 0 15,..10.18

.30 39 1 f.5- 30..02 63.t.- Er.0 23 33.5- 10.5..0 9.¶- 0;.5.1028 .0900 .10O28 .2369 .0085 .6008 n.005'7 .1105 .1075 .6611 1.585 p.00 1 8.0858 .1300 .3079 .6678.0583 .0701 .ito6 .o1.110 .01" t.80 .0802 .166o .7050 o.6in*.091 .0680 .6336 r.61..Otp6 *.056T .6618 .7340 B...6106; .0036 .2232 1.0936 .5578.003" .1502 .794 o.oooo .0980 .31565 .9305 .o563 .0557 .0560 .60665.015 .107t .0918 .7008.oso4 .0736 .11.0 .671I.0603 .001.6 .07/78 .3085 .6355 .610t3 8.090"6 .0837 .t303 .30)3 6.8167 0 0 1?.1335 .2130 t.5009 2.3033.1001 .3665 2.0108 4.5079.0507C .0618 .1013 .6637.0790 .o0;5 .5505.3631 .Ol6a .60t9 .7?90.6o .Oso6 .1116 .Otto .92 .81.06.1009. .7667 3.9636.053 .0337, .06e1 6.02917.0631... 0130 ,.0167 6.6016:.0691.0771. .0651 .1716 .9367 .6,208.01.53 .0750 .2391 .91o3 .3306h.0789 .1087 3 .06. p35.o66o .0780 .1813 t.6s 0 .o.060 .o767 .1553 .661*.o1. .0853 .30.8 .o1o9 .61..o061 ,11.63 .2581 6.,769 p.030O.0601 .1885 .5971 2.3657.os90 .659. .7809B 3.6031.0369 .0575 .ooS7 .56723.o769 .a1o1. .011 " .6608 .o336 .0066 1.1630 .0067.6072 .3781. 1.030 4o.9603.0O370 .01.67 .Ohl6 .1.99.0336 .6301 .0805 .5965.0E92 .0285T .6071 0,0651 ,1131 .300 1.1 7 .0707.01080 .3097 1.512 6.0011.062k. .0331 .1037 .3873.o0748. .o1B.3 .06o l .6066.0078 .0505 .0790 .7026 .13.0290 .071.2 .5537 .8078 *.0617.1302 .22 10, .30319 .o2.093 .3532 1.0773 '~0 02 0*Sb UT 03 a ft F.'.Sb a..141 Ox 3>QO< -.05 r.)03 a>03 (-'C'zJ C)Oar core T3.9921. 1..oowy 6.6312 00.3338 130.7993 36.1603 0.6715 3.6550 7.8303 09.0509 153.9901.

40.3708 IgloO Deco OIled,. 1.5806 .6700 .01.60 0.0600 16.6901 1..0553 .0807 .1.080 .6053 3.2606 67.0837 4.6869 erevocroeIoerco., SOW 00,002 avrragee SO values averageo Onarelaa trnrflegLea" (O percent) La Peter D~rectio 2.50 teen 6.13 tHns Angqle Innldnen at nnideant at lauren borne 0.0-12.9 0 o.o- 90.0 ,.tcO4 .261.1 .5917 .TA8 2.0325 *.n"6 9.0201 .55n1 .5015 .7c65 1.6155 *.0253 da 3 9.4 .0 .32 .59 .4 6 1 c1 .3969 1-.3760 .5706 39 .3658 .97'82 .6771. 1.5993 .8196 0.0- 90.0 :.l,04 .260T .26s7 *.06 1.81.17 ÷.o1o .3745 .14o6o .6263 6.O80.6 12o.0-150.0

.2762) .265 .251 1.5101. .2324 .1.1.25 1.362 O 25,O 0.0.68 .5172 .1012 .7281l 2.6510 ,.9746 .4eo7 .7553 1.0208.10 0.05 20.0 .2097 .31,50 .0787 2.9079 .9005 .8120 .8333 a 4 13 90.20-133.0

.1059 .161O .1054 .5358 1.096 .7sc6 .3171 .2392 .0197 1.3260 .7796 15: 20.0- 6o.o .2171 .278 .7080 2.6329 .& 751 173 0.0- 39.0 .1702 .5001 .7007 4.0172 .6902 2.5121L 17 157.7-190).0

.1255 .1958 .359 1.1t235 .2720 .2877 .1.813 2 .1600 16 5 .12197 .1837 .3611 .3554 .2787 ,6271 1.3820 18 112.5-125.0

.2169 .2501 ,.2090 1 .1815 .2477 .3554 .5974 1.4045 44I.4-55.0 21 67.5- 90.8 .1967 .23.2? .5179 1.6821 .2868 .3546 .5982 1.6678 ,.216 23 22.-5 4s%0 .29127 1.1071t 4.5478 .3952 .52789 .7898 5.33552 25 13.35-190.0

.1.710 -.1207 .38.32 .8189 .1532 ,.2076 .5229 1.246 96125.0-157.5

.1246 .2990 .9659 .o16o .299 .529~ 1.556 27 2 12.5-135.0

.1262 .1922 .2672 .3198 .059 .2655 .1.531 1 .406 Pa 289 0.0-113.5

.1202 .0952 .1548 .1.69 1.1363 .8703 .211, .1222 .1857 .9567 1.541 .7354 55,3-84.6 59 92.0 o:.0091 .1108 .5159 .1.99 1.777'5 t .0557 .331 .5778 .101.6 1,7929 So 95.o- 67.5 .1761 .2275 2.9'726 .26 .2505 .5513 1.7036 31 22.-5 1.5. .2155 .5521 1 5.2597 .3077'; .2.447 .71.18 3.0467 3.3 163.0-180.0 .C265 .1659 .2237 .1251 .257T7 .3581 1.1051"7 150.0-965.0

.2905, .1822 .1015 .9652 17 .104 .30(5 1.272..,9 50 3

.592 .5539 .2059 ,0027 ,.1905 2 ,1086 .2051 1.1391 675 36 122.6-155.8006

.16 .1220 2.702 .11.7 .9589 .21427 1 .232 275 1o5.o--12s.C

.1559 .1145~ .22552 .o4615 .C666 .12 .5711,Ed 1.117 4 e 3 90.9-105.0 111 .05 .9187,t2 .29 1.927 .931 .1508 .2145 .2339 .20(2 1.55 .739 646 k. 9 705.0-1900, .20 .8055 0h .16 .25 1so .0 41.1. ,.5-69 .1875 .51.37 .45-0 1.555 .2 4O 0 60.0. 75.0 .i64s,21 .1927 .5496 2.279 ,.9o,637 oJ .2171 .9037 1.7667 "'-00 1. 75.0- 1.5.0 :E C7 .z1.s .3310 1.i755 6.2233 .55 .0 .209 .56si .6151 ,2.8391 1.

.05'56 .2441 .0483 .2327 .1509 ,t901 .708,7 96 i50.o-165.o

.0577 .0122 l.o'-4 l.s04i .0597r .2256 .62 .2o 6 , 8713.20150.0

.0570 .2459 .9=227 .5675 .2005 .9503 1.55492 4o 0 oe loo 120.2135.0.9550 I.055o 3.t7200 L.52.4 5 n9.3 1.651. .1.921 3.1221 1.Ir.28 18 105.0-133.0 l mo .6670s .029 .05 595.L32 .57e1o1 .5 62 0'0 0e 0l-IT~ 73, (On tot (0'C 0 -o 2,20< NJ 07 m r C)02 5-a is.0 n ci, C,)z C)C)r C)C-I C z Cl)z I c-el 0 NJ:59£12'7-on-073 75 SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-1 CALC NO. SNC024-CALC-007E N E R C 0 N CALCULATION COVER ver proec Evr doy SHEET REV. 0 PAGE NO. I of 8 Til:Level for Initiating Condition E-HU1 rjc dntfe:SC2 Item Cover Sheet Items Yes No i Does this calculation contain any open assumptions, including preliminary z] [information, that require confirmation? (If YES, identify the assumptions.)

2 Does this calculation serve as an "'Alternate Calculation"? (If YES, identify the design [][verified calculation.)

Design Verified Calculation No. __________

3 Does this calculation supersede an existing Calculation? (If YES, identify the design [][verified calculation.)

Superseded Calculation No. __________

Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation r- Final Calculation Safety-Related EL Non-Safety-Related (Print Name and Sign)Design Reviewer:

Curt Lindner Date: i1) 7 v .Digealry signed by JayJ. Mai,~er, cHP Approver:

Jay Maisler, CHP 6...........

Date: 1 0/9/201 5 SMN H-I13-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-2 CALC NO. SNC024-CALC-007

~jEN E R CON CALCULATION RV Excd,, .... REVISION STATUS SHEET RV PAGE NO. 2 of 8 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 10/07/2015 Initial Submittal PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-8 0 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO. OF REVISION ATTACHMENT NO. OF REVISION PAGES NO. NO. ' PAGES NO.

SMN H-I13-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-3 CALC NO. SNC024-CALC-007

~E N E R CON TABLE OF CONTENTS REV. 0 Excdllence--Ever project. Erery day, I_ _ _PAGE NO. 3 of 8 Section 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 Purpose and Scope Summary of Results and Conclusions References Assumptions Design Inputs Methodology Calculations Computer Software Page No.4 4 5 5 6 7 8 8 SMNH-13-021 Attachment L SHEET L-4_________________

ENERCON Calculation for E-HU1 CALC NO. SNC024-CALC-067 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 F~E N E R C 0 N FOR INITIATING CONDITION E-Exo, o0-E~ ry d,.o~y~ HU1I PAGE NO. Page 4 of 8 1.0 Purpose and Scope The purpose of this calculation is to determine the emergency action level (EAL)thresholds for the initiating condition (IC) E-HU1, which is defined as damage to the confinement boundary of a storage cask containing spent fuel, as described in NEI 99-01 Rev. 6 [1]. The IC is defined as an "on-contact" radiation reading greater than two times the allowable dose readings as specified in the technical specifications listed in the cask's Certificate of Compliance (CoC). A dose rate reading greater than EAL threshold value indicates that there is degradation in the level of safety of the spent fuel cask.This calculation is performed under guidance from NEI 99-01 Rev. 6 [1], which describes development of a site-specific emergency classification scheme.2.0 Summary of Results and Conclusions The emergency action levels for initiating condition E-HU1 are calculated based on the HI-STORM 100 and HI-TRAC 125 cask system technical specification for spent fuel cask surface dose rates [2]. An elevated cask surface dose rate is indicative of degradation of the cask confinement barrier. The calculated elevated dose rates used as emergency action level thresholds are provided in Table 2-1.

SMNH-13-021 SMNH-1 3-021 Attachment LSHEL-SHEET L-5_________________

IJF-I '.r-r .,Jl~ L,¢dIL, U cIdLIUII IUi L-I-U I CALC NO. SNC024-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0~'E N E R C 0 N FOR INITIATING CONDITION E-ey ..... .'aory HU1 PAGE NO. Page 5 of 8 Table 2-1 Emergency Action Level Spent Fuel Cask Surface (Neutron + Gaimna) Dose Rates for IC E-HU1 LocationEt j(mrem/hr)

HI-TRAC 125 Side -Mid -height 450 Top J 110 HI-STORM 100 Side -60 inches below mid-height 80 Side -Mid -height 80 Side -60 inches above mid-height 30 Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 3.0 References

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors." Nuclear Energy Institute.

November 2012.2. 10 CFR 72.212 Report. Edwin I. Hatch Nuclear Plant Independent Spent Fuel Storage Installation.

Docket Number 72-36. Revision 17.4.0 Assumptions There are no assumptions made in this calculation.

SMN H-i13-02 1 Attachment L ENERCON Calculation for E-l-SHEET L-6 tU1 CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 FIE N E R C 0 N FOR INITIATING CONDITION E-£xcII, .. ~eydao HU1 PAG E NO. Page 6 of 8 5.0 Design Inputs 1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask system technical specification

[2, Table 6.2-2] are provided below in Table 5-1. These source values are scaled to develop the emergency action levels for initiating condition E-HUI.Table 5-1 Technical Specification Dose Rate Limits (Neutron + Gamma) for HI-STORM 100 and HI-TRAC 125 Loctin Number of Technical Specification Measurements j Limit (mrem/hr)HI-TRAC 125 Side -Mid -height 4 224.9 Top [ 4 52.8 HI-STORM 100 Side -60 inches beiow mnid-height 4 38.9 Side -Mid -height 4 39.7 Side -60 inches above mid-height 4 15.6 Top -Center of lid 1 6.0 Top -Radially centered 4 8.4 Inlet duct 4 72.0 Outlet duct 4 18.6 SMN H-i13-021 Attachment L

fnr 14 SHEET L-7 L.-I '41.1 .-I\ ,..JI~ '4 JClI...UIC l.I.JI I IUl L...I IU I CALC NO. SNC024-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0E N E R C 0 N FOR INITIATING CONDITION E-da HUI PAGE NO. Page 7 of 8 6.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-I100 and HI-TRAC 125 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6[1], for use in initiating condition E-HUI.

SMNH-13-021 SM NH-I 3-021 Attachment LSHE L-SHEET L-8l k..I kS./lJIl HUH L.-1I IL.., I CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0E N E R C 0 N FOR INITIATING CONDITION E-,t £~erday HU1 PAGE NO. Page 8 of 8 7.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose rate limits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits (Neutron + Gamma)Technical LoainSpecification Scaling Calculated Value EAL'LoainLimit Factor (mrem/hr) (mrem/hr)________________________

j(mrem/hr)

____ ________ ______________ ______ ______ ______HI-TRAC 125 _ ______Side -Mid -height 224.9 2 449.8 450 Top j 52.8 2 105.6 110 HI-STORM 100 Side -60 inches below mid-height 38.9 2 77.8 80 Side -Mid -height 39.7 2 79.4 80 Side -60 inches above mid-height 15.6 2 31.2 30 Top -Center of lid 6.0 2 12 10 Top -Radially centered 8.4 2 16.8 20 In let duct 72.0 2 144 140 Outlet duct 18.6 2 37.2 40 8.0 Computer Software Microsoft WORD 2013 is used in this calculation for basic multiplication.

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-9 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION RIEV. 0 SEN E R CON CHECKLIST NO. PagelIof 8 CHECKLIST ITEMS 1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision?Client procedure is not used in this calculation.

ENERCON QA procedures used throughout this fl E] [project.2. Are the proper forms being used and are they the latest revision?

I ] I] [3. Have the appropriate client review forms/checklists been completed?

Client procedure is not used in this calculation.

ENERCON QA procedures used throughout this [ ][project.4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

Client procedure is not used in this calculation.

ENER.CON QA procedures used throughout this [] El [project.5. Is all information legible and reproducible?

El[][6. Is the calculation presented in a logical and orderly manner? [ ][7. Is there an existing calculation that should be revised or voided?This calculation does not replace any ENERCON produced calculation.

Information generated

[ ][by this calculation will be used by SNC to update their HNP EAL report.8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation?

No current ENERCON calculations exist that are similar to this calculation for addressing the El [] El SNC Hatch EAL update.9. If an existing calculation is being used for design inputs, are the key design inputs, I assumptions and engineering judgments used in that calculation valid and do they [ ][apply to the calculation revision being performed.

___ _______

SMNH-13-021 Attachment L SHEET L-10_________________

ENERCON Calculation for E-HU1 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0~jE NE R CON CHECKLIST

&',ydoy. PAGE NO. Page 2of 8 CHECKLIST ITEMIS 1 YES NO N/A 10. Is the format of the calculation consistent with applicable procedures and expectations?

I ] [][11. Were design input/output documents properly updated to reference this calculation?

No ENERCON design inputs or outputs are affected by this calculation.

This calculation will o affect the Hatch EAL evaluation.

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification?

[ ][OBJECTIVE AND SCOPE 13. Does the calculation provide a clear concise statement of the problem and objective of 0 U the calculation?

_____ _____ ____14. Does the calculation provide a clear statement of quality classification?

0] [15. Is the reason for performing and the end use of the calculation understood?

0] I 16. Does the calculation provide the basis for information found in the plant's license basis?The plant's license basis is not applied in this evaluation.

0] [] 0 16. Does the calculation provide the basis for information found in the plant's design basis docume ntatio n?The plant's license basis is not applied in this evaluation.

[] [][

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-1 1 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0~jEN ER C ON CHECKLIST e-Eery...

PAGE NO. Page 3of 8 CHCKIS IEM1YES NO NIA Calculation is applied in the development of the HNP EAL evaluation, not the plant license basis. 2l 21. If so, is this documented in the calculation?T Calculation is applied in the development of thle HINP EAL evaluation, not the plant license basis. LI III {22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal?

__ 1] _ _Calculation is applied in the development of the IINP EAL evaluation, not the plant license basis. ] ]DESIGN INPUTS 23. Are design inputs clearly identified?

El[][24. Are design inputs retrievable or have they been added as attachments?

I[25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

Attachments are not included in this calculation.

i] [] [26. Are design inputs clearly distinguished from assumptions?

I[ II F DESIGN INPUTS (Continued)

27. Does the calculation rely on Attachments for design inputs or assumptions?

If yes, are the attachments properly referenced in the calculation?

Attachments are not included in this calculation.

[] [] 1 28. Are input sources (including industry codes and standards) appropriately selected and ]are they consistent with the quality classification and objective of the calculation?

[ ][29. Are input sources (including industry codes and standards) consistent with the plant's ] ii design and license basis? [ ][30. If applicable, do design inputs adequately address actual plant conditions?

I 12 12 SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-12 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0 SEN ER C ON CHECKLIST Cxellenc-"erypojec. PAGE NO. Page 4of 8 CHECKLIST ITEMIS' YES NO N/A 31. Are input values reasonable and correctly applied? I E] El 32. Are design input sources approved?

I Eu [33. Does the calculation reference the latest revision of the design input source? ) L] I [] [34. Were all applicable plant operating modes considered?

[] Li LI ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective?

[] El El 36. Is adequate justification/basis for all assumptions provided?

[] El []37. Are any engineering judgments used? El [] Li 38. Are engineering judgments clearly identified as such?I No engineering judgments were applied in this evaluation.

j] LI El 39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards,I engineering principles, physical laws or other appropriate criteria?

El Li [No engineering judgments were applied in this evaluation.j________-

METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing Ell 41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated?

Plant licensing basis was not affected by this evaluation.

[ ][42. Is the methodology used consistent with the stated objective?

I El El [I. j SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-13 CALC NO. SNCO24-CALC-007 E E C N CALCULATION PREPARATION REV. 0 Ex eo,-Exeryprojxc,.

Exeryx. PAGE NO. Page 5of 8 CHECKLIST ITEMS 1 YES NO NIA 43. Is the methodology used appropriate when considering the quality classification of the ~ E calculation and intended use of the results? [ ][BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? ____ ______45. Is there reasonable justification provided for the use of equations not in common use?Equations applied in this evaluation are in common use in the industry.

I] ][46. Are the mathematical operations performed properly and documented in a logical fashion? [ ][47. Is the math performedocorrectly?

[] [E I []48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? [ ][49. Has proper consideration been given to results that may be overly sensitive to very small changes in input?Results generated by calculations performed in this evaluation are not significantly affected by [] [] [minor perturbations of variables.

SOFTWARE/COMPUTER CODES 50. Are computer codes or software languages used in the preparation of the calculation?

Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

LI [][51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met?.Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

l I SOFTWARE/COMPUTER CODES (Continued)

52. Are the codes properly identified along with source vendor, organization, and revision level? [ ][53. Is the computer code applicable for the analysis being performed?

El El 0]

SMNH-1 3-021 Attachment L FNFRC(')N CAI~Ikltl~inn fnr F-HIlli SHEET L-14 CALC NO. SNC024-CALC-007 CALCULATION PREPARATION REV. 0E NE R CON CHECKLIST verp,0ject PAGE NO. Page 6of 8 CHECKLIST ITEMVS' YES NO I N/A 54. If applicable, does the computer model adequately consider actual plant conditions?

r LI ][55. Are the inputs to the computer code clearly identified and consistent with the inputs and I assumptions documented in the calculation?

[ ][56. Is the computer output clearly identified?I Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

j[] LI [57. Does the computer output clearly identify the appropriate units?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

IW L I [58. Are the computer outputs reasonable when compared to the inputs and what was expected?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

I L RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?

This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria required for this evaluation.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use?This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria LI [] [required for this evaluation.

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-15 CALC NO. SNCO24-CALC-007 mCALCULATION PREPARATION REV. 0 F' E NE RC ON CHECKLIST NO. Page 7of 8 CHECKLIST ITEMS 1 YES NO N/A 62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

This calculation provides results for the SNC HNP EAL evaluation.

No acceptance criteria LI LI [required for this evaluation.

63. Do the calculation results and conclusions meet the stated acceptance criteria?This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria LI I required for this evaluation.

64. Are the results represented in the proper units with an appropriate tolerance, if I L applicable?

[ 65. Are the calculation results and conclusions reasonable when considered against the I I I stated inputs and objectives?

[ ][66. Is sufficient conservatism applied to the outputs and conclusions?

I [ I L 67. Do the calculation results and conclusions affect any other calculations?

No ENERCON calculations are affected by this evaluation.

Results are provided to SNC HNP LI I for input into the Hatch EAL evaluation.

68. If so, have the affected calculations been revised?No ENERCON calculations are affected by this evaluation.

Results are provided to SNC HINP LI I for input into the Hatch EAL evaluation.

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation?

Calculation is based on design input and assumption data provided and used by client in their 10 LI [] LI CER 72.212 Report. Parameters maintained for consistency.

70. If so, are they properly identified?

No open assumptions applied in this evaluation.

Assumptions have basis based on information LI I provided by the client.

SMN H-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-16 CALC NO. SNC024-CALC-007 CALCULATION PREPARATION REV. 0E NER C ON CHECKLISTPAGE NO. Page 8of 8 DESIGN REVIEW 71. Have alternate calculation methods been used to verify calculation results? [ ][Note: 1. Where required, provide clarification/justification for answers to the questions in the space provided below each question.

An explanation is required for any questions answered as "No' or "N/A".Originator:

/Date Print Name and Sign SMNH-13-021 SMN -I 3021 HEETM-1Attachment M ENERCON Calculation for RA1 SHEET M-1 CALC NO. SNCO24-CALC-OO8 I~EN E R CON CALCULATION COVER Excllence-£,ep~oectC ,yd SHELET REV. 0 PAGE NO. I of 10 Til:Hatch EALs RA1 Threshold to Address NEI CletSC 99-01 Revision 6 PoetIetfe:SC2 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary

[] [information, that require confirmation? (If YES, identify the assumptions.)_________

2 Does this calculation serve as an =Alternate Calculation"? (If YES, identify the ][design verified calculation.)

Design Verified Calculation No. __________

___3 Does this calculation supersede an existing Calculation? (If YES, identify the ][design verified calculation.)

Superseded Calculation No. __________

___ ___Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation El Final Calculation

[]Safety-Related El Non-Safety-Related

[](Print Name and Sign)Originator:

David Hartmangruber~;

• Date: 3 ./Design Reviewer Dominic Napolitano, Ph. D / /, Approver:

Jay Maisler, CHPDigitally signed b M ~tMais lr, CHP'P,,I ,-~l I tI,nr C"I-IP c=FNF:RCON_

niic=USDate':2015.10.23 17:29:,54

-04'00' SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-2 CALC NO. SNC024-CALC-008

~EN ER C ON CALCULATIONREVISION STATUS SHEET RV PAGE NO. 2 of 10 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 10/23/2015 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-10 0 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO. OF REVISION ATTACHMENT NO. OF REVISION PAGES NO. PAGES NO.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-3 CALO NO. SNC024-CALC-008

~E N E R CON TABLE OF CONTENTS REV. 0 ExcelleflCe Evypoet.Eryd.

PAGE NO. 3 of 10 Section 1.0 Purpose and Scope.......................................................

2.0 Summary of Results and Conclusions

.....................................

3.0 References........;........................................................

4.0 Assumptions..............................................................

5.0 Design Inputs..............................................................

6.0 Methodology..............................................................

7.0 Calculations

..............................................................

8.0 Computer Software........................................................

9.0 Results and Conclusion

...................................................

Page No.4 5 6 6 7 8 8 9 10 SMN H-I13-02 1 Attachment M SHEET M-4 C..Ir-nI~tinn fnr IPA1 CALC NO. SNC024-CALC-008 Hatch EAL RAI Threshold to FIE N E R C ON Address NEI 99-01 Revision REV. 0 6 PAGE NO. 4 of 10 1.0 Purpose and Scope The purpose of this calculation is to calculate the Emergency Action Level (EAL)threshoids for the update of the RAI calculation in the Southern Nuclear (SNO) Design Calculation SMNH-05-009 (Reference

1) in response to the changes made to the Initiating Condition (IC) AA1 in Revision 6 of NEI 99-01 (Reference 2). Calculation RA1 is meant to address the IC AA1 (Section 4.1 of NEI 99-01 Revision 6 states "R may be used in lieu of A" for this recognition category provided the change is carried through for all the associated IC identifiers).

Revision 6 of NEI 99-01 IC AA1 identifies an EAL threshold for a release of gaseous or liquid radioactivity resulting in an offsite dose to a member of the public greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (CDE). IC AA1 is applicable to all operating modes and there are 4 EALs outlined in NEI 99-01 for IC AAI.1. Reading of site specific radiation monitors greater than threshold values that would generate a dose rate greater than the dose criterion established in IC AA1 in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Reading must be shown for 15 minutes or longer.2. Dose assessment using actual meteorology indicates doses greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (CDE) at or beyond site boundary 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond site boundary for one hour of exposure.4. Field survey results indicate either of the following at or beyond site boundary.

A closed window dose rate greater than 10 mR/hr expected to continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer an analyses of field survey samples indicates a thyroid CDE greater than 50 mrem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-5 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to~0 E N E R C ON Address NEI 99-01 Revision REV. 0 6 PAGE NO. 5 ofl10 The scope of this calculation is to determine site specific instrument readings for the RA1 EAL 1 threshold.

The IC RA1 EAL 2, 3, and 4 are not evaluated in this calculation.

The quality rating of this calculation is non-safety related due to results only being used to generate a revised set of EALs for submission by the Hatch nuclear power plant (HNP).2.0 Summary of Results and Conclusions The instrument readings that indicate an EAL threshold value has been reached for IC RA1 are calculated in this calculation.

IC RAI is the release of gaseous or liquid radioactivity resulting in offsite dose to a member of the public greater than 10 mrem TEDE or 50 mrem thyroid ODE.The RA1 EAL 1 is the valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown in Table 2-1.Table 2-1 Radiation Monitor RA1 EAL Threshold Values 1/2011 P005 Reactor Building Vent 2.6E-02 pCi/cm 3 Detector:

112011 N048 and 1/2011 N049 1 D11P006 Main Stack 8.1 E+01 pCi/cm 3 Detector:

10D11 N0055 and 10D11 N056 10D11 R763 A and B Recombiner Building Vent Off-Scale 1'The maximum range of the recombiner building went monitor is 1 .0E+06 cps. The calculated EAL thresholds for RAl, RS1, and RG1 are 1.27E+07, 1.27E+08, or 1 .27E+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMNH-13-021 Attachment M SHEET M-6___________________

ENERCON Calculation for RA1 ________CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to ENERCON Address NEI 99-01 Revision REV. 0 PAGE NO. 6 ofl10 3.0 References

1. SMNH-05-009 Rev 2, NEI 99-01 EAL Calculations, Southern Company, November 19 2014.2. NEI 99-01 Rev 6, Development of Emergency Action Levels for Non-Passive Reactors, November 2012, Nuclear Energy Institute.

4.0 Assumptions

Based on the analysis of the methodology of Reference 1, the following assumption is consistent with the previously performed calculations, but were not included in the listed assumptions of Reference 1.4.1 Perfect Monitor Response This assumption is applied to be consistent with the calculations performed in Attachment I of Reference

1. It is assumed in this calculation that the monitors at the end of each pathway are not energy dependent or that the monitor response accounts for the relative energy spectrum associated with the thresholds determined in this calculation based on the expected proportion of each isotope in the overall concentration.

This is a simplifying assumption applied due to the limited information provided about the monitoring equipment.

SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-7 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to jE N E R CO0N Address NEI 99-01 Revision REV. 0 6 PAGE NO. 7 oflO0 5.0 Design Inputs 5.1 Hatch Indication and RS1 EAL I Thresholds SMNH-05-009 (Reference

1) addresses the IC RSI, which is based on the release of gaseous radioactivity resulting in offsite dose to a member of the public greater than 100 mrem TEDE or 500 mrem thyroid ODE. SMNH-05-009 evaluates three release pathways and determines the monitor readings that would indicate an EAL threshold value has been reached for IC RSI. The monitor readings that would indicate an EAL threshold value for IC RSI are provided in Table 5-1.The indicating ranges for the radiation monitors are also provided in Table 5-1 and the ranges are based on Pages 51 and 52 of Reference

1. These values are used to calculate the IC RA1 EAL threshold values.Table 5-1 Radiation Monitor RS1 EAL Threshold Values

..

...Vent Path , % iMonitor

Indicating Range" 1/2011 P005 Reactor Building 2.6E-01 pCi/cm 3 1 E-03 to 1E+06 Vent i~/m Detector:

1/2D11 N048pi/m and 1/2D11 N049 1011 P006 Main Stack 8.1E+02 JCi/cm 3 1E-03 to 1E+05 Detector:

10D11 N0055 pCi/cm 3 and 1011N056 1011R763 A and B2 Recombiner Building 1.27E+08 cps 1E-01 to 1E+06 Vent cps 2 Based on Page 20 of Reference 1, the recombiner building detector reads in cps, not in jtCilcm 3.

SMN H-1 3-021 Attachment M SHEET M-8 ENERCON CIriik~tinn fnr PAl CALC NO. SNCO24-CALC-008 Hatch EAL RAI Threshold to ____ ________SE N E R C ON Address NEI 99-01 Revision REV. 0 Excdellece-Ev'ery project. Every day.6 PAGE NO. 8 of 1O 6.0 Methodology In the SMNH-05-009 (Reference

1) RSI evaluation, EAL 1 thresholds were set based on readings of radiation monitoring equipment for several effluent pathways.

The thresholds are shown in Table 5-1 of this calculation.

The calculations for dose rate estimates is linear, therefore the RS1 readings are scaled down by a factor of 10 (multiple of 0.1) for the RAl evaluation performed in this calculation resulting in EAL 1 threshold values reflecting an offsite dose to a member of the public greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (ODE). The calculation of the RA1 EAL I threshold values is provided in Section 7.0.7.0 Calculations As discussed in Section 6.0, the values provided in Table 5-1 are multiplied by a scaling factor of 0.1 for the RAl EAL I thresholds.

The resultant threshold values for RAl EAL 1 are shown in Table 7-1. The reactor building vent and main stack threshold values are within the radiation monitor 1/2D11P005 and ID11P006 indicating range. The calculated RAl PSI, and RGI EAL I threshold values for the recombiner building vent are beyond the indicating range of the radiation monitor 1D11R763.

The maximum range of the recombiner building vent monitor is 1E+06 cps. The calculated EAL thresholds for RA1, RSI, and RG1 are 1.27E+07, 1.27E+08, or 1.27E+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-9 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to*0 E N E R CO0N Address NEI 99-01 Revision REV. 0 Excellen £-vry proect.Cveyday.

6 PAGE NO. 9 ofl10 Table 7-1 Radiation Monitor RA1 Threshold Values Radiation. , Vent Path: , :Scaled:>

,. Scalin ,:: Resultant:

Indicating ... .i ..Facto Monitor Range ..1/2D11IP005 Reactor Building 2.6E-01 0.1 2.6E-02 pCi/cm 3 1 E-03 to Deetr /D 08 Vent pCVcm 3 1E+06 pCi/cm 3 and 1/2D11N049 1D11P006 Main Stack 8.1E+02 0.1 8.1E+01 pCi/cm 3 1E-03 to Detector:

1 D11 NOO55 pCi/cm 3 1E+05 pCi/cm 3 and 1D11N056 1D11R763 Aand B 3 Recombiner 1.27E+O8 cps 0.1 Off-Scale 4 1 E-01 to Building Vent 1E+06 cps 8.0 Computer Software No computer software was used in the creation of this calculation.

Based on Page 20 of Reference 1, the recombiner building detector reads in cps, not in jiCilcm.4 The maximum range of the recombiner building vent monitor is 1 .0E+06 cps. The calculated EAL thresholds for RAl, RS1, and RG1 are 1.27E+07, 1.27E+08, or 1.27B+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMN H-I13-021 Attachment M SHEET M-10 ENFRC.nM fnr PAl CALC NO. SNCO24-CALC-008 Hatch EAL RA1 Threshold to Ej E N E R CO0N Address NEI 99-01 Revision REV. 0 6 EaceI/aence--Ev~ery proect. Every day PAGE NO. 10 ofl1O 9.0 Results and Conclusion The purpose of this calculation is to calculate the EAL thresholds for the RA1 calculation in the SNC Design Calculation SMNH-05-009 (Reference

1) for HNP use in development of an EAL submittal based on NEI 99-01 Revision 6. Table 2-1 contains the threshold values associated with the RAI EAL 1. These values will be applied to the HNP Emergency Action Level Scheme developed using the guidance of NEI 99-01 Revision 6.

SMN H-i13-02 1 Attachment M ENERCON Calculation for RA1 SHEET M-11I CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 SEN E R CON CHECKLIST e-.ver.....

c y PAGE NO. Page 1 of 8 CHECKLIST ITEMS 1 YES NO NIA GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision?Client procedure is not used in this calculation.

ENERCON QA procedures used throughout

(] I] 0 [this project.2. Are the proper forms being used and are they the latest revision?

] i [ E 3. Have the appropriate client review forms/checklists been completed?

I Client procedure is not used in this calculation.

ENERCON QA procedures used throughoutI u] El 0 this project.4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

Client procedure is not used in this calculation.

ENERCON QA procedures used throughout El l thifs project.5. is all information legible and reproducible?

0 E El 6. Is the calculation presented in a logical and orderly manner?.0 El E 7. Is there an existing calculation that should be revised or voided? T This calculation does not replace any ENERCON produced calculation.

Information generated 5 El by this calculation will be used by SNC to update their Hatch Nuclear Power Plant EAL report.8. is it possible to alter an existing calculation instead of preparing a new calculation for ] ]this situation?

No current ENERCON calculations exist that are similar to this calculation for addressing the El E SNC Hatch EAL update.9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they __] Elf [apply to the calculation revision being performed.

___ _______

SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-12 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 SE NE R CON CHECKLIST Foco .....-Erey projec,. Prry d PAGE NO. Page 2 of 8 CHECKLIST ITEMS 1 YS N I 10. Is the format of the calculation consistent with applicable procedures and expectations?

I ] I{iI [11. Were design inputfoutput documents properly updated to reference this calculation?

No ENERCON design inputs or outputs are affected by tliis calculation.

This calculation will affect the Hatch EAL evaluation.

12. Can the calculation logic, methodology and presentation be properly understood i J l iz without referring back to the originator for clarification?

[ ][OBJECTIVE AND SCOPE 13. Does the calculation provide a clear concise statement of the problem and objective of 0 LIi the calculation?

[ ][14. Does the calculation provide a clear statement of quality classification?

0] i Li 15. Is the reason for performing and the end use of the calculation understood?

0] L LI 16. Does the calculation provide the basis for information found in the plant's license basis?The plant's license basis is not applied in this evaluation.

LI [][17. If so, is this documented in the calculation?

The plant's license basis is not applied in this evaluation.

LI LI 0]18. Does the calculation provide the basis for information found in the plant's design basis documentation?

The plant's license basis is not applied in this evaluation.

LI [][19. If so, is this documented in the calculation?

The plant's license basis is not applied in this evaluation.

LI LI 0]

SMN H-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-13 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 jjE NE R CON CHECKLISTprojec. £re,' day PAG E NO. Page 3 of 8 CHECKLIST ITEMS 1 YES NO N/A 20. Does the calculation otherwise support information found in the plant's design basis documentation?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, I] [] []not the plant license basis.21. If so, is this documented in the calculation?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, [ ][not the plant license basis.22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, El El 0]not the plant license basis.DESIGN INPUTS 23. Are design inputs clearly identified?

__ ElI []24. Are design inputs retrievable or have they been added as attachments?

Jo] Elf El[25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

Attachments are not included in this calculation.

Ill El 0 26. Are design inputs clearly distinguished from assumptions?

loj] El 27. Does the calculation rely on Attachments for design'inputs or assumptions?

If yes, are the attachments properly referenced in the calculation?

Attachments are not included in this calculation.

El [][28. Are input sources (including industry codes and standards) appropriately selected and 0 El E are they consistent with the quality classification and objective of the calculation?

[ ][29. Are input sources (including industry codes and standards) consistent with the plant's 0 l E design and license basis? ] [ I ]

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-14 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 1j EN E RC:ON CHECKLIST EweIenc-EeryproJoe,. doy PAG E NO. Pag e 4 of 8 CHECKLIST ITEMS 1 YES NO I NIA 30. If applicable, do design inputs adequately address actual plant conditions?

0] El I [31. Are input values reasonable and correctly applied? 0] 12 El[32. Are design input sources approved?

0] LI 1 33. Does the calculation reference the latest revision of the design input source? 0] 12 [ [ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective?

El El [36. Is adequate justificationfbasis for all assumptions provided?

0] I [] [37. Are any engineering judgments used? LI 0 L 38. Are engineering judgments clearly identified as such?No engineering judgments were applied in this evaluation.

LI [][39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, engineering principles, physical laws or other appropriate criteria?

12 [] [No engineering judgments were applied in this evaluation.__

E0 METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing 1 41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated?

Plant licensing basis was not affected by this evaluation.

[ ][

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-15 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0ENER RCON CHECKLISTE NO. Page 5 of 8 CHECKLIST ITEMS 1 ES NO NIA 42. Is the methodology used consistent with the stated objective?

El[][43. Is the methodology used appropriate when considering the quality classification of the T T [ E calculation and intended use of the results? I I ] [BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice rE and the plant's design and license basis? [ ][45. Is there reasonable justification provided for the use of equations not in common use?Equations applied in this evaluation are in common use in the industry.

0][][46. fahinAre the mathematical operations performed properly and documented in a logical 0] El []47. Is the math performed correctly?

0] t 48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? [] El Il 4,9. Has proper consideration been given to results that may be overly sensitive to very small changes in input?Results generated by calculations performed in this evaluation are not significantly affected by Q] El 0 minor perturbations of variables.

SOFTWARE/COMPUTER CODES 50. Are computer codes or software languages used in the preparation of the calculation 9 No software languages or codes were used in the development of this calculation.

El[][51. Have the requirements of CSP 3.09 for use of computer codes or software languages,T including verification of accuracy and applicability been met?El El 0]No software languages or codes were used in the development of this calculation.

SMN H-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-16 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 0, E N ER CON CHECKLISTPAGE NO. Page 6 of 8 CHECKLIST ITEMS 1 YES NO N/A SOFTWAREJCOMPUTER CODES (Continued)

52. Are the codes properly identified along with source vendor, organization, and revision level?No software languages or codes were used in the development of this calculation.

53. Is the computer code applicable for the analysis being performed?

No software languages or codes were used in the development of this calculation.

El []I [54. If applicable, does the computer model adequately consider actual plant conditions?

No software languages or codes were used in the development of this calculation.

El [][55. Are the inputs to the computer code clearly identified and consistent with the inputs andI assumptions documented in the calculation?I No software languages or codes were used in the development of this calculation.

[ ][56. Is the computer output clearly identified?

No software languages or codes were used in the development of this calculation.

[ ][57. Does the computer output clearly identify the appropriate units?No software languages or codes were used in the development of this calculation.

[1 LI 0]58. Are the computer outputs reasonable when compared to the inputs and what was expected?No software languages or codes were used in the development of this calculation.

l E0 59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results?No software languages or codes were used in the development of this calculation.

l E0 SMN H-I13-021 Attachment M ENERCON Calculation for RAl SHEET M-17 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 0EN E R CON CHECKLIST...

NO. Page 7 of 8 CHECKLIST ITEMS 1 YES NO NIA RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?

This calculation provides results for the SNO Hatch nuclear power plant EAL evaluation.

No acceptance criteria required for this evaluation.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use?This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No El El 0]acceptance criteria required for thfis evaluation.

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No 5] El 0 acceptance criteria required for thils evaluation.

63. Do the calculation results and conclusions meet the stated acceptance criteria?This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No ~ E acceptance criteria required for this evaluation.

___I 64. Are the results represented in the proper units with an appropriate tolerance, if 0 I applicable?

[ ][65. Are the calculation results and conclusions reasonable when considered against the 0 n E stated inputs and objectives?

[ ][66. Is sufficient conservatism applied to the outputs and conclusions?

0] El El[67. Do the calculation results and conclusions affect any other calculations?

No ENERCON calculations are affected by this evaluation.

Results are provided to SNC Hatch__ njo[nuclear power plant for input into the Hatch EAL evaluation.

E 68. If so, have the affected calculations been revised?No ENERCON calculations are affected by this evaluation.

Results are provided to SNC Hatch nuclear power plant for input into the Hatch EAL evaluation.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-18 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0~ NERCON CHECKLIST....

NO. Page 8 of 8 CHECKLIST ITEMS 1 YES NO N/A 69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation?

Calculation is based on design input and assumption data provided and used by client in their [] 0] El current EAL evaluation.

Parameters maintained for consistency.

70. If so, are they properly identified?

No open assumptions applied in this evaluation.

Assumptions have basis based on information I provided by the client.Note: 1. Where required, provide clarification/justification for answers to the questions in the space provided below each question.

An explanation is required for any questions answered as "No' or "N/A".Originator:

David Hartmangruber I Date Print Name and Sign Vogtle Electric Generating Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 3 Vogtle EAL Calculations

~ER~A 1~V'* ~~#rIZbd?Southern Nuclear Design Calculation SCalculation Number: X6CNAI5 Plant: Vootie Electric Generatino Plant Unit: fl 1 0] 2 [ 1 & 2 i Discio)line:

Mechanical Title: NEI 99-01 Rev 6 EAL Calculations

Subject:

Emergency IAction Level Setjpoints Purpose I Objective:

Document Emer eq.nc Action Level Values to support conversion to NEI 99-01 Rev 6 System or Equipment Tag Numbers: NIA Contents Topic Page "Attachments

1. of (Computer Printouts, Technical Papers, Sketches, Pages_____________Correspondence)

Purpose I A -SNC EP Concurrence 1 criteria 1 8- Reserved 0)Conclusion 3 C -References from X6CNA14 31 Design Inputs ...6 D0- TEDE and Thyroid CDE Calculations 18 Assumptions 18 E E- Shielding Calculations 29---References 22F -Evaluation of 52 psig Pressure on Mechanical 1 P.en.rtrations Method of Solutions 26 G -References from X6CNA16 Body of Calculation 31 H -VEGP 1, EAL Thresholds 7!- VEGP 2 EAL Thresholds

• 7i___________

___J -Reviewer Alternate Calculation 11"-'K -SNC024-CALC-005 Vogtle EALs RA1 thresholds 1 7______ to address NEI 99-01 rev 6 ______......_____L

-SNCQ24-CALC-004 VEGP EAL for IC E-HUI 16-'--Total # of Pages including cover sheet & 22_8 Attachments:

Nuclear Quality Level 0DSafety-Related E] Safety Significant 0l Non- Safety -Significant' erinRecord

__________________________

vmo origlnator' Revl~wer Approval I Approval2 No. IDescriptioun

_______ _______ _______" sudo~w. Wu C.T, Martin A.T Vieira DL.L Lamnbert____ ___________

Sgonfie Sign. on file Sign. on filegnonfl 2 :Complete Revision ' :.. \L. -.",, ___ r in Hawkinson Jay Maisler Keith Drudy Notes: 1. Additional work~ and changes to this calculation are required.

The calculation is "APPROVED PENDING" NRC review and acceptance of the Emergency Plan (EP) submittal.

2. The RTYPE for this document is CV805,015.

NMP-ES-039-F01 V4,0 NMP-ES-039-001 PURPOSE The purpose of this calculation is to provide values/data/curves and bases for use in development of the Vogtle Electric Generating Plant Emergency Action Levels (EAL) using NEI 99-01 Revision 6 guidelines.

This combined calculation document includes all unique calculations required to support emergency action level threshold as well as references to calculations used to create thresholds, but serve purposes beyond emergency action levels.The contents of this calculation are primarily an amalgamation of the calculations which supported the previous emergency action level scheme. The work performed in calculations X6CNA14 and X6CNA16 is directly transposed into this document and edited to account for the differences between NEI 99-01 Rev 4 and 6.Several emergency action level thresholds have been eliminated due to the new scheme. The thresholds previously calculated supporting ICs RU2, RA2, RA3, and SU4 have not been carried into this document.

Attachments K and L of this document contain calculations supporting new emergency action level thresholds and represent the only portion of this document which is not directly transposed from X6CNA14 and X6CNA16. The transposed material in this calculation has been further altered to reflect the new language and organization of NEI 99-01 Rev 6. These changes in language and organization are administrative in nature and have no impact on the calculation output.Criteria: The calculation performed will support the development of guidelines for NEI 99-01 EALs RU1, RA1, RSI, RO1, CAl, CS1, OGI, E-HU1, Fuel Clad Barrier Loss 3.A, RCS Barrier Loss 3.A, Containment Barrier Potential Loss 3.A and Containment Barrier Potential Loss 4.8.1. Declaration of an emergency, when such a declaration is not required, involves risk to the public as does the failure to make such a declaration, should one be warranted.

Therefore, this calculation shall develop a "best estimate" value for the dose rates or curie concentrations sensed at the monitors chosen for the Emergency Action Level (EAL) set points. When judgments are necessary, these judgments shall be as close to anticipated conditions as possible.2. If a particular monitor is to be used for an EAL, then the dose rate or curie concentration set point developed for that specific monitor shall be within the range of the monitor, or the monitor shall not be cited as applicable for the EAL.3. In accordance with the guidance of Regulatory Guide 1 .97 Revision 2, post-accident radiation monitors must read within a factor of 2 of actual radiation conditions.

Therefore, changes in the set points of this revision that are within a factor of 2 of the previous revision's set point for the same EAL do not invalidate the previous set point. It is up to the ultimate user of these calculations to determine if change to the EAL set point guidance document(s) is warranted.

4. Methods and Assumptions shall comply with the guidance of NEI 99-01 Revision 6.

Southern Nuclear Operating Company SOUTHERNE,.

Pln:VG Title: NEI 99-01 Rev 6 EAL Calculations XIN1 I COMPANY Unit: 1&2 SHEET 2 Note: NEI 99-01 Rev 6 states that the "A" Recognition Category designation may be changed to "R" provided the change is carried through for all of the associated IC identifiers.

As such, the Vogtle Electric Generating Plant Emergency Action Levels use the Recognition Category designation of "R" for the Abnormal Rad Levels / Radiological Effluent Recognition Category.

Southern Nuclear Operating Company SIm Plant: VEGP I X6CNA1 5 COMplant Unit: 1&2 Title: NEI 99-01 Rev 6 EAL CalculationsSHE3

==

Conclusions:==

The results of the calculations of the values / data required for the develoPment of the Vogtle Electric Generating Plant EALs are provided in the Body of the Calculations.

Initiating Condition RUl The liquid and gaseous release paths are continuous (per Tables 2-4 and 3-4 of VEGP ODCM), with release permits and alarm setpoints generated weekly.Thus, default setpoints are deleted from this version of the calculation.

Initiating Condition RA1 Greater than any of the following monitor readings, for 15 minutes or longer, serves as the threshold for EALI.Radiation Vent Path Reading Monitor RE-12444E Plant Vent -High Range 0.50 pCi/cc RE-I12839E Turbine Building Vent (SJAE) High 21 pCi/cc____ ___ ___ Range _ _ _ _ _ _Initiating Condition RS1 Greater than any of the following monitor readings, for 15 minutes or threshold for EALl.longer, serves as the Radiation Vent Path Reading Monitor RE-12444E Plant Vent -High Range 5 pCi/cc RE-12839E Turbine Building Vent (SJAE) -High Range 210 pCi/cc Southern Nuclear Operating Company SOT AN4 Plant: VEGP TteNE990Re6EACacliosX6CNA15 SOUTERA Unit: 1&2 Tite NE 90 e A acltosSHEET 4 Initiating Condition RGI Greater than any of the following monitor readings, for 15 minutes or longer, serves as the threshold for EALI.Radiation Vent Path Reading Monitor RE-12444E Plant Vent 50 jiCi/cc RE-12839E Turbine Building Vent (SJAE) -High 2100 Range Initiating Condition CA1 The following indication serves as a threshold for EAL 1 RPV water level less than 185'-10" Initiating Condition CSI The following indication serves as a threshold for EAL 1 .b.RPV water level less than 1 85'-4" The following indication serves as a threshold for EAL 2.b RPV level less than 181'-1 0" Greater than or equal to the following monitor reading serves as a threshold for EAL 3.b RE-005 or Containment Operating Deck High Range > 40 REM/hr RE-006 Initiating Condition CGI The following indication serves as a threshold for EAL 1l.b RPV level less than 181 '-1 0" Greater than or equal to the following monitor reading serves as a threshold for EAL 2.b RE-005 or Containment Operating Deck High Range > 40 REM/hr RE-006 Southern Nuclear Operating Company souTHl Pat:VG il:NE 90 e EAIacuain X6NA5 COMPANY Unit: 1&2 Til:HEEE-1Te ELCacla5n Initiating Condition E-HUI Greater than any of the following on-contact radiation readings (gamma + neutron) serve as the thresholds for EAL 1.LocationEA

______________________(m remlhr)HI-TRAC 125 Side -Mid -height 950 Top J 200 HI-STORM 100 Side -60 inches below mid- 170 Side -Mid -height 180 Side -60 inches above mid- 110 Top -Center of lid 50 Top -Radially centered 60 Inlet duct 360 Outlet duct 130 Fuel Clad Barrier Loss Threshold 3.A Greater than the following monitor readings serve as Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 > 2.6E+5 mR/hr RCS Barrier Loss Threshold 3.A Greater than the following monitor readings serve as Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 >= 8.7E+2 mR/hr Containment Barrier Potential Loss Threshold 3.A Greater than the following monitor readings serve as Potential Loss Threshold 3.A Containment radiation monitor RE-005 OR -006 >= 1 .3E+7 mR/hr.

Southern Nuclear Operating Company SOUHEA, Plant: VEGP Titl:NI9-1Rv6ELCluain X6CNA15 ISUHUnit:

1&2 TteNE990Re6EACacliosSHEET 6 DESIGN INPUTS Radiation Monitor System Parameters

1. Vogtle Radiation Monitoring System Operating Ranges Liquid Effluent Monitor Release Path Operating Range Page RE-0018 Liquid Radwaste Effluent Line 1.0E-06 to 1,0E-01 jiCi/cc 10 RE-0021 SIG Blowdown Effluent Line 4.0E-07 to 4.0E-02 pCi/cc 11 RE-0848 Turbine Building Drain Effluent Line 4.0E-07 to 4,0E-02 pCi/cc 12 Gaseous Effluent Monitor Release Path ARE-0014 Waste Gas Process Effluent Line 1.0E-01 to 1.0E+04 6 RE-I124420 Plant Vent 5.0E-07 to 5.0E-02 pCi/cc 3 RE-I124440 Plant Vent -Normal Range I1.0E-07 to 3.4E-03 pCi/cc 3 & 4 RE-I12444D Plant Vent -Mid Range 3.4E-03 to 4.0E-01 jiCi/cc 3 & 4 RE-I12444E Plant Vent -High Range 4.0E-0I to 5.8E+04 p, Ci/cc 3 & 4 RE-I128390 Steam Jet Air Ejector -Normal 1 .0E-07 to 3.4E-03 liCi/cc 1 RE-I12839D Steam Jet Air Ejector -Mid Range 3.4E-03 to 4.0E+01 #iCi/cc I RE-I12839E Steam Jet Air Ejector -High Range 4.0E+01 to 5.8E+04 pCi/cc I Area Radiation Monitors Monitor Release Path RE-0002 Containment Operating Deck I1.0E-01 to I1.2E+03 mREM/hr 4 & 5 RE-0003 Low Range RE-0004 Containment Access Hatch Area I1.0E-01 to I1.2E+03 mREM/hr 13 RE-0005 Containment Operating Deck 1 .0E-01 to 1 .2E+08 REM/hr 5 RE-0006 High Range RE-0008 Fuel Handling Building 1.0E-01 to I.2E+03 mREM/hr 13 RE-0011 Seal Table Room 1.0E-01 to I.2E+03 mREM/hr 13

Reference:

Attachment C, X6AZ01A.

2. Liquid Effluent Monitors' Alarm Setpoints Monitor Release Path Release High Alarm Reference Type* Setpoint (iC i/cc)RE-0018 Liquid Radwaste Effluent Line Batch Procedure 3431 1-C During Release Permit Scin Dependent 5.1 & 5.3 Between Releases 9.99E+20 RE-0021 SIG Blowdown Effluent Line Continuous Procedure Permit 34306-C During Release Section Dependent 5.1 & 5.3 No Activity 1 .00E-05 RE-0848 Turbine Building Drain Effluent Continuous Procedure Line 34310-C During Release Permit Scin____ ___ ____ ___ ____ ___ ____ ___ __ D pen ent 5.1 & 5.3 No Activity 1 .00E-05* Table 2-4, page 2-18, VEGP ODCM Refer to the following VEGP ODCM figures for the listed radiation monitors: Monitor VEGP 00CM Figure VEGP 00CM Page RE-0018 Figures 2-1 & 2-2 Pages 2-14 & 2-15 RE-0021 Figure 2-3 Page 2-16 RE-0848 Figure 2-3 Page 2-16 3. Gaseous Effluent Monitors' Alarm Setpoints Monitor Release Path Release High Alarm

Reference:

Type* Setpoint Procedure (pl~i/cc) 34333-C ARE-0014 Waste Gas Process Effluent Batch Sections Line 10.1 & 10.2 During Release Permit Dependent Between Releases 9.99E+20 RE-I12442C Plant Vent Continuous Permit Sections Dependent 7.1 & 7.3 RE-12444C Plant Vent Continuous Permit Sections Dependent 8.1 & 8.3 RE-I128390 Turbine Building Vent Continuous Sections (Steam Jet Air Ejector -9.1 & 9.3 Normal)______

No Confirmed Primary-to-Secondary 7.84E-04 Leakage Confirmed Primary-to-Secondary Leakage Permit_______________________________________

Dependent______

• Table 3-4, page 3-21, VEGP 00CM Refer to the following VEGP ODCM figures for the listed radiation monitors: Monitor VEGP ODCM Figure VEGP ODCM Page ARE-0014 Figure 3-1 Page 3-15 1IRE-12442C Figure 3-2 Page 3-16 2RE-1 24420 Figure 3-3 Page 3-17 RE-12839C Figure 3-4 Page 3-18 Plant Vent release type confirmed in 27AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager (copy in Attachment 03).The SJAE discharges to the environment via the Turbine Building Vent. As the SJAE must operate to maintain condenser vacuum while at power, it is a continuous release path.Turbine Building Vent release type confirmed in 28AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager (copy in Attachment 04).

Southern Nuclear Operating Company SO~a k Plant: VEGP X6CNA15*o~l Uit 12 Title: NEI 99-01 Rev 6 EAL Calculations SHEET9 Spent Fuel Pool Parameters

4. SFP Elevations Elevation Value Reference SFP Floor 1 79'-01/4" AX4DR023 Fuel Transfer Tube Centerline 186'-9%" AX4DR023, 1X2D48E007, 2X2 D48 E007 Top of spent fuel racks 1 93'-5" AX4DR023 Elevation of SFP Water Normal Level 21 8'-6"~ AX4DR023 Elevation of SFP Water Low Level 21 7'-0" AX4DR023 Containment Dimensions

5. Containment Elevations

& Dimensions Elevation/Dimension Value Reference Operating Deck Elevation 220'-0" 1x2D48E007, 1X2D48E008, 2X2D48E007, & 2X2D48E008 Containment Inside Diameter 140 ft 1X2D01A001

& 2X2D01A001 Operating deck thickness 2'-9" 1X2D48E007, 1X2D48E009, (above Seal Table Room) & 2X2D48E009 Top of inside Containment*

397'-.9" 1X2D01A001

& 2X2D01A001 Fuel Transfer Tube Centerline 186'-9%" AX4DR023, 1X2D48E007, 1 XD48E008, 2XD48E007 Elevation of RPV flange 1 94'-0" AX4DR023* Elevation of liner inner surface at top = spring line elevation 327'-9" + 70'-0" radius 6. Containment volume = 2.95E+06 cu ft

Reference:

Table 1 5A-1, VEGP FSAR Revision 20 (December 2015)7. Containment volume fraction above operating deck = 77.1%Unsprayed net free volume above operating deck = 68,900 cu ft Total net free volume below operating deck = 603,200 cu ft Volume fraction above operating deck = 1 -[(68,900 ft 3+ 603,200 ft 3)/(2,932,000 ft 3)]

Southern Nuclear Operating Company SOUTHERNA Ulnit: VEGPi Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 Volume fraction above operating deck = 0.771

Reference:

Table 6.5.2-1, VEGP FSAR Revision 19 (February 2014)8. Containment liner: 1/4" carbon steel

Reference:

VEGP FSAR sections 1.2.5, 6.2.7.2, & 6.5.3.1 and drawings 1X2D01A001

&2X20D01A001 Reactor Coolant System Parameters

9. Reactor Pressure Vessel & RCS Piping Dimensions Parameter Value Reference RPV Inside Diameter 173" VEGP FSAR Table 5.3.3-1 Hot Leg centerline elevation 187'-0" AX4DR023, IX4DL4A017-1, &(76% RVLIS) 2X4DL4A01 7-1 Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2A" AX4DR023 Top of Active Fuel 181'-10" AX4DR023 (63% RVLIS)Cold Leg Pipe ID 27%" 1X4DL4A017-1

& 2X4DL4A017-1 Hot Leg Pipe ID 29" 1X4DL4A017-1

& 2X4DL4A017-1 RCS Coolant Parameters Parameter Value Reference Full Power Tavg 588.4 °F Table 2-1, page 2-3, WCAP-1 6736-P VEGP FSAR Table 15.0.3-3 RCS operating pressure 2250 psia Full power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attached to WEC-SNC letter GP-1 8006 and Table 7.8-3 of WCAP-1 6736-P 10.11. Fuel Assembly outside dimensions

= 8.424" x 8.424"

Reference:

1 X6AN09-1 0000-2 & 2X6AN09-1 0000-0 12. Core effective diameter = 132.7 inches x 1 footl12 inches = 11.06 ft

Reference:

Table 5-1, page 5-4, 1/2X6AAI10-00095 Source Terms Southern Nuclear Operating Company SOUHERN A1A Plant: VEGP I X6CNA15 COMPANY Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SET1 13. Reg Guide 1.183 Core Release Fractions Noble Gases (Xe & Kr): 1.0 Halogens (I & Br): 0.4

Reference:

Tables 2 and 5, Reg Guide 1.183 14. Core Fission Product Radionuclide Inventory

& RCS Coolant Equilibrium Activity Isotope Core Inventory RCS Activity Kr-85 I1.04E:+06 8.37 Kr-85m 2.68E:+07 2.04 Kr-87 4.93E:+07 1.28 Kr-88 7.02E:+07 3.68 Xe-131m 7.13E:+05 2.02 Xe-133 2.12E:+08 256 Xe-133m 3.01E:+07 17.6 Xe-1 35 4.65E::+07 8.30 Xe-135m 4.18E:+07 0.56 Xe-138 1.69E:+08 0.74 1-131 1.03E+08 2.91 1-132 1.501E+08 2.96 I-133 2.10E+08 5.56 I-134 2.26E:+08 0.69 1-135 1.95E:+08 2.72 15. RCS Coolant Dose Equivalent 1-131 Activities Isotope 1.0 DE I- 60 iLCilg DE 1-131 131 Coolant Coolant Activity Activity (jltCi/g) (jiCilg)I-131 0.74 44.4 1-132 0.75 45.0 1-133 1.41 84.6 1-134 0.18 10.8 1-135 0.69 41.4

Reference:

Page 2 of LTR-CRA-06-179 RI attached to WEC-SNC letter GP-1 8006 and Table 7.8-1 of WCAP-1 6736-P Note: Per section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gas activity in the reactor coolant assumes 1 %failed fuel, which closely equals the LCO limit of pCi/gm for gross specific activity.

Reference:

Page 3 of LTR-CRA-06-1 79 RI attached to WEC-SNC letter GP-1 8006 and Table 7.8-2 of WCAP-1 6736-P Note: Iodine concentrations developed using thyroid DCFs from FGR #11.

Southern Nuclear Operating Company SOUTaN 4 Plant: VEGP X6N1 COMPANY Unit: 1&2 Til:NI9-1Rev 6 EAL Calculations SHEET 12 16. Deleted 17. Core Fission Product Inventory

-Fuel Rod Gap Isotope Fuel Rod Isotope Fuel Rod Isotope Fuel Rod Gap Gap Gap Inventory Inventory inventory_ _ _ _ _ (Ci) (Ci) (Ci)I-131 1.03E+07 Kr-85m 2.68 E+06 Xe-131 m 7.13E+04 1-132 1.50E+07 Kr-85 3.12E+05 Xe-133m 3.01E+06 1-133 2.10E+07 Kr-87 4.93E+06 Xe-133 2.12E+07 I-134 2.26E+07 Kr-88 7.02 E+06 Xe-135m 4.18E+06 1-135 1 .95E+07 Xe-1 35 4.65E+06 Xe-I138 1 .69E+07

Reference:

Table 15A-3, VEGP FSAR, Revision 19, February 2014 Update Dose Rate vs. SFP Water Depth 18. VEGP SFP Irradiated Fuel Dose Rate vs. Water Depth above fuel See Assumption

1. 8 for discussion of adjusting this data in Attachment E2 for use with irradiated fuel in Vogtle RPV Depth Dose Rate (ft) (mREM/hr)8 1 .2696E+04 10 6.3753E+02 11.1 1.2712E+02 12 3.1412E+01 14 1.8272E+00 16 1.1519E-01

Reference:

Appendix D, X6CDE.01 Bases:* 193 fuel assemblies in spent fuel storage racks* 100-hours after S/D* Equivalent cylinder diameter = 13.7 feet* Core source term multiplied by 0.72 to account for larger cross sectional area of effective cylindrical source in SFP Release Path Flow Rates 19. SJAE via Turbine Building Vent Flow rate = 900 CFM (4.25E+05 mL/s)Release Elevation

= Ground Level X/Q = 2.55 E-06 sec/in 3

Reference:

Table 3-4, VEGP-ODCM~

20. Plant Stack Vent Southern Nuclear Operating Company SUH , Plant: VEGP I X6CNA15 SnitTH1&2 TiteNE990Rv6EA Clua ion SHEET 13 I COMPANY Unt &Til:EI9-1Rv6ELCluaos U1 Flow rate = 187,000 OEM (8.83E+07 mL/s)U2 Flow rate = 112,500 CEM (5.31 E+07 mL/s)Release Elevation

= Mixed-Mode X/IQ = 4.62E-07 sec/rn 3

Reference:

Table 3-4, VEGP-ODCM Southern Nuclear Operating Company SOUHERM Plant: VEGP X6N1 SOTE~Unit:

t &2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET1 Conversion Factors 21. FGR 12 Effective Dose Equivalent (EDE) Dose Conversion Factors for external exposures Isotope EDE Air EDE Air Immersion Immersion DCF DCF (Svlsec)l (mREMIhr)/

______(Bqlm^3)

Kr-85 1.19E-16 1.59E+03 Kr-85m 7.48E-15 9.96E+04 Kr-87 4.12E-14 5.49E+05 Kr-88 1 .02E-1 3 1 .36E+06 Xe-131m 3.89E-16 5.18E+03 Xe-133 1.56 E-15 2.08 E+04 Xe-133m 1.37E-15 1.82 E+04 Xe-135 1.19E-14 1.59E+05 Xe-135m 2.04E-14 2.72 E+05 Xe-138 5.77E-14 7.69E+05 I-131 1.82E-14 2.42 E+05 I-132 1.12E-13 1.49E+06 1-133 2.94E-14 3.92E+05 I-134 1.30E-13 1.73E+06 1-135 7.98E-14 1.06E+06

Reference:

"Effective" column, Table 111.1, "Dose Coefficients for Air Submersion," in FGR 12, "External Exposure to Radionuclides in Air, Water, and Soil" Geometry:

semi-infinite (i.e., infinite hemispherical) cloud source, air density = 1 .2 kg/in 3.To derive coefficients for other densities, multiply DCFs by (1 .2/p), where p = air density in kg/rn 3 EDE DCF EDEODCE (REM/hr) = (Svlsec) x 100 REM x 3600 sec x 1 Bq x 1.0 Ci x 1(Bq/m^3) 1 Sv 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2.703E-11 Ci 1.0E+06 j, Ci 1 m^3 1 .0E+06 cc EDE DCF (mREM/hr)(1.iCilcc)EDE DCF= (REM/hr)x 103 mREM 1 REM Southern Nuclear Operating Company SOUMANY Uk lnit: VEGPI Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 SoUHer UntI& SHEET 15 22. FGR 11 CEDE Dose Conversion Factors Isotope CEDE Air CEDE Air Thyroid CDE Air Thyroid CDE Air Inhalation Inhalation Inhalation Inhalation DCF DCF DCF DCF (SvlBq) (Sv/Bq) (mREM/jItCi)

Kr-85 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-87 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-88 0.00E+00 0.00E+00 0.O0E+00 0.00E+00 Xe-131 m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-1 33 0.00E+00 0.00E+00 O.00E+00 0.00E+00 Xe-1 33m O.OOE+OO 0.00E+00 0.00E+00 0.00E+O0 Xe-1 35 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-I135m O.OOE+O0 0.00E+00 0.00E+00 0.00E+00 Xe-I138 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-131 8.89E-09 3.29E+01 2.92E-07 1.08E+03 I-132 1.03E-10 3.81E-01 1.74E-09 6.44 E+00 I-133 I.58 E-09 5.85E+00 4.86E-08 1.80E+02 1-134 3.55E-11 1.31E-01 2.88E-10 1.07E+00 1-135 3.32E-10 1.23E+00 8.46E-09 3.13E+01

Reference:

"Table 2.1, Federal Guidance Report 11 CEDE DCF: Column labeled "Effective" Thyroid CDE DCF: Column labeled "Thyroid" Per page 121, FGR-1 1: CEDE DCF (mREM/piCi)

= 3.7x109~ x CEDE DCF (Sv/Bq)

Southern Nuclear Operating Company SOUTHER£ Ulnit: VEGP X6N1 U lnit: VE&2 Title: NEI 99-01 Rev 6 EAL Calculations

'SET1 23. Unit Conversions Conversion Reference secant = 1/cosine Page 2-25, "Marks' Standard Handbook for Mechanical Engineers" 1 atmosphere

= 76 cm Hg = 14.7 psia Page F-303, "CRC Handbook of Chemistry

& Physics" 1 cubic foot = 0.02831 6847 cubic meters Page F-308, "CRC Handbook of Chemistry

& Physics" I day = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Page F-308, "CRC Handbook of Chemistry

& Physics" 1 foot = 12 inches Page F-310, "CRC Handbook of Chemistry

& Physics" 1 foot = 0.3048 meter Page F-31 0, "CRC Handbook of Chemistry

& Physics" 1 foot = 30.48 centimeters (cm) Page F-31 0, "CRC Handbook of~Chemistry

& Physics" 1 gallon = 3.7854118 liters Page F-31 1, "CRC Handbook of Chemistry

& Physics" 1 gram = 0.001 kilogram Page F-312, "CRC Handbook of Chemistry

& Physics" 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> = 3600 seconds Page F-31 3, "CRC Handbook of Chemistry

& Physics" 1 milliliter

= 1 cubic centimeter Page F-31 8, "CRC Handbook of Chemistry

& Physics" 1 pound = 453.59237 g Page F-320, "CRC Handbook of Chemistry

& Physics" 1 pound/cubic foot = 0.016018463 g/cc Page F-321, "CRC Handbook of Chemistry

& Physics" 1 year = 365.25 days Page F-325, "CRC Handbook of Chemistry

& Physics" I Becquerel (Bq) = 2.703x1 0-11 Curie (Ci) Page 22, Lamarsh, "Introduction to Nuclear Engineering" 1 Sievert (Sv) = 100 REM Page 404, Lamarsh, "Introduction to Nuclear Engineering" Temperature Scale Conversions Page 16, Holman, "Heat Transfer'0 F = 1.8 x °C + 32°R = F +460 _____________

Miscellaneous Design Inputs 24. Iodine boiling point = 184 °C = -363 °F

Reference:

Page B-I, "CRC Handbook of Chemistry

& Physics" 25. Density of Refueling Cavity and Spent Fuel Pool Water @ 130 °F = 61.55 Ibm/cu ft

Reference:

See Attachment C2.

Southern Nuclear Operating Company SOUHERNA4t Plant: VEGP II X6CNA15 SOMPNY Uit 12 Title: NEI 99-01 Rev 6 EAL Calculations SET1 26. Deleted 27. Deleted Radiation Monitor System Parameters

28. VEGP Digital Radiation Monitoring System (ORMS) post-accident monitors required sensitivity

& accuracy shall be in accordance with Reg Guide 1 .97 R2.

Reference:

Sections 2.6 & 3.1.23, Specification X6AZ01A, VEGP 1&2 Digital Radiation Monitoring System (DRMS)29. Deleted Southern Nuclear Operating Company IsoutHERN.

Plant: VEGP TteNE990Rv6EACacliosX6CNA1 5 COMpa~nY Unit: 1&2 TteNI990Re6EACacliosSHEET 18 ASSUMPTIONS

1. TEDE and Thyroid dose calculations based on one hour of inhalation Justification:

Page 33, NEI 99-01 Revision 6 2. Breathing rate = 3.47x10-A m 3/sec Justification:

VEGP-FSAR Table 15.6.3-8 and Section 4.1.3 of Reg Guide 1.183 3. The following partition factors are assumed to determine release activities Radionuclide PF Justification Noble Gases 1.0 NUREG-0017 (PWR-GALE), Section 1.5.1.8 Iodines Steam Generator 0.01 VEGP FSAR Table 15.2.6-1 (sheet 2 of 2)NUREG-0017 (PWR-GALE), Section 1.5.1.8 Air Ejector 1 .0E-04 FNP FSAR Table 12.2-1 Reasonable to assume Iodines behavior in VEGP Main Condenser same as in FNP Main Condenser.

Liquid leakage to Auxiliary 0.01 VEGP FSAR Table 12.2.2-1 (sheet 1 of 2)Building (Round up from 0.0075)4. No noble gases are retained in the S/G: i.e., all noble gases leaked to the secondary system are continuously released with steam through the SJAE.Justification:

VEGP FSAR sections 15.2.6.3.1.3 (page 15.2-11), 15.3.3.3.1.3 (page 15.3-9), & 15.4.8.3.1.3 (page 15.4-31)5. Core inventory release fractions Noble gases: 1.00 Iodines: 0.40 Justification:

Table 2, "PWR Core Inventory Fraction Released into Containment," page 1.183-14, Regulatory Guide 1.183 6. Specific volume of steam release = 26.804 cu ft/Ibm (ASME Steam Tables)Justification:

Specific volume of saturated steam at atmospheric pressure (Attachment C2)7. Specific volume of main steam = 0.4461 cu ft/Ibm (ASME Steam Tables)

Southern Nuclear Operating CornpanyPlant: VEGP ITteNE990Re6EACacliosI X6CNA1 5 SOUTPANY Unit: 1&2 Til:NI9-1Rv6 LCluaion SHEET 19 Justification:

Full power main steam pressure = 1000 psia (VEGP FSAR Table 10.3.2-1, sheet I)8. The VEGP SEP dose rate at water surface vs. SFP water depth assessment in Appendix D of calculation X6CDE01 is acceptable for estimating the water surface dose rate vs. depth for fuel in the reactor vessel at Vogtle.Justification:

• The source is assumed to consist of an offloaded core (193 fuel assemblies) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.*The VEGP SFP analysis assumed each fuel assembly's source term was spread across a storage cell cross sectional area of 110.03 sq in. This reduced the source term (MeV/sec-cc) by a factor of 0.72.*Adjusting the source term back to a geometry matching the closer spacing in the reactor increases the source term by 1/0.72, or 1.39. Since dose rate is proportional to source strength, the VEGP SFP dose rates are multiplied by 1.39.* The effective diameter of the VEGP SFP cylindrical source is 13.7', compared to the Vogtle reactor core effective diameter of 11.06 ft (Design Input #12). This effect is evaluated in Attachment E2 of this calculation.

9. The reflected dose rate at the operating deck area radiation monitors will be calculated using the methods of Davisson's "Gamma Ray Dose Albedos" (copy in Attachment C1).The calculation will be based on an iron reflector at the top of containment, with a diameter equal to the reactor pressure vessel inside diameter (RPV ID), and a distance (r feet) from the reflector to the radiation monitor equal to the hypotenuse of the triangle formed by the difference in elevations of the reflector (177.75 feet = 397'-9" -220"-0";

Design Input #5)and the monitor and one-half of the Containment ID (70 ft = 1/2Ax 140 feet; Design Input #5).Justification:

The iron reflector is selected because the Vogtle containment has a carbon steel liner. The reflected dose rate is proportional to the area of the reflector.

Assuming the reactor vessel functions as a collimator with reduced ROS inventory will reduce the reflected area. This in turn reduces the dose rate at the radiation monitor and, therefore, the EAL threshold for reduced RCS inventory.

Simplified diagram, based on 1X2D48E008 (2X2D48E008 dimensions same)10. The RCS concentration of 1-131 (lo) at a known fuel clad defect level (Do) may be used to determine the defect level (Di) at another RCS 1-131 concentration (1i) with a simple ratio: Dlh= Do/Io Di = (li/lo) x Do Justification:

Sheet 4-2 of WCAP-8253 (excerpt in Attachment G1 of this calc)11. The temperature of the air in containment is -235 0 F Justification:

VEGP FSAR Figures 6.2.1-4 & 6.2.1-5 12. The pressure of the air in containment is -,30 psig = -45 psia Justification:

VEGP FSAR Figures 6.2.1-1 & 6.2.1-2

13. The density of the air in containment (pair) at 45 psia and 235 0 F is 2.811 kg/rn 3 Justification:

Air is assumed to behave as an ideal gas. The ideal gas law (page 3-44,"Marks' Standard Handbook for Mechanical Engineers")

is (P1 X V1)IT1 = (P2 X V2)/T2 '4 v2 = Vl x (PI/P2) x (T2/'rl)where P = pressure (psia)T = temperature

(°R)v = specific volume (ft 3/lbm)subscript 1 = reference condition subscript 2 = containment conditions, 45 psia & 235 °F (695 °R)The reference conditions for the air density of FGR-12 Table II1.1, 1.2 kg/rn 3 , correspond to 20 °C (68 0 F =528 R) and 1 atmosphere (14.7 psia)per the dry air density table on page F-11 of "CRC Handbook of Chemistry

&Physics" v2 = (1/1.2 kg/in 3) x (14.7 psia/45 psia)x (695 0 R/528 0 R)v2 = (0.8333 m 3/kg) x (0.3267) x (1.3068)v2 = 0.3583 m 3/kg pair2 = 1/v2 = 1/(0.3583 m 3/kg)pair2 = 2.791 kglm 3 DENSiTY 01' DRY AIR A2 -= TmulaarwaI

~ii uz m Pmssm. H xow Maou smr thism b1, w, pi m, St (=~i 76l.? c m b H tel PhaqOn o.pulbi Pains Him Cern a 72.0 78.0 [~j 78.0 EOIj ~I1 12 14 16l 12 18 13 0.01~1 o.oo1141 171 o.oolnel 181 loi 144 141 0.001101 I11 101 2.001181i 1uN 161 172 ill 0.00118 184 1.t6 p,00t121 210 S20 1M 186 181 17"7 168 0.001113 143 146 142 186 0.601331 221 201 211 214 0.601.214 201 201 101 141 341 141 381 m 181 127 175 011*166 161 157 158* .061324 1321 1311'3.001W 101 141 141 0.001141 Ii'171 141 0.001264'U 281 281 244 0.001241 231-I.221'U I3.00I22~r 214 211'U 004 0.001204 134 161 141 184 17 0.1 0.8 0.4 0.8 0.8 0.7 0.8 0.3 16 0.1 0.4 0.5 0.6 0.7 0.8 0.3 15 em 0.1 0.3 0.4 0.6 0.6 0.7 0.8 0.3 2 8 8 10 11 18 14t 2 8 8t 7 8 10 1,2 14 15t 1 I 6 7 9 12t 11 0.0 110410.6011110.001184iS 110118010.00116810.001180 I-IT368 Fz=20C-0.0012 g/mil- 1.2 kglm 3 Southern Nuclear Operating Company souTiERNA U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 REFERENCES

1. VEGP 1 &2 FSAR, Revision 20, December 2015 Update 2. VEGP-1 Technical Specifications, Amendment 177, December 2015 Update 3. VEGP-2 Technical Specifications, Amendment 158, December 2015 Update 4. X6CNA14, V7.0, "[VEGP] NEI 99-01 EAL Calculations" 5. X6CNA16, V1 .0, "[VEGP] NEI 99-01 Revision 4 Fission Product Barrier EAL Setpoints" 6. SNC024-CALC-005, Rev. 0, "Vogtle EALs RAl Threshold to Address NEI 99-01 Revision 6" 7. SNC024-CALC-004, Rev. 0, "VEGP Determination of Emergency Action Level for Initiating Condition E-HUt" Methods 8. NEI 99-01 Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012, (http://pbadupws.nrc.ciov/docs/MLi232/M LI2326A805.pdf)

9. Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (http ://pbadupws.nrc..qov/docslMLOO371MLOO3716792.pdf)

10. Regulatory Guide 1.195, Revision 0, "Methods and AssumPtions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003 (http://pbadupws.nrc..qov/docs/M L0314/ML031490640.pdf)

11. NUREG-001 7, Revision 1, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1985 Reg Guide 1.197 12. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 (http://pbadupws.nrc.qov/docs/MLO6071MLO60750525.pdf)

System Specifications Containment

& Containment Penetration Specifications

13. DC-1818, Revision 6, "Electrical Penetration System" 14. DC-2101, Revision 5, "Containment Building" 15. X3AR01-E3, Revision 6, "[Specification for] Reactor Containment Electrical Penetrations for Vogtle Electric Generating Plant Units 1&2" 16. X4AQ10, Revision 7, "Specification for Pipe Penetrations for [Vogtle Electric Generating Plant Units 1&2]" Radiation Monitoringq System Specifications

17. X6AZO1A, VI.0, "Specification for Digital Radiation Monitoring System (DRMS) for Alvin W. Vogtle Electric Generating Plant -Units 1, 2," 31 August 2007 Southern Nuclear Operating Company SOUTHERNA Plant: VEGP Til:HEE9-TRv 2A3CluatosX6N1 CPNY Unit: 1&2 TileHEEE-0TRv6 A2Clcltin Emergency Plan & Procedures

18. VEGP-EPLAN, Revision 65, "Vogtle 1&2 Emergency Plan" (available at SNC Emergency Planning website, http:l/n uclear.southernco.com/support-services/emerqency-planning/default.html)

19. NMP-EP-1 10-GL03, V5.2, "VEGP EALs -l~s, Threshold Values, and Basis", November 2014.Procedures Dose Calculations

20. VEGP-ODCM, V30.0, "Vogtle Offsite Dose Calculation Manual" (available at SNC Regulatory Affairs Vogtle Licensing Documents website, http://n uclear.southernco.com/regulatory-affairs/Voqtle-Licensingq-Documents.

html)Radiation Monitoringq System Setpoints 21. 34306-C, VI19.1, "Operation of DRMS Steam Generator Blowdown Liquid Process Monitor RE-002 1" 22. 34310-C3, V21.0, "Operation of DRMS Turbine Building Drain Liquid Effluent Monitor RE-0848" 23. 34311-C, V28.0, "Operation of DRMS Liquid Release Monitors 1(2)RE-0018" 24. 34333-C, V12.0, "Gaseous Effluent Monitor Setup For Releases" Refueling Containment Integrity 25. 14210-1, V18.4, "Containment Building Penetrations Verification

-Refueling" Drawings Shared Drawingqs 26. AX4DR023, V4.0, "Volumes & Water Elevations in the Primary System" Unit I Drawings 27. 1X2D01A001, Revision 5, "Containment Concrete Forming General Arrangement" 28. 1X2D48E007, Revision 4, "Containment Internals General Arrangement Section Looking North" 29. 1X2D48E005, Revision 5, "Containment Internal General Arrangement Plan at EL 220-0" 30. 1X2D48E008, Revision 4, "Containment Internals General Arrangement Section Looking West" 31. 1X2D48E009, Revision 2, "Containment Internals Slab Thickness Plan at EL 220-0" 32. 1X4DL4A013, Revision 7, "Containment Building Unit I Containment Wall Pipe Penetration Design List" 33. 1X4DL4A014, Revision 9, "Containment Building Unit I Containment Wall Pipe Penetration Design List"

34. 1X4DL4A017-1, V25.0, "Containment Bldg. Piping Area 4A, B, C & 0 Level B -Plan &Sections Reactor Coolant Loops" 35. 1X5DS4B002, Revision 9, "Instrument Location Drawing Area 4B -Levels 1 & 2 Containment Building Plan -EL 220-0 to EL 238-0" 36. 1X5DS4D002, Revision 8, "Instrument Location Drawing Area 4D -Level 1 Containment Building Plan -EL 220-0 to EL 238-0" 37. 1X6AN09-1 0000, Revision 2, "Top and Bottom 17 X 17 High Burnup Fuel Assembly Outline& Reprocessing" Unit 2 Drawingqs 38. 2X2D01A001, Revision 5, "Containment Concrete Forming General Arrangement" 39. 2X2D48E005, Revision 6, "Containment Internal General Arrangement Plan at EL 220-0" 40. 2X2D48E007, Revision 0, "Containment Internals General Arrangement Section Looking North" 41. 2X2D48E008, Revision 1, "Containment Internals General Arrangement Section Looking West" 42. 2X2D48E009, Revision 0, "Containment Internals Slab Thickness Plan at EL 220-0" 43. 2X4DL4A01 3, Revision 5, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" 44. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" 45. 2X4DL4A017-1, V18.0, "Containment Bldg. Piping Area 4A, B, C & D Level B -Plan &Sections Reactor Coolant Loops" 46. 2X5DS4B002, Revision 3, "Instrument Location Drawing -Containment Building -Area 4B-Level I Plan -EL 220-0 to EL 238-0" 47. 2X5DS4C002, Revision 3, "Containment Building -Instrument Location Drawing Area 4C-Level 1 Plan -EL 220-0 to EL 238-0" 48. Deleted 49. 2X6AN09-1 0000, Revision 2, "Top and Bottom 17 X 17 High Burnup Fuel Assembly Outline& Reprocessing" Calculations

50. X6CDE.01, V5, "[VEGP] Spent Fuel Pool Shielding" 51. X6CNA11, V10.0, "Severe Accident Management Guideline (SAMG) Calculations" Source Term 52. WCAP-1 6736-P, Revision 1, "Vogtle Electric Generating Plant Measurement Uncertainty Recapture Power Uprate Program Engineering Report," May 2007 53. GP-1 8006, "Vogtle Electric Generating Plant Units 1 and 2 -Revised Source Terms for Measurement Uncertainty Recapture Power Uprate Program," 21 September 2006 Southern Nuclear Operating Company souTH=l,,'Pat V IG HEETNE 201Rv6EA acuainsXCA5 Um lnit: VEGP SHCNA25 I COMPANY' Ui:12 Tte E 901Rv6ELCluain

54. VEGP 1&2 Technical Specifications Bases, Revision 36, 29 December 2015 Dose Conversion Factors 55. Federal Guidance Report #11 (EPA 520-1-88-020), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," September 1988 (http://nepis.epa..qov/Simple.

html)56. Federal Guidance Report #12 (EPA-402-R-93-081), "External Exposure to Radionuclides in Air, Water, and Soil," September 1993 (http://nepis.epa.,qov/Simple.html)

GRODEC Computer Code 57. F-86-03, Revision 3, "Verification of the GRODEC Computer Program" Engineering References

58. ANSI/ANS-6.4.3-1991, "American National Standard for Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials," 26 August 1991 59. Davisson, "Gamma Ray Dose Albedos," pages 5-27 thru 5-38 in ANS/SD-76/14, A Handbook of Radiation Shielding Data, edited by J. C. Courtney, July 1976, (copy in Attachment C1)60. Holman, "Heat Transfer," fourth edition, 1976 61. Lamarsh, "Introduction to Nuclear Engineering," second edition, 1983 62. Singer, "Strength of Materials," second edition, 1962 63. CRTD-VOL 58, "ASME International Steam Tables for Industrial Use," second edition, September 2008 64. "CRC Handbook of Chemistry

& Physics," 5 7 th edition 65. "Marks' Standard Handbook for Mechanical Engineers," 8th edition 66. Geick & Geick, "Engineering Formulas," seventh edition 67. Murphy & Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," pages 401 -430, CONF-740807, "Proceedings of the Thirteenth AEC Air Cleaning Conference," Ma~rch 1975 (http:llwww.osti.qov/scitech/servletslpurl141 79572)68. Jaeger (editor-in-chief), "Engineering Compendium on Radiation Shielding:

Volume I: Shielding Fundamentals and Methods," 1968 69. Shultis & Faw, "Fundamentals of Nuclear Science and Engineering," 2002 Southern Nuclear Operating Cornpany SOUTH NY' Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 26 METHOD OF SOLUTIONS NEI 99-01 Revision 6 Methods conform to the guidance of NEI 99-01 Revision 6. Detailed descriptions of the methods are included in the individual EAL threshold calculations in the Analysis section of this calculation.

Use of Regulatory Guide 1.183, Alternate Source Term Method The NEI 99-01 Revision 6 Recognition Category A (Abnormal Rad Levels/Radiological Effluent)Initiating Conditions (l~s) for declaring an Alert, a Site Area Emergency and a General Emergency (Emergency Action Levels RS1 and RG1, respectively) are expressed in terms of Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (ODE).Regulatory Guide 1 .195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," is the current license basis for performing dose calculations for Vogtle. However, it expresses doses in terms of Whole Body and Thyroid.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," does express doses in terms of TEDE and CDE, but it is not the current licensing basis for performing dose calculations for Vogtle. However, per section 1.1.4 on page 1.183-6, "This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with emergency dose projections, protective measures, and severe accident management guides." Per section 4.1.1 of RG 1.183, TEDE is defined as the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure.Per section 4.1.2 of RG 1.183, Table 2.1 of Federal Guidance Report 11 provides tables of conversion factors acceptable to the NRC staff. The dose conversion factors (DCFs) factors in the column headed "effective" yield doses corresponding to the CEDE.Per sections 4.1.4 and 4.1.5 of Reg Guide 1.183, the DDE should be calculated assuming submergence in a semi-infinite cloud for the most limiting person at the EAB. The effective dose equivalent (EDE) from external exposure is nominally equivalent to the DDE, thus EDE may be used in lieu of DDE in determining the external dose contribution to the TEDE. Table II].1 of Federal Guidance Report 12 provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.TEDE & Thyroid CDE Calculations The basic process for calculating an offsite dose consists of first determining the concentrations of radionuclides in the release stream, be it air, steam, or water. The release stream concentration is determined by dividing the release rates of the radionuclides of interest, expressed as microcuries per unit time, by the release fluid's volumetric flow rate, expressed as cubic centimeters (cc) per unit time:

Southern Nuclear Operating Cornpany SOUTHaEm 41 Plant: VEGP Title: N M EI 99-01 Rev 6 EAL Calculations X6CNAI 5 COMPANY Unit: 1&2 SHEET 27 liCi/cc =

time]/[cc/unit time]As we are back-calculating release concentrations based on pre-established dose limits (100 mREM TEDE and 500 mREM Thyroid CDE), the upstream modeling of the specific release paths is not necessary.

The gaseous effluent noble gas radiation monitors care not a whit how those radionuclides arrive at them.Step 1: Identify the radionuclides of interest.

Select the same radionuclides used to calculate doses for the design basis accidents in FSAR chapter 15: the fission product noble gases and iodines. The other fission products and activated corrosion products are particulates and will not contribute significantly to the offsite dose.Step 2: Determine the ROS coolant radionuclide activity for each radionuclide (Xrcs-i pCi/g). This is assumed to be the sum of core fission product inventory released during a LOCA divided by RCS coolant mass (Mrcs g) and the equilibrium RCS coolant activity (Xeq-i) for that rad ion uclide.Xrcs-i = Xeq-i + (1 .0E+06 Ci) x [Core Inventory (Ci)] x [Release Fraction]/(Mrcs g)For no fuel damage, the release fraction is 0 and the RCS activity is the equilibrium RCS coolant activity.

If fuel damage is assumed (release fraction > 0), the quotient of core inventory and RCS coolant mass will be orders of magnitude greater than the contribution from the coolant equilibrium activity.Step 3: Convert coolant activity (Xrcs-i to release stream activity (Xris-i jiCi/cc).

This conversion is accomplished by multiplying the ROS coolant activity by a dimensionless partition factor (PFi) and an arbitrarily selected density, prls g/cc: Xrls-i

= (Xrcs-i x PFi x (pris g/cc)The partition factor will depend on the radionuclide and the release path. The partition factors used in this calculation are discussed in Assumption

1. 3 of this calculation.

Arbitrarily set pris = 1.0 g/cc to make the math easy. The justification for this will be provided in Step 9.Step 4: Determine radionuclide concentration at Exclusion Area Boundary (XEAB-I liCi/cc).

This is done using standard dose assessment methods. The release concentration is multiplied by the release volumetric flow rate (Qris m 3/sec) and the diffusion coefficient

[(X/Q) m 3/sec]: XEAB-i

= Xris-i x [Qris (m 3/sec)] x [(X/Q) (m 3/sec)]Step 5: Calculate the TEDE for each radionuclide for one hour exposure time at EAB. This is done using the appropriate FGR-11 and FGR-12 dose conversion factors (DCFs), as discussed in the previous subsection.

TEDE 1 (mREM) = External Exposure + Internal Exposure TEDEi (mREM) = XEAB-I (pCi/cc) x texp (hours) x DDEDcF-i

+XEAB-i (j, Ci/cc) X texp (hours) x BR (cc/hr) x CEDEDCF-I TEDEi (mREM) = XEAB x texp (hours) x TEDEDcF-i

[(mREM/hr)/(pCi/cc)]

where TEDEDcF-i

[(mREM/hr)/(p.Ci/cc)]

= DDEDcF-i [(mREM/hr)l/QiCi/cc)]

+BR (cc/hr) x CEDEDcF-I (mREM/pCi)

BR (cc/hr) = breathing rate Step 6: Add the individual TEDEs to obtain the TEDE for the release (TEDEris):

TEDEris = [TEDEI]TEDEris = [Xris-i x (X/Q) x Qris X texp X TEDEDcF-i]

TEDEris = [(X/Q) x Qris X texp ] X [Xrls-i X TEDEDcF-i]

Step 7: Calculate Thyroid ODE for each Iodine isotope for one hour exposure time at EAB. This is done using the appropriate FGR-11I dose conversion factors (DCFs), as discussed in the previous subsection.

CDETHY-i (mREM) = XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CDETHY-DCF-i (mREM/4LCi)

Step 8: Add the individual Thyroid ODEs to obtain the Thyroid ODE for the release (CDEris): CDEris = [CDETHY-i]

CDEris = [Xrls-i X (X/Q) x Qris X BR X CDETHY-DCF-i]

CDEris = [(X/Q) x Qris x texp X texp] X [Xrls-i X TEDEDcF-i]

Step 9: Determine the 100 mREM TEDE threshold release concentrations for each noble gas (Xi oo-i This is done by multiplying each noble gas' release concentration (Xris-i #C~i/cc)determined in Step 3 by the quotient of 100 mREM and the sum of the TEDEs for all of the radionuclides considered (TEDErls mREM). Only noble gas concentrations are adjusted because the gaseous effluent monitors are noble gas detectors.(Xia0-i p.Oi/cc)/(Xris-i pCi/cc) = (100 mREM)/(TEDEfls mREM)Xioo-i (liCilcc)

= (Xrls-i jiC i/cc) x (100 mREM)I(TEDEris mREM)The following demonstrates that the arbitrarily assumed release stream density has no effect on the final result.Xris-i x (100 mREM)XlOO-1 =[(X/Q) x Qris x texp ] X Z [Xrls-i X TEDEDcF-i]

= Xrcs-i X (1.0) X pris X (100)Xrcs-i X prls X (100)Xpr001x (X/Q Qris x texp X T- [Xrcs-i X PFi X TEDEDcF-i]

Southern Nuclear Operating Company SOUTHERZ U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CN15~COMPANY Pln:VGI Xi 0-iXrcs-i X pr-le X (100)paIs X (XIQ) x Qris x [Xrcs-i X PFi X texp X TED EDcF-i]= Xrcs-i X (1 00)(X/Q) x Qris x texp X [Xrcs-i X PFi x TEDEDcF-I]

The assumed release stream density has no effect on the final result: it cancels out. Thanks to the power of Excel, it is easier to calculate a postulated dose rate and adjust release concentrations than to set up the above equations.

Now to perform a dimensional check:= Xris-i X (100 mREM)1, [Xris-i X (X/Q) x Qris x PFi x texp x TEDEDcF-i]

"7 (jiCi/cc) xmREM-(pCi/cc) x (sec/in 3) x (m 3/sec) x (hour) x [(mREM/hour)/(pCi/cc)](= ./,,G,,G}.

v x (fna/seG) x x [(mREM/h~eu-)I(iCi/cc)]= 1/(pC i/cc)]J4Ci/cc = Step 10: Determine the 500 mREM Thyroid CDE threshold release concentrations for each noble gas (X500T-i ptCi/cc).

This is done using the same method as in Step 9. Again, the arbitrarily assumed release stream density cancels out and has no effect on the final result.Several general trends can be inferred from the equation derived in Step 9 above. Holding other factors constant:* Increasing the diffusion coefficient (XIQ m 3/sec) will reduce the 100 mnREM release concentration.

• Increasing the release flow rate (Q CFM) will reduce the 100 mREM release concentration.

• Increasing the exposure time (t hours) will reduce the 100 mnREM release concentration.

• Increasing the total release [Xrcs-i x PFi x pris X TEDEDcF-I])

will reduce the 100 mREM release concentration.

When assessing TEDE and Thyroid CDE doses against NEI 99-01 Rev. 6, initiating conditions RA1, RS1 and RGI, there are two release paths that will be evaluated:

One release path via the Plant Stack Vent and one release path via the condenser steam jet air ejector (SJAE). The pathways and their major assumptions are summarized below.

Radiation Monitor Release Path Core Damage Partition Factors RE-I12442C Plant Vent Stack Yes Noble Gases: 1.0 RE-I12444C/D/E Iodines: 0.01 RE-12839 SJAE Two Cases: With and Noble Gases: 1.0 Without Core Damage Iodines: I1.0E-06[S/G PF (0.01) x SJAE PF (1 .0E-04)]The release paths via the S/G SRV or PORV and the Turbine Driven AFWP Turbine Exhaust are not evaluated because the Main Steam Line radiation monitors (RE-i13119 thru RE-I13122) were deleted per the 30 September 2014 license amendment (ADAMS # ML14170A91 1).Effects of Plant Configuration on Dose Conversion Factors The FGR 12 Table II1.1 air immersion DCFs are based on immersion in an infinite hemispherical cloud in which the radioisotopes are dispersed.

Per page 415 of the 1 3 th AEC Air Cleaning Conference (CONF-740807 VOL 1), the dose rate at the center of a finite hemispherical cloud of volume V cu ft is the dose rate from an infinite hemispherical cloud divided by a geometry factor (GF), where GF = 1173 /V 0" 3 3 8 Depending on area radiation monitor location, a portion of the dispersed radionuclides may be shielded by structures such as the steam generator enclosures.

This reduces the finite cloud volume.

Southern Nuclear Operating Company VEGP X6N1 SOMPERAN Unit: 1,&2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET3 BODY OF CALCULATION RU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:

Emergency Action Levels: All (1 OR 2OR 3)1. Reading on ANY effluent radiation monitor greater than two times the ODCM limits for 60 minutes or longer.Liquid & Gaseous Effluent Monitors' Default Alarm Setpoints Per the guidance of NEI199-01 Rev 6 for Initiating Condition RU1 and its associated Emergency Action Level 1 (EAL 1), which are stated as follows: From page 25 of NEI 99-01 Rev 6: EAL #1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.As a release permit alarm setpoint is dependent upon the mix of radionuclides, their concentrations, the effluent release rate, the dilution stream flow rate, and the presence of concurrent contaminated releases, determining a default alarm threshold is infeasible.

The following liquid and gaseous release paths are continuous (per Tables 2-4 and 3-4 of VEGP ODCM), with release permits and alarm setpoints generated weekly (per 27AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager; copy in Attachment C3 of this calculation).

Liquid Releases Release Path Monitor SIG Blowdown RE-0021 Effluent Line Turbine Building RE-0848 Drain Effluent Line Gaseous Releases Release Path Monitor Plant Vent Stack RE-i12442C Plant Vent Stack RE-I124440 Turbine Building RE-12839 (SJAE)Thus, default setpoints are deleted from this version of the calculation.

Southern Nuclear Operating Company SOTENr Plant: VEGP Tite NI9-1Rv6ALClua ion6N1 CPNY Unit: 1&2 , SHEET 32 2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer: Liquid Effluent Release Path Monitors:

None Gaseous Release Path Monitors:

None Main Steam Line radiation monitors (RE-131 19 thru RE-13 122) were deleted per the 30 September 2014 license amendment (ADAMS # ML14170A911).

3. Sample analysis for gaseous or liquid release indicates a concentration or release rate greater than two times the ODCM limits for 60 minutes or longer.

Southern Nuclear Operating Company soutPEl Y. Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SET3 RAI: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20R3OR4 1. Reading on ANY of the following radiation monitors greater than the readings shown for 15 minutes or longer.RE-12444E 0.50 pCi/cc RE-I12839E 2.1 x 101 pCi/cc This calculation is performed in Attachment K 2. Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the site boundary.3. Analyses of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the site boundary for one hour of exposure.4. Field survey results indicate EITHER of the following at or beyond the site boundary:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer* Analysis of field survey samples indicate thyroid ODE greater than 50 mrem for one hour of inhalation.

Southern Nuclear Operating Company SUHERN, 4.Plant: VEGP X6N1 SOUT en Unit: 1&2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET3 RSI: Release of gaseous radioactivity resulting in offsite dose greater than 100 mREM TEDE or 500 mREM thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20OR3 1. VALID reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer: Radiation Vent Path Reading Monitor RE-I12444E Plant Vent -High Range 5 jiCi/cc RE-I12839E Turbine Building Vent (SJAE) -210 PCi/cc High Range Plant Vent (RE-12444E)

The threshold calculations are performed using the Excel spreadsheet in Attachment D1 of this calculation.

The thresholds calculated in the spreadsheets have been rounded off to one significant below to reflect the radiation monitoring system accuracy.100 mREM TEDE threshold

= 500 mREM Thyroid CDE threshold

=

Limiting threshold:

100 mREM TEDE = 5/,uCi/cc Steam Jet Air Ejector (RE-12 839)The threshold calculations are performed in the Excel spreadsheet in Attachment D2A (no core damage) and D2B (core damage) of this calculation.

The thresholds calculated in the spreadsheets have been rounded off to one significant below to reflect the radiation monitoring system accuracy.Dose Threshold No Core Damage Core Damage 100 mREM TEDE 2E+03 210 500 mREM Thyroid CDE 3E+07 4E+06 jiCi/cc Limiting threshold:

100 mREM TEDE, with core damage = 210 ,iCi/cc Southern Nuclear Operating Company SAT=m 1 Plant: VEGP ITteNE991Re6EACacliosj X6CNA15 S ;OUTLNY Unit: 1&2 TiteNE9-0Rv6EA Clua ion SHEET 35 2. Dose assessment using actual meteorology indicates doses greater than 100 mREM TEDE or 500 mREM thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the site boundary.* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Southern Nuclear Operating Company SUHm-- Plant: VEGP I X6CNA15 UnSO12 Til:UET9-1He 6ELacuaion SHEET 36 I COMPANY Unt & Til:NI9-1Rv6ELCluaos RGI: Release of gaseous radioactivity resulting in offsite dose greater than 1000 mREM TEDE or 5000 mREM thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20OR3 1. Reading on ANY of the following radiation monitors greater than the readings shown below, for 15 minutes or longer: Because the RG1 EALI dose limits are ten times the RS1 EALI dose limits, these threshold values are ten times the RS1 EAL1I threshold values.Radiation Vent Path RS1 EAL1 RG1 EAL1 Monitor Threshold Threshold RE-I12444E Plant Vent 5 50 j.+/-Ci/cc RE-I12839E Turbine Building 210 !.iCi/cc 2100 (Steam Jet Air Ejector)2. Dose assessment using actual meteorology indicates doses greater than 1000 mREM TEDE or 5000 mREM thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the site boundary.* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

Southern Nuclear Operating Company SOUTHE=B U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6EETNA15 CA1: Loss of RPV inventory.

Operating Mode Applicability:

Emergency Action Levels: Cold Shutdown, Refueling 1 OR2 1. Loss of RPV inventory as indicated be level less than elevation 1 85'-1 0" (73% on Full Range RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2 x ID 13.75" Bottom ID = Centerline

-(1/2Ax ID) 185'-1 0.25" Hot Leg Inside Diameter 29.0" Sx ID 14.5" Bottom ID = Centerline

-(A x ID) 185'-9.5" The dimensions and elevations are taken from Design Input #9. The RPV water level elevation corresponding to the Bottom ID of the RCS piping is ~185'10".Because the core barrel is a right circular cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearly interpolating between the TOAF (EL 181"-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):

Southern Nuclear Operating Company s Plant VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations j XCA5 S Unit: 1&2 SHEET 38 VEGP RVLIS Indication vs. RPV Water Level Elevation 80 75 IL. 70 80 183 184 185 186 RPV Water Level Elevation (feet)188 The RPV water level elevation corresponding to the Bottom~73% on Full Range RVLIS.ID is 185'-10" or 2. a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.

Southern Nuclear Operating CompanyPlant: VEGP Titl:NI9-1Rv6ELCluain6N1 5 0~PIN Unit: 1&2 Til:NI9~ e A acltosI SHEET 39 0S1: Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

Emergency Action Levels: Cold Shutdown, Refueling 1 OR20OR3 1. a. CONTAINMENT CLOSURE not established AND b. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) and hot leg (HL) bottom/IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2%x ID 13.75" Bottom ID = Centerline

-(1/2 x ID) 185'-1 0.25" 6" Below CL Bottom ID I185'-4.25" Hot Leg Inside Diameter 29.0" Sx ID 14.5" Bottom ID = Centerline

-(1/2 x ID) 185'-9.5" 6" Below HL Bottom ID 1 85'-3.5" The dimensions and elevations are taken from Design Input #9. The elevation corresponding to 6" below the Bottom ID of the RCS piping is ~185'4".Because the core barrel is a right circular cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearly interpolating between the TOAF (EL 181"-1O" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):

Southern Nuclear Operating Company SO1J, Unit: 1&2 ITitle: NEI 99-01 Rev 6 EAL Calculations I SHEET 40 VEGP RVLIS Indication vs. RPV Water Level Elevation... ..... ..... .... Centerline

-7 __ -__ 6" Below RCS __--1........ .. .. ..........

.. .Piping Bottom~ID C 70 70~I 181 1 82 183 184 185 188 18 RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is 185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established AND b. RPV level less than 181'-1 0" [TOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by ANY of the following:

RE-005 O...R 006 >40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Southern Nuclear Operating CompanyPant: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 Unit: 1&2 S HEET 41 CG1 : Loss of RPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

Cold Shutdown, Refueling Emergency Action Levels: 1 OR 2 1. RPV water Level: a. Less than TOAF: EL 181'-1 0" or 63% on Full Range RVLIS.AND b. ANY indication from the Containment Challenge Table C1.Containment Challenge Table C1 CONTAI NMENT CLOSU RE NOT established*

Explosive mixture inside containment greater than OR equal to 6% H2 greater than OR equal to 13 psig WITH CONTAINMENT Containment PressureCLSResalhd greater than O..E equal to 52 psig WITH Tech Spec containment integrity intact* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.2. a. RPV water level cannot be monitored for 30 minutes or longer: AND b. Core uncovery is indicated by ANY of the following:

Monitor Location EAL Threshold RE-0005 or RE-0006 Containment Operating

> 40 REM/hr Deck High Range Source Range N/A Erratic Indication Southern Nuclear Operating Company smHm 1 Plant: VEGP Til:NI9-1Rv6ELCluaion6N1

• ui~ Unit: 1&2 TileHEE9-0ERv6TA Clcltin UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. ANY indication from the Containment Challenge Table C1 (above).Containment Operating Deck High Range (RE-O005 or RE-O006): This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel in the RPV with an RPV water level of less than TOAF (181 '-10" or 63% on Full Range RVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Mon itor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen: Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.Pressure>_

14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per uOperating Procedure 142 10-1/2, Containment Building Penetrations Verification

-Refueling." Section 6.0 of 14210-1/2 lists the acceptance criteria for CONTAINMENT CLOSURE, among them the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

If Containment pressure (PCTMT) exceeds the static head (Al-) due to the difference between the Transfer Tube centerline elevation (EL 186"-93/4";Design Inputs #4 & #5) andPTT the SFP low operating water level (EL 217"-0"; Design Input #4), the Transfer Tube air-to-air.

barrier is not maintained.--...-

A1H (ft) = 217"-0" -186'-9. 75" = 30'-2. 25" = -30 ft Pctmt (psig) > AH (ft) x p (Ibmlft) x g (ft/sec 2) x 1 ft 2 ge (Ibm-ft)/(Ibf-sec

2) 144 in 2 Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec 2 J x 1 ft 2 ift 32.2 (Ibm-ft)/(Ibf-sec

2) 144 in 2 (Design Input #25)Pc~r > -13 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-EP- 11 0-GLO3 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement

3. 6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier. Thus, these penetrations must meet the design accident pressure requirement of section 3. 4.5 of DC-2101, 52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager; see Attachment C5) and Ron Cowen (Westinghouse Site Services Manager," see Attachment C6).

Southern Nuclear Operating CompanyPlant: VEGP X6N1 CPAY Unit: 1& il: E 90 Rev 6 EAL Calculations SHEET 44 SOU MPATil:NEI990 Piping Penetrations The piping penetrations are listed in drawings 1X4DL4A0 13, 1X4DL4A0 14, 2X4DL4A0 13, and 2X4DL4A0 14. Cross-sectional views are shown in drawings 1X4DL4A014 and 2X4DL4A0 14.Per section 4.1.2 of specification X4AQIO, these penetrations provide part of the containment boundary.

Section 4.1.3.3.2 of this specification directs the user to Attachment 2 for the design temperature and pressure for these penetrations.

Per Attachment 2 of specification X4AQIO, the emergency operation design containment pressure is 50 psig.The evaluation in Attachment F of this caic demonstrates that the pipe penetrations should not fail due to a containment pressure of 52 psig.VEGP Condition Report 876376 has been submitted to review and resolve this difference between the Containment and the piping penetration accident design pressure criteria.

There are no operability or functionality issues because the peak containment DBA pressure is ~37 psig (VEGP FSAR Tables 6.2.1-1 & 6.2.1-66).

Electrical Penetrations Per section 3.1.3 of DC-1818, the electrical penetration assemblies shall withstand the pressure, temperature, and environmental conditions resulting from a DBA without exceeding the electrical penetration design leakage rate.Per section E3. 6.2 of specification X3ARO1-E3, the electrical penetration design leakage rate is 0.01 cc/sec at DBA conditions.

CONTAINMENT CLOSURE no.t established.

Basis: NEI 99-01 Rev 6, page 81.

Southern Nuclear Operating Company SOUTHERN U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNA15 I E-HUI Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability:

ALL Emergency Action Level: 1 1. Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table El. Readings are combined gamma + neutron.Location A H I-TRAC 125 Side -Mid -height 950 Top J 200 HI-STORM 100 Side -60 inches below mid- 170 Side -Mid -height 180 Side -60 inches above mid- 110 Top -Center of lid 50 Top -Radially centered 60 Inlet duct 360 Outlet duct 130 This calculation is performed in Attachment L

Fission Product Barrier Emergency Action Levels Fuel Clad Barrier Fuel Clad Barrier Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 > 2.6E+5 mR/hr The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300pCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the fuel clad barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss Threshold 3.A since it indicates a loss of both the fuel clad barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

Evaluation of Containment High Range Rad Monitor Locations Readings at Power A review of the instrument location drawings (Unit 1:1IX5DS4BO02

& 1X5DS4DO02; Unit 2: 2X5DS4BO02

& 2X5DS4CO02) indicates the rad monitors 1/2RE-0005

& 1/2RE-0006 are located above the Containment operating deck of EL 220'-0".The following full power area rad monitor data demonstrates that the monitors do not measure radiation from nearby piping and components containing elevated reactor coolant activity.

The data are documented in Attachment G2; Attachment G3 documents that the plant was at 100%power.

Unit 1 1 RE-005 1 RE-006 Tie Reading Tie Reading Tie (mREMI~hr)

Tie (mREM/hr)12:00 505 12:00 250 12:10 484 12:10 232 12:20 509 12:20 193 12:30 511 12:30 189 12:40 502 12:40 240 12:50 499 12:50 238 13:00 493 13:00 250 Average 500 Average 227 Unit 2 2RE-005 2RE-006 Tie Reading Tie Reading Tie (mREMIhr)

Tie (mREM/hr)12:00 42.2 12:00 19.2 12:10 29.7 12:10 21.3 12:20 53.3 12:20 23.1 12:30 54.9 12:30 23.9 12:40 47.4 12:40 15.4 12:50 61.3 12:50 18.8 13:00 52.8 13:00 16.6 Average 48.8 Average 19.8 The differences between the Unit 1 and Unit 2 readings are due to their installation locations as shown on the next sheet. The Unit 1 rad monitors are mounted on the S/G enclosures on opposite sides of the reactor vessel. The Unit 2 rad monitors are mounted near the outer containment wall, further away from the reactor vessel as compared to the Unit 1 monitors.Containment Geometry While the released radioisotopes are divided by the total containment volume (Vctmt), the operating deck (EL 220'-0") area radiation monitors locations limit them to "seeing" the sprayed portion of the containment volume. This view may be further reduced by containment structures.

As a result, the containment volume used to determine the geometry factor GF is expressed as: Vmon = fview x fspray x Vctmt where Vmon = containment volume "seen" by area radiation monitor (cu ft)fvwew = fraction of containment volume above operating deck "seen" by area radiation monitor fsry= fraction of containment volume above operating deck = 0. 771 Vctmt = Containment volume = 2. 95E+06 cu ft The fraction of the sprayed containment volume seen by the area radiation monitors is estimated below.The containment operating deck area radiation monitor locations identified in drawings 1X5DS4BO02, 1X5DS4DO02, 2X5DS4BO02, and 2X5DS4CO02 are overlaid on the containment structural drawings below:

~Southern Nuclear Operating Company=UI Al 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNAI5 I s@J~ham, Unit: 1&2 SHEET 48 Unit I Unit 2 While the Unit 2 area radiation monitors have a fairly unrestricted "view" of the containment volume above the operating deck, the Unit I area radiation monitors are each screened from a significant portion. The difference between the Unit 1 and Unit 2 radiation monitor locations explain the differences between their full power readings.The fraction of the sprayed containment volume seen by the Unit I area radiation monitors is fvjew = I -A segmnen/A op deck where A segment = area of segment outboard of dashed gray line Aop, deck = area of operating deck = (Hix lDctmnt 2)/4 IDctrnt = containment inside diameter = 140 ft The area of the segment is calculated as shown below.

A circular segment (in green) is enclosed between a chord (dashed line) of length c and the arc of length s shown above the green area.R = radius of circle = % x lDctmt R = 70Oft e = central angle (degrees)c = chord length s = arc length h = height of segment d = height of triangular portion S

References:

http://en.wikipedia.orq/wiki/Circular seqment and Geick&Geick, "Engineering Formulas" The central angle is The radius is 6 = 2 arccos = 2 arcsin2-R R =h+d The height of the triangle is d =R -h Area of segment is 80o Scaling from drawing 1X2D48EO05, h = 50 ft and d = 20 ft. The chord length c is determined using the Pythagorean Theorem: (c/2)2 + d 2 = R (c/2)2 = R 2_ cl c/2 = [R 2 -d 2c = 2 x [R 2 -l1 c = 2 x [(7 0)2- (20)21 c =134 ft e = 2 x arcsin[c/2R]

e = 2 x arcsin[(134)/(140)]

e = 2 x arcsin[O.

958]eO= 2x 73.40 Southern Nuclear Operating CompanyPlant: VEGP I X6CNA15 sOuy pnay Unt1& Title: NEI 99-01 Rev 6 EAL Calculations SET5 6 = 146.80 A segment = [R 2/2] X [(917/180)

-sinS7]A segment = [(70)2/2] x [(146. 8 x 17/180) -sin(146. 8)]A segmnent = [2450] x [(2. 562) -(0. 548)]A segment = [2450] x [(2.014]A segment = 4935 sq ft A op deck = Hl1Dctmt 2/4 A op deck = H7(140)2/4 Aop deck = 15,394 sq ft fview, = I -A segmnent/A op deck fview = 1 -- (4935/15,394) fview = 1 --0.321 fview = 0.6 79 For Unit 2, fview, = 1.0O.Thus Vmon = fview X fapray x Vctmt GE = 11 73AVmon 0" 3 3 8 The calculation of the geometry factor will be performed in the spreadsheets in Attachments H and I. Performing it here using a calculator will produce different results than the spreadsheet due to round off errors.RCS Barrier RCS Barrier Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 >= 8.7E+2 mR/hr The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold 3.A since~it indicates a loss of the RCS Barrier only.Evaluation of the Fuel Clad Failure Southern Nuclear Operating Company ISOUTHERN Uk lnit: VEGPI Title: NEI 99-01 Rev 6 EAL Calculations I SHETA15 This threshold is determined assuming fuel clad failure and resultant instantaneous release and dispersal into containment the reactor coolant noble gas and iodine inventory, with RCS activity at the Tech Spec allowable limits.The iodine concentrations are set equal to the corresponding to the 1.0 ,uCi/g DE/1-131 values.The corresponding noble gas reactor coolant concentrations are expected to be proportional to the DE I-131 concentrations.

The ratio of each Iodine radionuc/ide's reactor coolant DE/1-131 concentration above is divided by its equilibrium reactor coolant concentration.

The minimum ratio is then used to multiply the noble gas equilibrium reactor coolant concentrations.

RCS Concentrations (pCi/g)Iodies quilriu 1 Ci/g (1 i.LCi/g)/lodnes Equibrum DE 1-131 Equilibrium 1-131 2.91 0.74 0.25 I-132 2.96 0.75 0.25 I-133 5.56 1.41 0.25 1-134 0.69 0.18 0.26 1-135 2.72 0.69 0.25 Minimum Ratio =0.25 RCS Concentrations (pCi/g)RCS Concentrations (pCiIg)NolI 1 Nobles Equilbrium DE 1-131 Noble Gases Equilbrium DE 1-131 GssEquivalent Equivalent Kr-85m 2.04 5.17E=-01 Xe-131m 2.02 5.121E-01 Kr-85 8.37 2.12E=+00 Xe-133m 17.6 4.46E+00 Kr-87 1.28 3.24E=-01 Xe-133 256 6.491E+01 Kr-88 3.68 9.321E-01 Xe-135m 0.56 1.421E-01 Xe-135 8.3 2.10E+00 Xe-138 ] 0.74 [1.88E-01 These concentrations

(/uCi/g) are then multiplied by the RCS coolant mass (grams) then divided by the containment volume (in 3) to determine the containment source strength (Ci/m 3).The resulting Unit I and 2 containment high range rad monitor dose rates for immersion in a finite hemispherical cloud of air are calculated in Excel spreadsheets in Attachments H2 and 12, respectively, of this calculation.

They uses the applicable FGR 12 Table III. 1 DCFs for immersion in a semi-infinite cloud of air and the geometry factor (GE) from sheet 30 of this Southern Nuclear Operating Company SOUTHEN ln:VG Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 I COMPANY Unit: 1&2 I SHEET 52 calculation to convert the semi-infinite (infinite hemispherical) cloud dose rate to a finite hemispherical cloud dose rate.The results of the Loss of RCS FP Barrier setpoint calculations in Attachments H2 and 12 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (REM/h r) (m REM/hr)VEGP 1 0. 870 8. 7E+02 VEGP 2 0. 991 9.9E+02 Containment Barrier Containment Barrier Potential Loss Threshold 3.A Containment radiation monitor RE-005 OR -006 >= 1 .3E+7 mR/hr.The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous fuel clad barrier loss and RCS barrier loss thresholds.

NUREG-1 228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20 percent in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS barrier and the fuel clad barrier. It is therefore prudent to treat this condition as a potential loss of containment that would then escalate the emergency classification level to a General Emergency.

Evaluation of the Potential Loss Threshold This setpoint is based on the release of all iodine isotopes and noble gases corresponding to 20% fuel clad failure. The core wide fuel rod gap noble gas inventories of FSAR Table 15A-3 are multiplied by this failed fuel clad fraction to determine the released activity.

This released activity is then divided by the containment volume to determine the source strength (Ci/m 3).The resulting Unit I and 2 containment high range rad monitor dose rates for immersion in a finite hemispherical cloud of air are calculated in Excel spreadsheets in Attachments H2 and 12, respectively, of this calculation.

They uses the applicable FGR 12 Table 1lL.1 DCFs for immersion in a semi-infinite cloud of air and the geometry factor (GE) from sheet 30 of this calculation to convert the semi-infinite (infinite hemispherical) cloud dose rate to a finite hemispherical cloud dose rate.

Southern Nuclear Operating Company SOUTHERN~

UPlnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNA153 The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H3 and 13 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (REMIhr) (m REM/h r)VEGP I I.31E+04 1.3E+07 VEGP 2 1.49E+i04 1.5E+07 Containment Barrier Potential Loss Threshold 4.B Containment Hydrogen concentration greater than 6%.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.

It therefore represents a potential loss of the containment barrier.Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.

Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNAI5 Sheet: A-I Attachment A -SNC Emergency Planning Concurrence Calculation Number: X6CNAI5 Calculation Version: 1 Calculation Title: NEI 99-01 Rev 6 EAL Calculations I the undersigned have reviewed the subject calculation and concur that:* Its Methods of Analysis conform to the guidance of NEI 99-01 Revision 6* Its Assumptions are consistent with the guidance of NEI 99-01 Revision 6* Its conclusions are consistent with the Methods of Analysis, Assumptions, and Design Inputs.g~4. c.II~:~-M~@d~rr 9111t'Name 1 / j(,. I SNC Emergency Planning/ Signature

/ Date / Organization Southern Nuclear Design Calculation Plant: Votl Unit: 1&2 Calculation Number: X6CNA15 Sheet: C-I ATTACHMENT C -REFERENCES DescrptionNumber Descrptionof Pages C1 -Davisson, "Gamma Ray Dose Albedos," from "A Handbook of Radiation 13 Shielding Data" 02 -Validation of Spirax Sarco On-Line Steam Tables 10 03 -Collins-Bornt e-mail, "Re: Vogtle EAL Setpoints," 27AUG14 2 04 -Collins-Bornt e-mail, "Re: Turbine Building Vent Release Permit -1/2RE- I 128390," 28AUG14 C5 -Stan ley-Bornt e-mail, "RE: Vogtle Containment Penetrations

-Cold Shutdown 2& Refueling Modes," 04SEP14 C6 -Cowman-Bornt e-mail, "RE: Vogtle Containment Penetrations

-Cold Shutdown 2& Refueling Modes," 05SEP14 1-4-1-4-1-4-Total Number of Pages Including Cover Sheet 3 31 Edwin I. Hatch Nuclear Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 3 Hatch EAL Calculations COMPANYJM Southern Nuclear Design Calculation

[Calculation Number: SMNH-;13-021 Plant:Hatch Nuclear Plant Unit:O-I 0 [ 2 E[] I & 2 fDiscipline:Mechanical Title: I

Subject:

Emergency NEi 99-01 Rev 6,EAL Calculations

.,.Action Level Setpoints"PurpoSe I Objective::

Document Emergency Action Level Values to support conversion to NEI 99-01 Rev 6 System or Equipment"Tag Number..:

'N/A .. .i..Contents Topic Page Attachments*

1. of (Computer Printouts, Technical Papers, Pages Sketches,, Correspondence)

Purpose 1 A -SNC EP Concurrence " ..1 Criteria 1 B -Reserved 0 conclusions 3 C -FuelClad Loss 4 .A Threshold calculation 19 Design Inputs ' 5 D -Fuel Clad Potential Loss 5.A Threshold 1.Assumptions*

10 11___References 14 !E -RCS Loss 4.A Threshold Calculation 16 Method of Solutions

...18i F -'Rc's Loss 5.A Thresh'old Calculation " .6 Body of Calculation

... ;: " 123 IG -Containment Potential Loss ,4A"Threshold 12 Calculation

"____ ._________

___H -Steam Table Validation

-... ' 'I- RSI Threshold Calculation 25 Total # of Pages including J8 -K Threshold Calculatio~n 4'1 cover sheet & Attachments:

18 K-G hrsodaclain1 L- E-HU.! Threshold Calculation

... 16 M -RA1 Threshold Calculation

.18 Nuclear Quality Level.0] Safety-Related IX] Safety Significant

.. I"1 Non- Safety --Significant Version Record _____.... ..nator .. Revioewer

'Approv'al I "

2 No. Description -__________

J htjSlrx D.S. 0. W. Wu A.T. Vieirs D.L. Lamlbert 1 Issued McCutcheon Sign. on file Sign. on file Sign. on file 2 Complete Revision 9I ooonakisn

'7 a oile KeithoDudy Notes: .. .1. Additional work and changes to this calculation are required.

The calculation is "APPROVED PENDING" NRC review and acceptance of the Emergency Plan (EP) submittal.

NMP-ES-039 NMP-ES-039-001 Purpose: The purpose of this calculation is to provide values/data/curves and bases for use in development of the Hatch Nuclear Plant Emergency Action Levels (EALs) using Nuclear Energy Institute (NEI) 99-01 Rev 6 guidelines.

This combined calculation includes all unique calculations required to support emergency action level thresholds as well as references to calculations used to create thresholds, but serve purposes beyond emergency action levels.The contents of this calculation are primarily taken from the calculation which supported the previous emergency action level scheme. The work performed in calculation SMNH-05-009 V2 is directly transposed into this document and edited to account for the differences between NEI 99-01 Rev 4and 6.The calculation Attachments L and M of this document contain calculations supporting new emergency action level thresholds and represent the only portion of this document which is not directly transposed from SMNH-05-009.

The unique work in this document includes those attachments and their associated content in the body of the calculation.

Several EAL thresholds were removed as they do not support the NEI 99-01 Rev 6 scheme. Thresholds removed include specific water elevations which previously supported l~s RU2, RA1, RA2, BFL2, and RCS Barrier. The transposed material in this calculation has been further altered to reflect the new language and organization of NEI 99-01 Rev 6. These changes in language and organization are administrative in nature and have no impact on the calculation output.Criteria: The calculation performed will support the development of guidelines for NEI 99-01 initiating conditions (ICs) RU1, RU2, RA1, RA2, RS1, RGI, CAl, CS1, CG1, E-HU1, Fuel Clad Barrier, RCS Barrier, Containment Barrier, and SU3.1. Declaration of an emergency, when such a declaration is not required, involves risk to the public as does the failure to make such a declaration, should one be warranted.

Therefore, this calculation shall develop a "best estimate" value for the dose rates or curie concentrations sensed at the monitors chosen for the Emergency Action Level (EAL) set points. When judgments are necessary, these judgments shall be as close to anticipated conditions as possible.2. If a particular monitor is to be used for an EAL, then the dose rate or curie concentration set point developed for that specific monitor shall be within the range of the monitor, or the monitor shall not be cited as applicable for the EAL.3. In accordance with the guidance of Regulatory Guide 1 .97, Revision 2, post-accident radiation monitors must read within a factor of 2 of actual radiation conditions.

Therefore, changes in the set points of this revision that are within a factor of 2 of the previous revision's set point for the same EAL do not invalidate the previous set point.It is up to the ultimate user of these calculations to determine if change to the EAL set point guidance document(s) is warranted.

4. Methods and Assumptions shall comply with the guidance of NEI 99-01 Revision 6.Note: NEI 99-01 Rev. 6 states that the "A" Recognition Category designation may be changed to "R" provided the change is carried through for all of the associated IC identifiers.

As such, the Hatch Nuclear Plant Emergency Action Levels use the Recognition Category designation of "R" for the Abnormal Radiation Recognition Category.

==

Conclusions:==

This calculation contains thresholds for the Hatch Nuclear Plant Emergency Action Levels. Site specific information used to develop the scheme is included in the Body of Calculation section beginning on page 23. In this document, ten thresholds are calculated.

The results of these calculations are as follows.Initiating Condition RA1 Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6E-02 pCi/cc 8.1E+01 pCi/cc Initiating Condition RS1 Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6E-01 pCi/cc 8.1E+02 pCi/cc Initiating Condition RGI Greater than any of the following monitor readings for 15 minutes or longer serves as the threshold for EAL 1.Reactor Building Vent Main Stack 2.6 E+OO pCi/cc 8.1E+03 pCi/cc Initiating Condition CG1 The supporting calculation for this threshold concludes that there are no monitors able to provide on-scale indications of core uncovery.Initiating Condition E-HU1 Greater than any of the following on-contact radiation readings serve as thresholds for EAL 1.

Southern Nuclear 0perating Cornpany SOU1HURN Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-04 COMPANY Unit: l&2SH T4 Table El Location of Dose Rate Total Dose Rate (Neutron + Gamma mR/hr)HI-TRAC 125 Side -Mid- height 450 Top 110 HI-STAR 100 or HI-STORM 100 Side -60 inches below mid- 80 height Side -Mid- height 80 Side -60 inches above mid- 30 height Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 Fuel Clad Barrier Loss Threshold 4.A A Drywell Wide Range Radiation Monitor (DWRRM) reading greater than 1,400 R/hr.Fuel Clad Barrier Loss Threshold 5.A The supporting calculation for this threshold concludes that the applicable monitors will be off scale.RCS Barrier Loss Threshold 4.A A DWRRM reading greater than 40 R/hr.RCS Barrier Loss Threshold 5.A The reading on drywell fission product monitor D11K630 of 1E+06 cpm will indicate a potential loss of the RCS barrier. Per SX18062 page 34, the monitor K630 range is 10 to 1E+06 cpm. A reading of 5.OE+05 cpm is an appropriate threshold to ensure an accurate reading is possible.Primary Containment Barrier Potential Loss Threshold 4.A A DWRRM reading greater than 26,000 R/hr.

Southern Nuclear 0perating Cornpany ISOUTHERNEI Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY' Unit: 1&2 SHEET 5 Design Inputs (additional design inputs and references are listed in the attachments as needed): Attachment C: 1. The core inventory source terms are listed in Enclosure 1 of NL-06-1637 page 16.Core Inventory Isoope Curies/MWt Kr-83m 3.30E+03 Kr-85 3.78E+02 Kr-85m 6.92E+03 Kr-87 1.32E+04 Kr-88 1.86E+04 Kr-89 2.26E+04 Xe-131m 3.03E+02 Xe-133 5.27E+04 Xe-133m 1.58E+03 Xe-135 1.89E+04 Xe-135m 1.09E+04 Xe-137 4.81E+04 Xe-138 4.52E+04 1-131 2.72E+i04 1-132 3.93E+'04 1-133 5.52E+04 1-134 6.05E+04 I- 135 5.16E+04 Per NUREG-1301 pg. 6, DE1-131 only considers the following iodine isotopes (1-131, 1-132, 1-133, 1-134, 1-135). Therefore other iodine isotopes have been excluded from this calculation

2. Dose Equivalent 1-131 is defined as that concentration of 1-131 (pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, 1-135 (NUREG-1 301 page 6).3. Core Power for radiological evaluations of 2818 MWt is provided in NL-06-1 637 page 14 Enclosure 1.4. RCS inventory of 9965 ft^3 is provided in NL-06-1 637 Enclosure 1 page 47 Table 34.5. The release fraction into the containment for noble gases and halogens (i.e. lodines) is 0.05 per RG-1.183 page 13.6. The dry well free volume (UI: 146010 ftA3 U2: 146266 ftA3) is listed in NL-06-1637 Enclosure 1 page 47 table 34.7. The suppression pool free volume (UI: 112900 ftA3 U2: 109800 ftA3) is listed in NL-06-1637 Enclosure 1 page 47 table 34.

8. UI: The Main feed water temperature is 392.4 °F and core recirculation temperature is 532.0 0 F. Per GE-NE-0000-0003-0634-01 pg. 15.9. U2: The Main feed water temperature is 425.7 °F and the core recirculation temperature is 535.0 0 F. Per GE-NE-0000-0003-0634-01 pg. 16.10. According to PDMS the Drywell Wide Range detector is MPL 1 D11 N003A&B and 2D11IN003A&B.

The detectors are connected to D11 K621 Recorders as described in the following table: MPLDraingInstrument Tag Range'MLSheet No R/hr Model No 1D1IN003A A16481 DuIB QD11-RE-N003A 1EO- 1E7 877-1 1D11N003B A16481 D11C QD11-RE-N003B 1E0 -1E7 877-1 2D11N003A A26481 D11A Q2D11-RE-NO03A lE0- 1E7 877-1 2D11NOO3B A26481 D11B O.2Dl1-RE-NOO3B lEO -1E7 877-1 Attachment D I11. Attachment D Factor Allocation 0.225 and Factor Safety 0.5 per 64CI-OCB-004-1/2 Page 56.12.0Off-Gas Post Treatment Radiation Monitor D11 K61 5 efficiency Factor 3.8E5 cps/pCi/cc per e-mail attached to attachment D 13. The High Purity Waste Stream leakage inside the Drywell is obtained from SMNH-93-029 page 6 and is 3900 gal/day.14. The Low Purity Waste Stream leakage inside the Drywell is obtained from SMNH-93-029 page 6 and is 700 gal/day.15. Fission Product Monitor D11K630 for Xe133 = 2.7E7 cpm/IpCi/mL and Kr85 = 2.29E8 cpm/pCi/mL.

The response factor was obtained from SX1 8062 pg. 32 and SX27520 pg.32 for equipment tag number P010.Attachment E 16. The radiation field from RPV at elevation 153.6 for Unit 1 and Unit 2 is listed below (BH2-M-V999-0047 Table 2 and 3).

Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-FO13A 6.44 2B21-AOV-FO13A 2.46 1B21-AOV-FO13B 4.41 2B21-AOV-F013B 9.96 1B21-AOV-FO13C 12.3 2B21-AOV-FO13C 9.96 1B21-AOV-F013D 13 2B21-AOV-F013D 2.46 1B21-AOV-FO13E 12.3 2B21-AOV-F013E 2.46 1B21-AOV-F013F 12.3 2B21-AOV-F013F 9.96 1B21-AOV-F013G 12.3 2B21-AOV-FO13G 9.96 1B21-AOV-FO13H 4.41 2B21-AOV-FO13H 2.46 1B21-AOV-F013i 4.41 2B21-AOV-FO13K 9.96 1B21-AOV-FO13K 12.3 2B21-AOV-FO13L 9.96 1B21-AOV-FO13L 12.3 2B21-AOV-F013M 9.96 Average = 9.7 7.2 17. The radiation field from Main Steam Lines (BH2-M-V999-0047 Table 5 and 6).Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-FO13A 17.57 2B21-AOV-FO13A 17.39 1B21-AOV-FO13B 17.28 2B21-AOV-FO13B 17.28 1B21-AOV-F013C 17.08 2B21-AOV-FO13C 17.08 1B21-AOV-F013D 17.01 2B21-AOV-F013D 17.39 1B21-AOV-F013E 17.16 2B21-AOV-F013E 17.28 1B21-AOV-F013F 16.13 2B21-AOV-FO13F 17.08 1B21-AOV-F013G 16.13 2B21-AOV-F013G 15.43 1B21-AOV-F013H 16.62 2B21-AOV-F013H 17.28 1B21-AOV-F013J 16.33 2B21-AOV-FO13K 15.43 1B21-AOV-F013K 16 2B21-AOV-F013L 14.78 1B21-AOV-F013L 13.4 2B21-AOV-F013M 14.78 Average = 16.4 16.5 18. The radiation field from Recirculation Lines (BH2-M-V999-0047 Table 9 and 10).'S Solenoid Valves Solenoid Valves for Ul Rad/hr for U2 Rad/hr 1B21-AOV-F013A 7.5 2B21-AOV-F013A 11.97 1B21-AOV-F013B 8.23 2B21-AOV-F013B 6.23 1B21-AOV-F013C 6.19 2B21-AOV-F013C 6.85 1B21-AOV-F013D 5.2 2B21-AOV-F013D 10.08 1B21-AOV-F013E 5.89 2B21-AOV-F013E 15.22 1B21-AOV-F013F 5.15 2B21-AOV-F013F 6.19 1B21-AOV-F013G 5.89 2B21-AOV-F013G 6.85 1B21-AOV-FO13H 7.5 2B21-AOV-F013H 15.96 1B21-AOV-F013J 7.5 2B21-AOV-F013K 6.85 1B21-AOV-F013K 6.22 2B21-AOV-F013L 7.8 1B21-AOV-FO13L 6.22 2B21-AOV-F013M 7.8 Average = 6.5 _______ 9.3 Attachment F 19.The Normal RCS Isotopes concentration is taken from FSAR Table 11 .1-1, and are listed in the attachment F.20. The core flow (UI: 78.5E6 Ibm/hr U2: 77E6 lbm/hr) can be found in GE-NE-0000-0003-0634-01 Figures Ia & lb.21. Per SX1 8062 page 34 the monitor K630 range is 10 to 1 0A6 cpm.Attachment G3 22. DWRRM is D11K621 NMP-EP-110-GL02 pg. 66.23. The source terms for noble gases and iodines are provided in NL-06-1 637 and are shown below. We will only consider the isotopes that have release fraction above 0 as described in RG 1.183 Table I and Table 5.24. The 20% is the core fraction gap release as described in NEI 99-01.Attachment H 25. Radiation from a HOLTEC Overpack from table I and 2 of S55932.26. The attenuation coefficient for 6 MeV gamma was used. The lowest attenuation was used thus giving us most conservative (lowest) shielding factor. Attenuation

= 0.234 cm^-l.27. MPC lid thickness 9.5 inches per $55932 pg. 7 bullet 9.Attachment I

28. The Core Release Fractions are taken from RG 1.183 Table 1. This fraction, describes release of isotopes to RCS water. Iodines 0.3 and Noble Gases 1.0.29. The Partition Coefficient is the ratio of the concentration of a nuclide in the gas phase to the concentration of that nuclide in the liquid phase when the liquid and gas are at equilibrium.

It is assumed that 100% Noble gases are released into the steam.According to NUREG-0016 Table 2-7 page 2-13 the Iodines partition coefficient is 0.004.30. X_Q = 8.37E-6 sec/m^3 ground release per Table 3-4 of HNP ODCM page 3-17 31. Reactor building ventilation flow rate 1 .42E8 mL/sec per Table 3-4 of HNP ODCM page 3-17 32. The Dose Conversion Factor (DCF) for Effective Dose Equivalent (EDE) was taken from FGR12 "Effective Column" of Table II1.1.33. The Dose Conversion Factor (DCF) for Committed Effective Dose Equivalent (CEDE)was obtained from FGR11 Table 2.1 column labeled "Effective".

34.The exposure time is provided in NEI 99-01, T~lhr.35.The breathing rate for persons offsite is listed in section 4.1.3 page 1.183-16 of RG-1.183. BR=3.5E-4 mA3/sec 36. Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE). However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent for external exposure, if the body is irradiated uniformly.

37. The Thyroid Committed Dose Equivalent (CDE) from inhalation was obtained from FGR 11 Table 2.1 column labeled "Thyroid".

38. X_Q = 4.10E-8sec/m^3 elevated release per Table 3-4 of H NP ODCM page 3-17.39. Main stack flow 9.44E6 mL/sec per Table 3-4 of HNP ODCM page 3-17.40. Recombiner building vent flow 2.36E5mL/sec per Table 3-4 of HNP ODCM page 3-17.Attachment J 41. Radiation field above SFP when water is at TOAF is 2.68E5 R/Hr per BH2-M-V999-0048 pdf page 39 column 3.42. The material of RPV is steel per FSAR U2 Table 5.2-6.43.The RPV inside diameter is 218 inches per FSAR U2 Table 5.4-1 and $15213.44. RPV vessel thickness is 5.38 inches per FSAR U2 Table 5.2-10.Attachment K 45. Roof elevation of 281 ft 9 in DWG H25963.

Assumptions (additional assumptions are listed in attachments as necessary):

Attachment C 1. The higher temperature of RCS will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-Q000-0003-0634-01 page 19 the reactor pressure is 1060 psia. According to http://www .spiraxsarco.com/resources/steam-tables .asp, the specific volume for 1060 psia and 480.35 °F conditions is 0.01 991 61 ft^3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to attachment C.2. According to http://www.spiraxsarco.com/resources/steam-tables.asp, the specific volume for 44.7 psia and 343 0 F conditions is 10.4690 ftA3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to Attachment C. The drywell pressure, and temperature were obtained from NL-06-1 637 Enclosure 1 table 11 and table 12.3. The Drywell Wide Range detectors are 1 D11N003A&B and 2D11IN003A&B.

Per H16241 and H26417 the instrument location is at elevation of 156 ft and approximately 27 ft from the center of RPV. As the result the geometry of detector in this volume will be modeled as a cylinder with of 41.5 feet high. This number corresponds to the elevation above drywell floor. The radius of the cylinder was adjusted to the drywell free volume that is in the line of site of the detector.

The geometry factor will be evaluated for the point P1. The uncollided fluxes at interior and exterior points of a non-absorbing cylindrical volume source are provided in Engineering Compendium on radiation shielding pages 381 and 382.T. .........

-" Fig'. 6.4.-?. Geometry3 for non-absorbing cylindrical, volume source; interior, surface and exterior points.At P 1 : 4. The mean free path was calculated for the containment air (Figure 1). It can be seen that the mean free path is over 650 feet, which is much larger than the reactor building of 71.75 ft+71 .75 ft = 143.5 ft (H25941).

Therefore it was concluded that the Southern Nuclear 0peratinl Cornpany OUHr 1 Plant: HNP SMNH-1 3-021 nt & Title: NEI 99-01 Rev 6 EAL Calculations SET1 containment atmosphere does not provide any appreciable gamma shielding and that the non-absorbing cylindrical model should be used.Figure 1: Mean Free Path for Drywell Steam..... .pa/_!=1.

Mass Absorption Coefficient (cm^2/g) --..... =Absorption Coefficient_(1l/cm_)

... .....;L= iMean Free Path (cm) !........... I ...' ........L -.....I .... .....Steam @ 44.7 psia & 343degF L---... .... ...... .....-- -...... ..... ... ... ....... ... ...- -... .-!.... L Calculated in AttachmentC p.= 11.53E-03 IEnergy J.aPLia (MeV) g)^2 (11cm) (cm) (ft)0.5 0.0330 5.05E-05 1.98E+04 6.50E+02 1.0 0.0311 4.76E-05 2.10E+04 6,89E+02 1.5 0.0285 4.36E-05 2.29E+04 7.52E+02 2.0 0.0264 4.04E-05 2.48E+04 8,12E+02 3.0 0.0233 3.56E-05 2.81E+04 9.20E+02 4.0 0.0213 3.26E-05 3.07E+04 1.01E+03 glcm^~3 to--.. .... ... .. .......- ..... ... ... .. .... ..... ..... .. ... .... .. .. ..... ......$

Reference:

Table 11.5, page 649, Lamarsh, "Introduction Nuclear Engineering," 2nd edition, 1983 -, I Attachment D 5. Since the D11 P010 gas detector is G-M counter ($30523 pdf page 76 section 1-4) the Xe1 33 response factor will applied to all isotopes of Xenon, while the Kr85 response factor will be applied to all isotopes of Krypton. For G-M detectors all gammas above detector energy threshold would initiate charged particles inside the detector.Attachment F 6. Noble gases are not typically retained in the RCS water, but are continuously released via offgas system. The source terms for noble gases are taken from Table 11.1-1. It is assumed that 100% of noble gases will leave the solution.Attachment I

Southern Nuclear 0 eratin9 Cornpany SOUTHERN Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET13-2 COMPANY Unit: 1&2 SET1 7. The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used in HNP FSAR Chapter 15 evaluations:

Iodine's and noble gases Table 15.3-4. It is assumed that only iodine's and noble gases needs to be considered, because particulates will be retained in the primary containment water.8. The RCS will have some equilibrium noble gases and some iodine in the RCS water from normally operating reactor. According to HNP FSAR U2 Table 11.1-2 the levels of iodines is on the order of 1 E-1 pCi/g which is negligible to calculated above order of magnitude of 1 E5 pCi/g. The reason these isotopes are in low concentrations because they are continually removed in the power plant steam condensers.

Thus initial equilibrium noble gases and iodines will be neglected.

Attachment J 9. The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top of Active Fuel. The calculation pdf page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Since the RPV core is assumed to be a columnized source, the same radiation level is assumed to be at the edge of the RPV. This is conservative and according to the BH2-M-V999-0048 pdf page 39 column 2 on the edge of the SFP the radiation level is 2.29E5 R/hr. As can be seen the radiation level does not very much from the middle of the core to the edge of the core.10. The page B-9 of BH2-M-V999-0048 provides the gamma source strength broken down by energy groups. It can be seen that the maximum gamma strength is spread between 0.4 to 1.8 MeV, therefore this calculation will use linear attenuations that are associated with 1 MeV gamma rays.Attachment K 11. The reflected dose rate at the operating deck area radiation monitors will be calculated using the methods of Davission's "Gamma Ray Dose Albedos" (copy in Attachment K).The calculation will be based on an iron reflector at the top of the secondary containment, with a diameter equal to the drywell, and a distance R from the reflector to the radiation monitor equal to the hypotenuse of the triangle formed by the difference in elevations of the reflector and the monitor and the distance from drywell center to the approximate detector placement.

The iron reflector is selected because the containment roof has metal roof deck. The reflected dose rate is proportional to the area of the reflector.

Assuming the reactor vessel functions as a collimator with reduced RCS inventory will reduce the reflected area. This in turn reduces the dose rate at the radiation monitor and, therefore, the EAL threshold for reduced RCS inventory.

f 4 Southern Nuclear 0 eratinl Cornpany SOTEN4 Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-1 3-021 COMPANY Unit: 1&2 SHEET 14

References:

1. NEI 99-01 "Methodology for Development of Emergency Action Levels" Rev 6.2. H16145 V18.0 "HNP UI Nuclear Boiler System P&ID Sheet 3" 3. H261 89 V21 .0 "HNP U2 Nuclear Boiler System P&ID Sheet 3" 4. NUREG-1301 "Offsite Dose Calculation Manual Guidance:

Standard Radiological Effluent Controls for Pressurized Water Reactors." April 1991.5. NL-06-1637 "Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term." August 29, 2006.6. GE-NE-0000-0003-0634-01 "Edwin I. Hatch Nuclear Plant, Units I and 2, 10-PSI Dome Pressure Increase." July 2003.7. S16039, V2.0 "Instrument Data Sheet Process Radiation Monitoring System" 8. A16481 Sheet D11IA V2.0 "STACK GAS VELOCITY PROBE AND TRANSMITTER DATA SHEET" 9. A16481 Sheet D11B V0.0 "AREA RADIATION MONITORING SYSTEM" 10.A16481 Sheet D11C V0.0 "AREA RADIATION MONITORING SYSTEM" 11 .A26481 Sheet D11A V0.0 "AREA RADIATION MONITORING SYSTEM" 12.A26481 Sheet D11B V0.0 "AREA RADIATION MONITORING SYSTEM" 13. H16566 V7.0 "PROCESS RADIATION MONITORING SYSTEM P&ID (SHEET 4)" 14. H26017 V22.0 "FISSION PRODUCTS/POST-LOCA MONITORING SYSTEMS P&ID" 15. H16032 V5.0 "EQUIP LOCATION REAC BLDG SECT A-A" 16. H16241 V29.0 "INSTRUMENT

& PRIMARY PT LOC REAC&RADWASTE BLDG EL 130" 17. H26417 V17.0 "INSTRUMENT

& PRIMARY POINT LOCATIONS-REACTOR BUILDING-EL.

130'-0" & RADWASTE BUILDING-EL.

132'-4"" 18. Engineering Compendium of Radiation Shielding, NY 1968, Volume 1.19.64CI-OCB-004-1 V7.0 "UNIT ONE POST TREATMENT RADIATION MONITORS" 20. 64CI-OCB-004-2 V7.0 "UNIT TWO POST TREATMENT RADIATION MONITORS" 21.S $6039 V2.0 "INSTRUMENT DATA SHEET PROCESS RADIATION MONITORING SYSTEM" 22. SX27520 VI.0 "IDS -PROCESS RADIATION MONITORING SYSTEM" 23. SX1 8062 V2.0 "IDS PROCESS RADIATION MONITORING SYSTEM" 24. S30523 VO.2 "IODINE -NOBLE GAS SAMPLE PANEL 133D9025G1-G6 INSTRUCTION MANUAL (OPERATION AND MAINTENANCE)" 25.S$19207 V2.0" INSTRUCTION MANUAL RADIATION MONITORING SYSTEM VOLUME 4" 26. SX29455 V1 .0 "RADIATION MONITORING SYSTEM VOLUME IV"

27. H 16568 V5.0 "REACTOR PROTECTION SYSTEM P&ID" 28. BH2-M-V999-0047 V2.0 "DRYWELL EQUIPMENT EQ DOSES FOR EXTENDED POWER UPRATE FOR REA HT-96660" 29. HNP Technical Specifications 273/218 01 16 30. NUREG-0016 "Calculation of releases of radioactive materials in gaseous an liquid effluents from boiling water reactors (BWR GALE Code)" April 1976.31. RG 1.183 "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." July 2000 32. BH2-CS-52-2P33-01 V3.0 "Containment Hydrogen Analyzer" 33. BH2-CS-52-2P33-02 V2.0 "Containment Oxygen Analyzer" 34. Regulatory Guide 1 .7 "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident" Rev 1, Sept 1976.35. SS2102005 V6.0 "Furnishing

& Delivery of Reactor Drywell & Suppression Chamber-Containment Systems" 36. CALC F-86-03 V0.0 "COMPUTER CODE: VERIFICATION OF THE F GRODEC COMPUTER PROGRAM" 37. Deleted 38. 64CI-OCB-008-0 V8.1 "PLANT SERVICE WATER RADIATION MONITORS" 39. 64CI-OCB-009-0 V5.3 "LIQUID RADWASTE RADIATION MONITORING" 40. H16564 V29.0 "PROCESS RADIATION MONITORING SYSTEM P&ID SHT. 2" 41. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2" 42.64CI-OCB-002-1 V12.0 "UNIT ONE REACTOR BUILDING VENT RADIATION MONITORING" 43. 64CI-OCB-002-2 V16.0" UNIT TWO REACTOR BUILDING VENT RADIATION MONITORING" 44. H26013 V7.0 "PROCESS RADIATION MONITORING SYSTEM l.E.D. SHEET 3" 45.64CI-OCB-003-1 V14.0 "RECOMBINER BUILDING VENT RADIATION MONITORING" 46. H16528 V12.0 "OFF GAS RECOMBINER BUILDING VENTILATION SYSTEM P & ID AND PROCESS FLOW DIAGRAM" 47. 64CI-OCB-001

-0 Vi13.0 "MAIN STACK RADIATION MONITORING" 48.5$56256 V1.0 "GEl4 FUEL BUNDLE INTERFACE CONTROL" 49.$S54974 V0.3 "BWR SPENT FUEL STORAGE RACKS -RACK LAYOUT MPL. F16" 50. S54975 V0.1 "BWR SPENT FUEL STORAGE RACKS -CONSTRUCTION (ELEVATIONS)

MPL. F16" 51 .H 15602 V1 .0 "REAC BLDG FUEL TRANS POUR NL PLAN SECT&DET"

52. H15336 V10.0 "REAC BLDG SPENT FUEL POOL-PLAN

@ EL 228&EL SECT&DET POUR NL" 53. Deleted 54. H25963 Vl0.0 "ARCHITECTURAL

-EXTERIOR WALL AND ROOF DETAILS" 55. H25621 V2.0 "REACTOR BUILDING-STRUCTURAL STEEL-ROOF FRAMING AT EL.280'- 0" -PURLINS AND TOPCHORD" 56. H25694 V0.0 "REACTOR BUILDING-STRUCTURAL STEEL 280'-0"- VERTICAL SECTIONS" 57. H25953 V3.0 "ARCHITECTURAL-REACTOR BUILDING-SECTION A-A" 58. H15880 V4.0 "ACCESS CONT FLOOR PLAN ELEVATION 203 228&244-1 0" 59. H45515 V1.0 "I/O LIST SPDS RTP NODE 2 (SHEET 1)" 60. H16562 V7.0 "AREA RADIATION MONITORING SYSTEM I.E.D" 61. H26010 V10.0 "AREA RADIATION MONITORING SYSTEM I.E.D" 62. H27805 V22.0 "AREA RADIATION MONITORING SYS. 2D21 ELEMENTARY DIAGRAM SHT. I OF 8" 63. H25990 V3.0 "ACCESS CONTROL -FLOOR PLAN -EL. 203' AND 228"'64. S55932 V1.0 "DOSE RATES FROM THE HI-STORM 100 SYSTEM F(OR THE HATCH ISFSI MPL F18" 65. HNP Dry Storage FSAR "Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System" Revision 7, http://nuclear.southernco.com/regqulatory-affairs/H NP-Dry-Cask.html

66. H15878 V14.0 "ACCESS CONT FLOOR PLAN ELEVATION 158" 67. HNP ODCM V23.0 "Offsite Dose Calculation Manual for Hatch Nuclear Plant" 68. FGR12 Sep 1993 "External Exposure to Radionuclides in Air, Water, and Soil" 69. FGR1 1 Sep 1988 "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." 70.5SX18761 V2.0 "P & I DIAGRAM KMG -HRH" 71.S$41454 V2.0 "P & I DIAGRAM KMG-HRH" 72. DoclD: RE203727981 VI .0 "GASEOUS EFFLUENT REPORTS" for procedure 64CH-R PT-007-0.73. DoclD: RE203186522 VI.0 "Form HPX-0893 to procedure 64CI-OCB-003-1" 74. S25213 Rev G "NUC BOILER SYS-DESIGN SPEC" 75. H16565 V5.0 "PROCESS RADIATION MONITORING SYSTEM P. & I.D. SHT. 3" 76. H26012 V23.0 "PROCESS RADIATION MONITORING SYSTEM I.E.D. SHEET 2" 77. H16014 V39.0 "REACTOR BUILDING REFUELING FLOOR VENTILATION SYSTEM P. & I. D."

Southern Nuclear 0perating Cornpany Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMvNH-13-021 COMPANY Unit: 1&2 SHEET 17 78. H27158 V6.0 "PROCESS RADIATION MONITORING SYSTEM 2D11 ELEMENTARY DIAGRAM SHEET 16 OF 22" 79. Lamarsh, "Introduction to Nuclear Engineering," 2nd edition, 1983 80. H25942, V14.O "ARCH REAC BLDG-FL PLN-EL 130" 81.ASME International Steam Tables for Industrial use. 2 nd edition.82.64CI-OCB-005-1 V15.0 "UNIT ONE FISSION PRODUCT MONITOR" 83.,64CI-OCB-005-2 V15.0 "UNIT ONE FISSION PRODUCT MONITOR" 84. SS6902005 V3.0 "REACTOR DRYWELL & SUPPRESSION CHAMBER VESSELS &APPURTENANCE" 85. BH1-CS-33-P33-06 VI1.0 "CONTAINMENT 02 ANALYZER" 86. SMNH-93-029 VO.0 "LIQUID EFFLUENT DOSES NORMAL OPERATION" 87. Shultis and Faw "Fundamentals of Science and Engineering" 2002 88. BH2-M-V999-0048 V0,0 "SPENT FUEL POOL BOILING DOSE RATE" 89. Gieck "Engineering Formulas" 7 th edition.90. SI15213 V0.2 "RPV VESSEL ASSEMBLY-PF-l1983-63-8" 91. Courtney "A Handbook of Radiation Shielding Data" July 1976 92. 64CI-OCB-006-1 VI14.0 "UNIT ONE PRETREATMENT RADIATION MONITORING" 93. 64CI-OCB-006-2 V12.0 "UNIT TWO PRETREATMENT RADIATION MONITORING" 94.31E0-EOP-014-1 V12.0 "SC -SECONDARY CONTAINMENT CONTROL RR -RADIOACTIVITY RELEASE CONTROL" 95. 31 EO-EOP-014-2 V1 1.0 "SC -SECONDARY CONTAINMENT CONTROL RR -RADIOACTIVITY RELEASE CONTROL" 96. SMNH-05-009 V2.0 "NEI 99-01 EAL Calculations" 97. SNC024-CALC-007 Rev. 0 "HNP Determination of Emergency Action Level for Initiating Condition E-HUI" (Attachment L)98. SNC024-CALC-008 Rev. 0 "Hatch EALs RA1 Threshold to Address NEI 99-01 Revision 6" (Attachment M)

Method of Solutions:

NEI 99-01 Revision 6 Methods conform to the guidance of NEI 99-01 Revision 6. Detailed descriptions of the methods are included in the individual EAL threshold calculations in the Analysis section of this calculation.

Use of Regulatory Guide 1.183, Alternate Source Term Method The NEt 99-01 Revision 6 Recognition Category A (Abnormal Rad Levels/Radiological Effluent)Initiating Conditions (ICs) for declaring an Alert, a Site Area Emergency, and a General Emergency (Emergency Action Levels RA1, RS1 and RG1, respectively) are expressed in terms of Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (CDE).Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," is not the current license basis for performing dose calculations for Hatch. However, it expresses doses in terms of Whole Body and Thyroid.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," does express doses in terms of TEDE and CDE, is the current licensing basis for performing dose calculations for Hatch. However, per section 1.1.4 on page 1.183-6, "This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with emergency dose projections, protective measures, and severe accident management guides." Per section 4.1.1 of RG 1.183, TEDE is defined as the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure.Per section 4.1.2 of RG 1.183, Table 2.1 of Federal Guidance Report 11 provides tables of conversion factors acceptable to the NRC staff. The dose conversion factors (DCFs) factors in the column headed "effective" yield doses corresponding to the CEDE.Per sections 4.1.4 and 4.1.5 of Reg Guide 1.183, the DDE should be calculated assuming submergence in a semi-infinite cloud for the most limiting person at the EAB. The effective dose equivalent (EDE) from external exposure is nominally equivalent to the DDE, thus EDE may be used in lieu of DDE in determining the external dose contribution to the TEDE. Table II1.1 of Federal Guidance Report 12 provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

Southern Nuclear 0peratinl Cornpany ISOUTH=ERNA.

Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMH1-2 COMPANY Unit: 1&2 SET1 TEDE & Thyroid ODE Calculations The basic process for calculating an offsite dose consists of first determining the concentrations of radionuclides in the release stream, be it air, steam, or water. The release stream concentration is determined by dividing the release rates of the radionuclides of interest, expressed as microcuries (pCi) per unit time, by the release fluid's volumetric flow rate, expressed as cubic centimeters (cc) per unit time: pJCi/cc = [pCi/unit time]/[cc/unit time]As we are back-calculating release concentrations based on pre-established dose limits (100 mREM TEDE and 500 mREM Thyroid ODE), the upstream modeling of the specific release paths is not necessary.

The gaseous effluent noble gas radiation monitors care not a whit how those radionuclides arrive at them.Step 1: Identify the radionuclides of interest.

Select the same radionuclides used to calculate doses for the design basis accidents in FSAR chapter 15: the fission product noble gases and iodines. The other fission products and activated corrosion products are particulates and will not contribute significantly to the offsite dose.Step 2: Determine the RCS coolant radionuclide activity for each radionuclide (Xrcs-i pCi/g). This is assumed to be the sum of core fission product inventory released during a LOCA divided by RCS coolant mass (Mrcs g) and the equilibrium RCS coolant activity (Xeq-i) for that radion ucl ide.Xrcs-i = Xeq-i + (1 .0E+06 pCi/1 Ci) x [Core Inventory (Ci)] x [Release Fraction]/(Mrcs g)For no fuel damage, the release fraction is 0 and the RCS activity is the equilibrium RCS coolant activity.

If fuel damage is assumed (release fraction > 0), the quotient of core inventory and RCS coolant mass will be orders of magnitude greater than the contribution from the coolant equilibrium activity.Step 3: Convert coolant activity (Xrcs-i pCi/g) to release stream activity (Xris-i pCi/cc). This conversion is accomplished by multiplying the RCS coolant activity by a dimensionless partition factor (PF 1) and an arbitrarily selected density, pris g/cc: Xrls-i (pCi/cc) = (Xrcs-i pCi/g) x PFi x (pris g/cc)The partition factor will depend on the radionuclide and the release path. The partition factors used in this calculation are discussed in Assumption

1. 4 of this calculation.

Arbitrarily set pris = 1.0 g/cc to make the math easy. The justification for this will be provided in Step 9.Step 4: Determine radionuclide concentration at Exclusion Area Boundary (XEAB-i pCi/cc). This is done using standard dose assessment methods. The release concentration is multiplied by the release volumetric flow rate (Q~s m 3/sec) and the diffusion coefficient

[(X/Q) m 3/sec]: XEAB-I (pCi/cc) = Xris-1 (pCi/cc) x [Qris (m 3/sec)] x [(X/Q) (m 3/sec)]Step 5: Calculate the TEDE for each radionuclide for one hour exposure time at EAB. This is done using the appropriate FGR-11 and FGR-12 dose conversion factors (DCFs), as discussed in the previous subsection.

Southern Nuclear 0peratinl Cornpany SOUTHERNZ[

lnt N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET3202 COMPANY Unit: 1&2 SET2 TEDEi (mREM) = External Exposure + Internal Exposure TEDE 1 (mREM) = XEAB-I (pCi/cc) X texp (hours) x DDEDCF-I [(mREM/hr)/(IJCi/cc)]

+XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CEDEDOF-i (mREM/pCi)

TEDEi (mREM) = XEAB (pCi/cc) X texp (hours) x TEDEDCF-i

[(mREM/hr)/(pCi/cc)]

where TEDEDCF-i

[(mREM/hr)/(pCi/cc)]

= DDEDCF-I [(mREM/hr)/(pCi/cc)]

+BR (cclhr) x CEDEDCF-i (mREM/pCi)

BR (cclhr) = breathing rate Step 6: Add the individual TEDEs to obtain the TEDE for the release (TEDEr~s):

TEDEris = Z [TEDEi]TEDEris = [Xris-i X (X/Q) X Qris x texp x TEDEDCF-I]

TEDEris = [(X/Q) x Qris x texp ] x $" [Xris-i x TEDEDcF-i]

Step 7: Calculate Thyroid CDE for each Iodine isotope for one hour exposure time at EAB. This is done using the appropriate FGR-11I dose conversion factors (DC Es), as discussed in the previous subsection.

CDETHY-i (mREM) = XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CDETHY-DCF-i (mREM/pCi)

Step 8: Add the individual Thyroid CDEs to obtain the Thyroid CDE for the release (CDEris): COEris = E' [CDETHY-i]

CDEris = Z [Xrls-i X (X/Q) x Qris x BR X CDETHY-DCF-i]

CDEris = [(XIQ) x Qris x texp x texp] x X [Xris-i X TEDEDcF-I]

Step 9: Determine the RS1 EALI 100 mREM TEDE threshold release concentrations for each noble gas (Xi00-1 pCi/cc). This is done by multiplying each noble gas' release concentration (Xris-i IJCi/cc) determined in Step 3 by the quotient of 100 mREM and the sum of the TEDEs for all of the radionuclides considered (TEDEr~s mREM). Only noble gas concentrations are adjusted because the gaseous effluent monitors are noble gas detectors.(Xi0o-i pCi/cc)/(Xris-i pCi/cc) = (100 mREM)/(TEDEris mREM)Xi00-i (pCilcc) = (Xrls-i pCi/cc) x (100 mREM)/(TEDEris mREM)The following demonstrates that the arbitrarily assumed release stream density has no effect on the final result.Xris-i x (100 mREM)Xl00-i =[(X/Q) x Qris x texp ] x $" [Xrls-i x TEDEDcF-i]

Xrcs-i x (1.0) X pris x (100)Xl0 0-i =[(X/Q) x Qris x texp X [Xrcs-i x PFi x pris x TEDEDcF-i]

Xrc-i x pris x (100)Xloo-J =pris X (X/Q) x Qris x texp x T' [Xrcs-i X PFi x TEDEDCF-i]

Xrcs-i X Prls X (100)Xl 0 0-i =pig x (X/Q) x Qris x Z [Xrcs-i x PFi x texp x TEDEDcF-i]

Xrcs-i X (100)X 1 o 0-i-=r(X/Q) X Qris x texp X E [Xrcs-i X PFi x TEDEDcF-i]

The assumed release stream density has no effect on the final result: it cancels out. Thanks to the power of Excel, it is easier to calculate a postulated dose rate and adjust release concentrations than to set up the above equations.

Now to perform a dimensional check: Xris-i X (100 mREM)Xloo-i =[Xris-i X (X/Q) x Qris x PFi x texp x TEDEDcF-i]

? ~(pCi/cc) x mREM IJCi/cc =(pCi/cc) x (sec/rn 3) x (m 3/sec) x (hour) x [(mREM/hour)/(IJCi/cc)]

9 X tRRE-M p Ci/cc=i,-,., ,c,. x x (ms/see) x (heufr) x [(mRE-M/heu-f)/(p Ci/cc)]pCi/cc 1 1/(p Ci/cc)]pCi/cc = pCilcc Step 10: Determine the RS1 EAL1 500 mREM Thyroid CDE threshold release concentrations for each noble gas (X50oTr-i pCi/cc). This is done using the same method as in Step 9. Again, the arbitrarily assumed release stream density cancels out and has no effect on the final result.Several general trends can be inferred from the equation derived in Step 9 above. Holding other factors constant:* Increasing the diffusion coefficient (X/Q m 3/sec) will reduce the 100 mREM release concentration.

• Increasing the release flow rate (Q CFM) will reduce the 100 mREM release concentration.

• Increasing the exposure time (t hours) will reduce the 100 mREM release concentration.

Southern Nuclear 0peratin9 Cornpany SOUTHERNA4 Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations I SMNH-13-021 COMPANY Unit: 1&2 I SHEET 22*Increasing the total release (£ [Xrcs-J x PFi x pris x TEDEDGF-i])

will reduce the 100 mREM release concentration.

Body of Calculation:

RU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2 OR 3)1. Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer.Liquid Effluent Monitors (HNP ODCM Table 2-4) 2 X Setpoints Liquid Radwaste Effluent Line (Detector D1 1-N007, Indicator D1 1-K604)(Ref. 64C1-0CB-009-0, H 16564, H26012)Service Water System Effluent Line (Detector D1 1-N008, Indicator D1 1-K605)(Ref. 64C1-0CB-008-0, H 16564, H26012)Gaseous Effluent Monitors (HNP ODCM Table 3-4) 2 X Setpoints Reactor Building Vent Stack (Detector 1D1 1-N020, Indicator Dl11K619, Detector 2D1 1-NO26A/B, Indicator 2D1 1K4636)(Ref 64CI-OCB-002-1, 64C1-0CB-002-2, H16564, H260 13, H260 12)Recombiner Building Vent (Detector D1 1-N078 Indicator Dl11R763A Detector D1 1-N079, Indicator D 11R763B)(Ref 64C1-0CB-003-1, H 16528)Main Stack (Detector D1 1-N071 Indicator D11IK6OOA, Detector D1 1N072 Indicator D 11K600B)(Ref 64C1-0CB-001-O, H 16564)2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the 0DCM limits for 60 minutes or longer.

Southern Nuclear 0perating Cornpany SOUTHERNE Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SM -102 COMPANYr Unit: 1&2 SHEET 24 RU2 UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability:

All Emergency Action Levels: (1)1.a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

LPersonnel report of iow water level.SEP Low level alarm (1/2G41N372)

Per PDMS 1/2G41N372 (LS-N3 72) spent fuel pool low level alarm is set at EL225'9" and personnel report of low water level.AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors 1 D21 -K601 A -Rx Head Laydown Area 1ID1I-K601 D -Refuel Floor 1 D21-K601 E -Drywell Shield Plug 1D21-K601 M -Spent Fuel Pool and New Fuel Storage area 2D21-K601 A -Rx Head Laydown Area 2D21-K601 M -Spent Fuel/Fuel Pool Areas 2D21-K601 E -Dryer/Separator Pool 2D21-K611 K -RPV Refuel Floor 228'2D21-K61 1 L -RPV Refuel Floor 228'The Drywe// radius is 18'10" -19' (H255 70) this approximation is reasonable for radiation calculations.

To calculate distance to drywel/ edge from radiation detector the radius of drywell is subtracted from distance from radiation detector to drywell center.

Southern Nuclear 0perating Cornpany SOT N4L Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY Unit: 1&2 SHEET 25 Distance to Monitor Detector Name Elevation edge of Range Ref drywell H 15880 1D21K601A 1D21NOO2A 233Head7"111548'H45515 Laydown Area 230 67-9=4'mR/'hr H66 H 15880 ID2Ik601B 1D21NOO2B Refueling Floor 233'0" 95"-19' = 76' 1-1E4 H45515 Stairway mR/hrH162 H 15880 1D21K601D 1D21NOO2D Spent Fuel Pool 23'" 75'"19' = 56' 1-1E4 H45515 Demin Equip mR/hr H15880 1D21K601E 1D2lNOO2E Drywell Shield 233'0" 21"-19' = 2' 1-1E4 H45515 Plug mR/hr H 16562 H15880 1D21K601M 1D21NOO2M 233'P 2"N9'W7'H45515 Fuel Storage 23'" 24'=7 mR/hr H66 H26010 2D21K601A 2D21NOO2A RHed 228'0" 60"-19' = 41' 1E4H27805 Laydown Area mR/hr___________H25990 H26010 2D21K601E 2D21NOO2E ryrSpo raol 228'0" 76"-19' = 57 1-1E4 H27805__________H25990 1E4H26010 2D21K601M 2D21NOO2M SFP Area 228'0" 75"-19' = 56' mhrH27805 H25990 H26010 2D21K611K 2D21NO12K Reco esl 228'0" 71'-19' = 52' 114H27809 Refueling mR/hr H59 H260 10 2D21K611IL 2D21NO12L Reactor Vessel 228'0" 62"19' = 43 1-1E4 H27809______ Refueling mR/hr H25990 Southern Nuclear 0peratinl Cornpany ISOUIHERNZ._

Pln: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-026 COMPANY Unit: 1&2 SET2 RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2 OR 3 OR 4)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer.This calculation is performed in Attachment M.Reactor Building Vent 2.6E-02 pCi/cc Range 1E-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D11P005 Detector 1/2D11N048 and 1/2D11N049 (Ref SX18761, S41454)Main Stack 8.1E+01 pCi/cc Range IE-3 to 1E5 uC~icc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1D11P006 Detector 1D11N0055 and 1D11N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the Site Boundary.3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the Site Boundary for one hour of exposure.4. Field survey results indicate EITHER of the following at or beyond the Site Boundary:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid ODE greater than 50 mrem for one hour of inhalation.

RA2 Significant lowering of water level above, or damage to, irradiated fuel.Operability Mode Applicability:

All Emergency Action Levels: (Table0R1 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on ANY Table R1 radiation monitors.Table Ri Refuel Floor Area Radiation Monitors Unit i Unit 2 1D21-K601 A -Rx Head Laydown Area 2D2l-K601 A -Rx Head Laydown Area 1D21 -K60 1 D -Refuel Floor 2D2 1-K601I M -Spent Fuel/Fuel Pool Areas 1D21-K601 E -Drywell Shield Plug 2D21-K601 E -Dryer/Separator Pool 1D21-K601 M -Spent Fuel Pool and New 2D21-K61 1 K -RPV Refuel Floor 228'Fuel Storage area____......

_________________,__________

2D21-K611 L -RPV Refuel Floor 228'Refuel Floor Ventilation Monitors Unit 1 Unit 2 ID11l-K609 A-D -Rx Bldg. Potential 2D1 1-K(609 A-D -Rx Bldg. Potential Contaminated Area Vent Exhaust Rad Contaminated Area Vent Exhaust Rad Monitor Monitor 1D 1l-K(611 A-D -Refuel Floor Vent Exhaust 2D1 1l-K61 1 A-D -Refuel Floor Vent Exhaust', :' ": .....2D1 1-K(634 A-D -Refuel Floor Rx Well Vent.______...._______________Exhaust

-. * ** ,2D1 1-K(635 A-D -Refuel Floor DW/Sep.________,_______,_______________________

Vent. Exhaust Southern Nuclear 0perating Cornpany ISOUTHERN E1, Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations j SM NH-13-021 COMPANY Unit: 1&2 SHEET 28 Monitor Detector Name Elevation Distance Range Ref to edge of_____ ____ _ ___ ____ drywell _ _ _ __ _ _ _Refuel Floor Area Radiation Monitors: 1D2IK601A ID21NOO2A Rx Head 233'0" 67'-19' = 1-1E4 H15880 Laydown Area 48' mR/hr H45515_________H16562 1D21K601B 1D21NOO2B Refueling Floor 233'0" 95"-19' = 1-1E4 H15880 Stairway 76' mR/hr H45515____ ____ ___ ____ __ _H 16562 1D21K601D 1D21NOO2D Spent Fuel Pool 233'0" 75"-19' = 1-1E4 H 15880 Demin Equip 56' mR/hr H45515 H 16562 1D21K601E 1D2INOO2E Dirywell Shield 233'0" 21'-19' = 1-1E4 H15880 Plug 2' mR/hr H45515_______H16562 ID21K601M ID2INOO2M SFP & New 233'0" 26"-19' = 1-1E4 H15880 Fuel Storage 7' mR/hr H45515 H16562 2D21K601A 2D21NOO2A Rx Head 228'0" 60'-19' = 1-154 H26010 Laydown Area 41' mR/hr H27805____________H25990 2D21K601E 2D21NOO2E Dryer/Separator 228'0" 76'-19' = 1-1E4 H26010 Pool 57' mR/hr H27805 H25990 2D21K601M 2D21NOO2M SFP Area 228'0" 75"-19' = 1-1E4 H26010 56' mR/hr H27805_________

_______ H25990 2D21K611lK 2D21NO12K Reactor Vessel 228'0" 71"-19' = 1-1E4 H26010 Refueling 52' mR/hr H27809 H25990 2D21K611L 2D21NO12L Reactor Vessel 228'0" 62"-19' = 1-1E4 H26010 Refueling 43' mR/hr H27809____ ___ __ ____ ___ ___ __ ____ _ _ ___ ___ H25990 Southern Nuclear Operating Cornpany Pan: N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET-029 COMPANY Unit: 1&2 SET2 Monitor Detector Name Range Ref Refuel Floor Ventilation Monitors: 1/2D11K609 1/2D11N010 Rx Bldg H16565 A-D A-D Ventilation H26013 Exhaust 0.01- FSAR Radiation lOOmR/hr Table Monitor 11.4-1 PDMS 112D11K611 1/2D11N012 Refuel Floor 0.01- H26012 A-D A-D Vent Exhaust lOOmR/hr H16014 FSAR Table 11.4-1___ ___ __ ___ ___ ___ ___ PDMS 2D11K634 2D11N024 Refuel Floor 0.01- H26012 A-D A-D Exhaust lOOmR/hr H27158 2D11K635 2D11N025 Refuel Floor 0.01- H26012 A-D A-D Exhaust lOOmR/hr H27 158 3. Lowering of spent fuel pool level to (site-specific Level 2 value)*.*The level 2 value was not available at the time of this calculation.

RS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR 2OR 3)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: This calculation is performed in Attachment I.Reactor Building Vent 2.6E-1 pJCi/cc Range 1E-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D1 1P005 Detector 1/2D1 1N048 and 1/2D1 1N049 (Ref SX18761, S41454)Main Stack 8.1E2 pJCi/cc Range IE-3 to lE5 uCi/cc (ESAR U2 Table 11.4-1 sheet 1)Monitor 1DI1PO06 Detector 1D11N0055 and ID1 1N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid ODE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the Site Bounday.* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Southern Nuclear Operating CornpanyPlant: HNP Title: NEI 99-01 Rev 6 EAL Calculations SMNH-13-021 COMPANY Unit: 1&2 SHEET 31 RG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid ODE.Operating Mode Applicability:

All Emergency Action Levels: (1 OR20OR 3)1. Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Reactor Building Vent 2.6E0 tpCi/cc Range 15-3 to 1E6 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1/2D11P005 Detector 1/2D11N048 and 1/2D11N049 (Ref SX18761, 341454)Main Stack 8.1E3 pJCi/cc Range 1E-3 to 155 uCi/cc (FSAR U2 Table 11.4-1 sheet 1)Monitor 1D11PO06 Detector 1D11NO055 and 1D11N056 (Table 2 SX18761, H 16564)2. Dose assessment using actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the Site Boundary:* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

CA1 Loss of RPV Inventory.

Operability Mode Applicability:

Cold Shutdown, Refueling Emergency Actuation Levels (1 OR 2)1. Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).

According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point. According to H16 145 and H26 189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35".2.a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level increase in any of the following due to a loss of RPV inventory.

Drywell Floor Drain Sumps Drywell Equipment drain Sumps Torus Torus room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks Southern Nuclear Operatingl Company SOUTHE=RNZ4 lnt N Title: NEI 99-01 Rev 6 EAL Calculations SMHEET3302 COMPANY Unit: 1&2 SET3 CS1 Loss of RPV inventory affecting core decay heat removal capability.

Operability Mode Applicability:

Cold Shutdown, Refueling Emergency Action Levels: (1 OR 2 OR 3)1.a. Secondary CONTAINMENT INTEGRITY not established.

AND b. RPV level less than -41" (6" below the Level 2 actuation setpoint).

According to S25213 page 12 and 14 ECCS actuates at level 2 actuation set point.According to H16145 and H26 189 the level 2 instruments are 1/2D21N692A-D and 1/2D21N682A-D.

According to PDMS the instruments are set for -35". Therefore

-35"-6" =-41" 2.a. Secondary CONTAI NMENT INTEGRITY established.

AND b. RPV level less than -158" (TOAF).The reactor vessel top of active fuel was calculated in Fuel Clad Baffier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158".3.a. RPV level cannot be monitored for 30 minutes or longer.b. Core uncovery is indicated by the following:

• UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery: Dryeli Floor Drain Sumps Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged.

Operating Mode Applicability:

Emergency Action Level: Cold Shutdown Refueling (1 OR 2)t.a. RPV level less than -158" (TOAF)The reactor vessel top of active fuel was calculated in Fuel Clad Barrier Potential Loss 2.A. The top of fuel was found to be at vessel level of -158.44".This can be rounded to -158" AND b. ANY indication from the Containment Challenge Table C1.2.a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by the following:

• UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery: Drywell Floor Drain Sumps Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps Turbine Building Floor Drain Sumps Torus Rad Waste Tanks Torus Room Sumps Radiation monitor readings indicative of core unco very are investigated in Attachments J and K resulting in no monitors able to provide on-scale indications of core unco very.AND c. ANY indication from the Containment Challenge Table C1 Containment Challenge Table C1 Containment H2 greater than or equal to 6% AND 02 greater than or equal to 5%Primary Containment Pressure:

greater than 56 psig Secondary CONTAINMENT INTEGRITY NOT established*

Secondary Containment radiation monitors greater than Max Safe values (SC EOP -Table 6)*If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Damage to a loaded cask CONFINEMENT BOUNDARY E-HU1 Southern Nuclear 0perating Cornpany SOUTHERNR lnt N Title: NEI 99-01 Rev 6 EAL Calculations SNHEET3502 COMPANY Unit: 1&2 SET3 Operating Mode Applicability:

Emergency Action Level: ALL (1)1. Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table El.The calculation is performed in Attachment L.Table El Location of Dose Rate [ Total Dose Rate____________________(Neutron

+ Gamma mRlhr)HI-TRAC 125 Side -Mid- height [450 Top 11l0 HI-STAR 100 or HI-STORM 100 Side -60 inches below mid- 80 height Side -Mid- height 80 Side -60 inches above mid- 30 height Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 Southern Nuclear 0perating Cornpany latHN Title: NEI 99-01 Rev 6 EAL Calculations SMH3-2 COMPANYW Unit: 1&2 SHEET 36 Fission Product Barrier Emergency Action Levels Fuel Clad Barrier: Emergency Action Levels Fuel Clad Barrier Potential Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be determined.(Ret. H16 145 and H26189)According to COLR for HNP the currently used fuel is GEl4. 'According to NED C-32868P Rev 5 Appendix A (Reference of the COLR) the fuel length for GEl4 fuel was increased from 148" to 150" inches. The Appendix A is attached below. Thus the top of the fuel per TS Bases 2.1.1.3 is 158.44 inches below instrument zero.According to 31EO-EOP-015-1 and 31EO-EOP-015-2 "CPI Flow Chart" operators are instructed to maximize water injection rates from alternate injection subsystems when reactor water level drops below -155 inches of instrument zero. This value is more conservative than the actual TOAF level.

NEDC-32868P

-Revision 5 GNF Proprietary Information

-Class ill Appendix A Minor Modifications to the GEl4 Design The nominal pellet density for thle GEI4 design has been increased from 95_5% to 97.0% of theoretical density (TD). The upper specification limit, however, has not been changed. In addition, for the full length U0 2 fuel rods, barrier and non-barrier tubing designs, the active fuel length has been increased from 3759 mm (148 in.) to 3810 mm (150 in.).All GESTAR II criteria identified in Sections 2.1 through 2.13 have been re-evaluated and it has been demonstrated that all criteria are satisfied.

The results of the evaluations are contained in DRF J11-03667-00 and DRF J1 1-03467-01.

As part of the evaluation, a new NRC approved method was used for the ATWS analysis (Section 2.14). The ODYN one-dimensional transient analysis code was used to perform the required ATWS analyses.

All results are below the specified limits for compliance and are documented in DRF B21-00676-00.

The NRC approved version is documented in the lollowing:

MFN-99-029, Letter, S. Richards (NRC) to J, Kiapproth (GE), "~Safety Evaluation by the Office of Nuclear Reactor Regulation on NEDC-24154P, Supplement 1", November 18, 1999.MFN-052-98, Letter, T. Essig (NRC) to J. Quirk (GE), "Safety Evaluation by the Office of Nuclear Reactor Regulation on NEDC-24154P, Supplement 1 (TAC No.MA3478)", November 17, 1998.

Fuel Clad Barrier Loss Threshold 4.A DWRRM greater than 1,400 R/hr.In Attachment C the detector radiation level of 1.4E3 R/hr was calculated.

The calculation used core inventory from NL-06-1637 to calculate isotopes concentrations.

The calculation for DEI131 was performed to find a ratio to DEl 300uCi/gm.

GRODEC was used to calculate the fluence within the drywell.Cylinder geometry was used to calculate the geometric fraction.Fuel Clad Barrier Loss Threshold 5.A Offgas Pre- and Post-Treatment Monitors Offscale High AND Fission Product Monitor Offscale High.Attachment D performed an evaluation for Offgas Pre- and Post-treatment monitors D1 1-K615 (section A of Attachment D) and Containment fission product monitors D11IP010 (section B of Attachment D). It was found that these instruments will be off scale.RCS Barrier: Emergency Action Levels RCS Barrier Loss Threshold 1I.A Primary containment pressure greater than 1.85 psig due to RCS leakage.LIS 1C71N650A-D 1.85 psig Ref. PDMS LIS 2C71N650A-D 1.85 psig Ref. PDMS References (H 16568, POMS)RCS Barrier Loss Threshold 2.A RPV water level cannot be restored and maintained above -155 inches or cannot be determined (Ref. H16145 and H26189)The reactor vessel top of active fuel was calculated in Fuel Cla'd Barrier Potential Loss 2.A.RCS Barrier Loss Threshold 4.A DWRRM greater than 40 R/hr.In Attachment E the detector radiation level of 40 R/hr was calculated.

The calculation used core inventory from NL-06-1 637 to calculate isotopes concentrations.

The calculation for DEI13I was performed to find a ratio to DEl Southern Nuclear 0peratinl Cornpany SOUTHERNE4i Plant: HNP Title: NEI 99-01 Rev 6 EAL Calculations I SMNH-13-021 COMPANY Unit: 1&2 I SHEET 39 2uCi/gm. GRODEC was used to calculate the f/uence within the drywell. Cylinder geometry was used to calculate the geometric fraction.RCS Barrier Loss Threshold 5.A Drywell Fission Product Monitor reading 5.0 x 105 cpm.This EAL is to cover drywve// fission product monitor indications that may indicate loss or potential loss of the RCS barrier. Attachment F determined that the reading on drywe// fission product monitor Dl 1K630 of 1E6 cpm will indicate potential loss of RCS barrier. Per SX18062 page 34 the monitor K630 range is 10 to lO^6 cpm.Primary Containment Barrier: Emergency Action Level: Primary Containment Barrier Potential Loss Threshold 1l.A Primary containment pressure greater than 56 psig.Containment Design Pressure:

56 psig (SS2 102005 section 304)(SS6902005 section 304)Primary Containment Barrier Potential Loss Threshold I .B Greater than or equal to 6% H2 AND 5% 02 exists inside primary containment.

Explosive mixture inside containment

>_6% Hydrogen (Ref. RG1.7 pgl. 7-6)> 5% Oxygen (Ref. CALC BH2-CS-52-2P33-2 pg 4 and 9)(Ref. CALC BH1-CS-33-P33-06 pg 8 & A-I)Primary Containment Barrier Potential Loss Threshold 4.A DWRRM greater than 26,000 R/hr.The evaluation of expected radiation readings on DWRRM (Dl11K621) was performed in Attachment G of this calculation.

The detector is expected to read 2.6E4Rlhr.

The range of this instrument is 1-10^7 R/hr (established in attachment c).

SU3 Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode: Power Operation Startup Hot Standby Hot Shutdown Emergency Action Levels: (1 OR 2)1. Pretreatment Radiation Monitor reading greater than 240,000 pCi/sec for greater than 60 minutes.According to Tech. Spec. Bases 3.7.6 page B3. 7-31 and Tech. Spec. section 3.7.6 page 3. 7-16, the gross gamma activity rate of the noble gases measured at the main condenser evacuation system pretreatment monitor station shall be <240mCi/second or <240, O00pCi/second.

According to 64C1-0CB-006-1/2 procedures the offgas pretreatment radiation monitors are 1/2D1 1-K601 and 1/2D1 1-K602 OR 2. Sample analysis indicates that the reactor coolant specific activity is EITHER:* Greater than 0.2 pCi/gm and less than or equal to 2.0 pCi/gm dose equivalent 1131 for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.* Greater than 2.0 pCi/gm dose equivalent h131.According to Tech. Spec. Bases 3.4.6 page B3. 4-25 and Tech. Spec. section 3.4.6 page 3. 4-11, the specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity 0.2 pCi/gm. A condition

>0.2 IJCi/gm but 2.0 IJCi/gm must be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A condition

>2pC~igm requires immediate action.

Southern Nuclear Design Calculation IPlant: HNP Unit: 1&2 Icalculation Number: SMNH'13-021 ISheet: A-I Attachment A -SNC Emergency Planning Concurrence Calculation Number: SMNH-13-021 Calculation Version: 1 Calculation Title: NEI 99-01 Rev 6 EAL Calculations I the undersigned have reviewed the subject calculation and concur that:* Its Methods of Analysis conform to the guidance of NEI 99-01 Revision 6* Its Assumptions are consistent with the guidance of NEI 99-01 Revision 6* Its conclusions are consistent with the Methods of Analysis, Assumptions, and Design Inputs..L 1 J' / t, Emergency PlanningSN Name / Signature 0 /Date / Organization Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 1 of 19 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors (Fuel Clad Barrier Loss Threshold 4.A).Defining Units that MathCAD will understand MWt :=r 106W Ci: *- Bq and 2.7.10[tCi := l0-6 Ci Rad := 0.01IGy Per FGR-11 page 219 Remn Remn:= 0.01Sv Per FGR-11 page 219 mRem := -1000 The source terms for noble gases and iodine's are provided in NL-06-1 637 Enclosure 1 page 16 and are shown below.I Isotopes : "Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe- 1381-13 1""I-132""1-133""I-134""1-135" Corelnventory

3.30E+03' 3 .78E+02 6.92E+03 1 .32E+04 1 .86E+04 2.26E+04 3.03E+02 5.27E+04 1 .58E+03 1.89E+04 1.09E+04 4.81IE+04 4.52E+04 2.72E+-04 3 .93E+04 5.52E+04 6.05E+04 5.16E+04 Ci MWt)

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 2 of 19 CorePower

=z 2818.MWt RCS inv :=9965ft 3 FRelease:=

0.05 DW_Volu 1 := 146010ft3 SPVol~u 1:= 112900ft 3 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.The dry well free volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.VolContainmentu1

= DWVo1lu 1+ SPVolu1 = 2.589 x 105-ft3 DW_Vo1u 2 :=146266ft 3 SP_Volu 2:= 109800ft The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol Containmentu 2:= DW_Volu2 + SPVolu 2= 2.561 x( 105.ft3 ContainmentVol Avg : VolContainm entul + VolContainmentu 2 53= 2.575 x 105.ft3 2 To determine the RCS inventory mass the water specific volume needs to found as follows: Ul: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 °F + 532.0 °F TAvgu 1 := 2= 462.2.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 °F TAvgu 2:= 2= 480.35.°F SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEl 99-01 Rev 6 EAL Calculations 3 of 19 The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is l060psia.ft 3 VRCS water :=0.0199161

--According to http://wwwv.spiraxsarco.

corn/resources/steam-tab les.asp the specific volume for 1060psia and 480.35 F conditions is 0.0199161ft^3/lbm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv8 RCS mass .- -2.27 x 108-gm'URCS water The concentration of isotopes in RCS is calculated as follows: CoreInventory.

CorePower.

F_Release COnCRCS isotopes : RCS mass augment(Isotopes, COnCRCS isotopes)

=0 1 0 "Kr-83m" 2.049"10-3 1 "Kr-85" 2.347" 0-4 2 "Kr-85m" 4.296"10-3 3 "Kr-87" 8.195" 0-3 4 "Kr-88" 0.012 5 "Kr-89" 0.014 6 "Xe-131lm" 1.881"10-7 "Xe-133" 0.033 C 8 "Xe-133m" 9.809"10-4' gin 9 "Xe-135" 0.012 10] "Xe-135m" 6.767"10-3 111 "Xe-137" 0.03 123 "Xe-138" 0.028 13 "I-131" 0.017 141 "I-132" 0.024 15 "I-133" 0.034 161 "I-134" 0.038 17 "I-135" 0.032 SPlant: HNP U1 & U2 SNC CALCULATION MN-301Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 4 of 19 The DEl for the RCS is calculated in the following work sheet: The Coolant Iodine activities are provided in NL-06-1637 Table 28 Enclosure 1.The dose conversion factors are provided in FGR 11. The RCS mass was calculated just above in this attachment.

The Iodine activities were taken from the calculated above matrix.gConCRCS isoto pes 1]3 Co"I-13_*t (0.017 " rI-132"15 COnCR S-i~ps4 /0.024 Iodines:=/'-3' Activity := ConcRCS isotopesl 5 /l0.034 C L0.038 gm k."-15"ConCRCS-isotopesl 6 .0.032 ConCRCS-isotopes 1 7 2.92E-07 1 .74E-09 BqFRlIhlaintbe DCFFGR11l:

4.86E-08 --~ The dose conversion factors are taken from K8.46E-09J The total exposure to iodines is calculated as follows: (1.826 x 104`157.23 5> Rem ExPiodines

= (DCFFGR1 1'Activity)

=6.169 x 103 gm 40.064 1.004 x 1O 3 SumExPiodines

=VEx..dne 4.6 x14Remn L~Podies 2.53 x10gm The DEI131 equivalent of the exposure from iodines calculated by dividing the Sum of iodines exposure by 1131 DCF factor: Sum-ExPiodines 14 ~ICi DEl131_eqv

= = 2.37 x 10 -DCFFGR110 gm Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-0 1 Rev 6 EAL Calculations 5 of 19 FDEJ300 := gm -0.0 13-~DE1131 _eqv This is the fraction that will be applied to iodine's and noble gases concentrations to adjust to RCS DEl 300 uCi/gm equivalent as described in NEI 99-01.Amount of isotopes in the drywell is found as follows: In this calculation the core inventory is multiplied by core power, containment release fraction, and the DEl of 300uCi/gm fraction.

The amount of total curies released are distributed throughout the drywell and suppression pool. The concentration then is multiplied by the U1 DryWell volume. This way we have the total curies inside the DryWell only. The DryWell for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.Corelnventory.

CorePower.F_Release.

FDE1300.DWVoI 1 1 2 Isotopes in DW: augment(Isotopes, Isotopes in DW) =ContainmentVol Avg*0 "Kr-83m" 3.343E+003

1. "Kr-85" 3.83E+002"2 "Kr-85m" 7,011tE+003 3 "Kr-87" 1.337E+004 4 "Kr-88" 1.884E+004 5 "r-9 2.29E+004 6 "Xe-131lm" 3.07E+002 7 "Xe-133" 5.339E+004 181 "Xe- 133m" 1.601E+003 9, "Xe-135" 1.915E+004

.10 "Xe-135mn" 1.104E+004 11: "Xe-137" 4.873E+004 12 "Xe-138" 4.579E+004 13 "I-131" 2.756E+004 14 "I-132" 3.981E+004 15 "1-133" 5.592E+004

16 "I-134" 6.129E+004 17 "1-135" 5.228E+004.Ci

[Plant: HNP Ul & U2 SNO CALCULATION SMNH-13-021 Attachment CI Title: NEI 99-01 Rev 6 EAL Calculations 6 of 19 I The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air mixture density for pressure 44.7psia and temperature of 343F (NL-06-1 637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

GRODEC does not have an input for absorption of coefficients.

DW_VoIlu 2 = 4.142 x 109.mL 1 -3 gm Psteam- -1.53xx 10 .ft 3 mL 10.4690 lb According to http://www, spiraxsarco.

com/resources/steam-tab les.asp the specific volume for 44.7psia and 343 F conditions is 10.4690ft^3/Ibm.

This web site was validated in Attachment H. The copy of the web page and the provided information is attached to this attachment below. The drywell pressure and temperature were obtained from NL-06-1637 Enclosure 1 table 11 and table 12.According to PDMS the Drywell Wide Range detector is MPL 1D11N003A&B and 2D11N003A&B.

The detectors are connected to D11K621 Recorders.

As described in the following table: DrawingInstrument Tag Range MPL Sheet No R/hr Model N 1D11N003A A16481 DuIB QD11-RE-N003A lE0- 1E7 877-1 1D11N003B A16481 D11C QD11-RE-N003B lE0- 1E7 877-1 2D11N003A A26481 DlIA Q2D1l-RE-N003A lE0- 1E7 877-1 2D11N003B A26481 DluB Q2D11-RE-N003B lEO- 1E7 877-1 DWG H16566 DWG H26017 I I I I I I!I I I I 1 t!L.DWG H16566 GROSS Q~IM~ GUt~SftS l~A~.-i These drawings show that the detectors are inside the drywell.

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 7 of 19 DWG H16032 Per H16241 and H26417 the instrument location is at elevation of 156ft and approximately 27 ft from the center of RPV. This model will approximate the volume seen by the detector through a cylinder of 41.5 ft high. This number corresponds to the elevation above drywell floor. The radius of the cylinder will adjusted to the drywell free volume that is in the line of site of the detector.

Through the visual inspection of drawings H26417 and H16241 (shown below) it can be seen that about 2/3 of the drywell volume can be seen by the detector.E1llAllwl SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 8 of 19 Drawing H26417 Shows that approximately 2/3 of the containment volume is seen by the detector.Drawing H16241 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 9 of 19 As the result the geometry of detector in this volume will be modeled as a cylinder with the height of 41.5ft. The cylinder radius will be adjusted to match 2/3 of free volume of the drywell. The geometry factor will be evaluated for the point P1.Fig. 6.4.-7. Geometry for non-absorbing cylindrical volume source; interior, surface and exterior points.The uncolided fluxes at interior and exterior points of a non-absorbing cylindrical volume source are provided in Engineering Compendium on radiation shielding Volume 1 pages 381 and 382.(6.4.-19)3 H16032 H16241 and H26417 At P 1 : Sv (.& ;Heightcy1

= 156ft -114.5ft = 41.5.ft DWVolu 2 = 1.463 x 105.ft3 Volumecylinder

=fi r.R2*height table 34 Enclosure 1.Standard equation for cylinder volume.This is fraction of the containment line of site for the detector.FLine :-3 F FLine. DWVolu 2 RadiUSdetector

= = 27.348.ft 7r.Heightcy 1 Heightcyl

-1.5 17 RadiUSdetector the function (aR is determined via page 382 table 6.4-3 of Engineering Compendium of Radiation Shielding Volume 1. Linear interpolation will be used.

[ Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment~il:

E 90 e A acltos1 Cf1 xI:= 1.0 Heightcy -1 .1 x.-RadiUSdetector Y :=0.628 x3:= 2.0 3:=0.390 (x 2 -Xl)(Y 3 -Yl)+ 050 Y := (x 3_ Xl) +Y .0 Pa := Y'Pa = 0.505 Sv-Heightcy 1= 2 .'a ---= 3.1929198406362356406.Sv~m or Heightcy 1 = 1.265 x 103.cm Heq : 1.265 x 103 cm= 2ve 319.31217771913149332.Sv.c4 The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.1.6 ergs*.. o-MV-cm 2" Ia 4 3600S 10O0 ergs/(g .rad)

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 11 of 19 1 -.16 ergs cm 3 6 0 s MeV gm h Rad-hr or Geometricfactor.

Sv 10ergs g~rad 1.6.10- x 3600 -5= 5.76 x 10 100 Defining Units that MathCAD can process.q:=1.60219.10-1 9.coul eV := qe-volt Fundamental charge of electron page 6 of "Fundamental of Nuclear Science and Engineering'.

defines the unit of charge.MeV := 106.eV cc :=mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

(3.00E-01 4.OOE-0 I 5.00E-01 6.00E-0 1 8.00E-01 I1.00E+00 1 .50E+00 2.00E+00 3.00E+i00 4.00E+00.MeVSv:

1.00E+05 3 .62E+04 2.77E+05 4.76E+05 4.96E+05 7.3 8E+05 4.00E+05 3.35E+05 0.00E+00 MeV CO" Sec tI absair : r0.0288" 0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.0238 0.02112 cm gm

[Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 12 of 19 HeightcyI

= 1.265 x 103.cm oa=0.505 Calculated above Calculated above Radiationfield

= ,,Hegt~ absair =0 1 2 3 4 5 6 7 8 9 0 48.637 54.513 19.8 151 253.343 255.767 347.937 175.324 130. 176 0 Rad hr The unit conversion was handled by MathCAD internally.

This unit conversion was explained above in this attachment.

SM Radiationf ied :=

=1.436 x 103.a-Ic , hr~ccording to the S43177 Operator Manual Figure 1-4 the model number 877-1 detector has 1 to 1I atio of radiation present to radiation measured or shown. Therefore, the detector would read i.436E3R/hr.

This can be rounded to 1 .4E3R/hr.

The range of this instrument is I to 1 0A7 R/hr k stablished above).GRODE Inpu 1 18 60,3.343e+3,0 61 ,3.830e+2,0 62,7.011 e+3, 0 63,1 .337e+4,0 64,1.884e-'-4,0 65,2.290e+4, 0 141 ,2.756e+4,0 142, 3. 981e+4,0 143,5.592e+4,0 144,6. 129e+4,0 145,5.228e+4.0 FPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEL 99-0 1 Rev 6 EAL Calculations 13 of 19 146,3.070e+i2,0 147,5.339e+4,0 148,1.601e+3,0 149,1. 915e+'4,0 150,1.104e+4,0 151,4.873e+4,0 152,4.579e+4, 0 OUTPUT OF GRODEC CALCULATION DATA FILE NAME:FCEAL3 INPUT DATA LISTING 1 K<I-I-I-I-ISOTOPE INITIAL ADDITION ACTIVITY(CI)

RATE(CI/HR)(R-83M 3.343E+03 0.000E+i0 (R-85 3.830E+02 0.000E+00 (R-85M 7.011E+03 0.000E+0 (R-87 1.337E+04 0.000E+00 (R-88 1.884E+04 0.000E+00 (R-89 2.290E+04 0.000E+00-131 2.756E+04 0.000E+00-132 3.981E+04 0.000E+00-133 5.592E+04 0.000E+00-134 6.129E+'04 0.000E+00-135 5.228E+04 0.000E+00 (E-131M 3.070E+02 0.000E+(.E-133 5.339E+04 0.000E+0(.E-133M 1.601E+03 0.000E+(.E-135 1.915E+04 0.000E+0((E-135M 1.104E+'04 0.000E+(.E-137 4.873E+-04 0.000E+0OC (E-138 4.579E+04 0.000E+0()0 0 00 0 00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

= 1000 OR 1.000000 HOURS Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 14 of 19 TOTALACTIVITY

=323379.000000 CURIES ISOTOPE ISOTOPI NUMBER NAME E INITIAL ADDITION ACTIVITY ACT(CI) RATE(CI/HR) (Cl)3.343E+03 0.000E+00 2.303E+03 60 61 62 63 64 65 75 76 77 84 141 142 143 144 145 146 147 148 149 150 151 152 161 163 164 171 KR-83M KR-85 KR-85M KR-87 KR-88 KR-89 RB-87 RB-88 RB-89 SR-89 1-131 1-132 1-133 1-134 1-135 XE-131M XE-I133 XE-I133M XE-I135 XE-I135M XE-I137 XE-138 CS-I135 CS-1 37 CS-i138 BA-137M 3. 830E+02 7.011 E+03 1 .337E+04 1 .884E+04 2. 290E+ 04 0.000OE+00 0.000E+00 0.000E+00 0.000E+00 2.756E+'04 3.981E+04 5. 592E+04 6. 129E+04 5.228E+04 0.000E+'00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000OE+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+'00

3. 830E+i02 5.989E+03 7.737E+03 1.471E+04 4.800 E-02 I1.696E-11I 1.441 E+04 4.006E+02 8.779E-0I 2.746E+04 2. 930E+04 5.404E+04 2.755E+04 4.71 3E+04 3.070E+02 0.000E+00 3.735E+02 5.339E+04 0.000E+00 5.341E+'04 1.601E+03 0.000E+00 1.581E+03 1.915E+04 0.000E+00 2.165E+04 1.104E+04 0.000E+00 7.681E+02 4.873E+04 0.000E+00 1.142E+00 4.579E+04 0.000E+00 4.256E+-03 0.000E+00 0.000E+00 5.404E-07 0.000E+00 0.000E+00 1.204E-02 0.000E+00 0.000E+00 9.922E+03 0.000E+00 0.000E+00 1.204E-02 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME RB -87 CS-135 ACTIVITY (CI)I1.695969E-11I 5.403755E-07 WHAT IS THE SOURCE VOLUME (CC)4. 142000E+09 WHAT IS THE SOURCE DENSITY (GMICC)1 .530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.

SPiant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 15 of 19 I HE NUMBER. OF ENERGY GROUPS INPU I 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 2 3 4 5 6 7 8 9 10 3.000000E-01 4.000000E-01 5.000000E-01 6.000000OE-01 8.000000E-01 1 .000000E+00 1 .500000E+00

2. 000000E+00 3.000000E+00

4. 000000E+00 MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO.UC/GM SOURCE STRENGTH UC/CC Cl KR 83M KR 85 KR 85M KR 87 KR 88 KR 89 KR 89 RB 88 RB 89 SR 89 45 47 46 48 49 50 51 67 68 21 4 5 6 7 8 I I I I I 131 132 133 134 135 3.634336E+0 6.044329E+01 9.450985E+0 1.220921 E+03 2.321087E+03 7.574059E-03 7.574059E-03 2.274495E+03 6.320660E+01 1 .385227E-01 4.333311E+03 4.622957E+03 8.527853E+03 4.347433E+03 7.436683E+03 5.89351 9E+C 2.494256E+IC 1 .212008E+C 1.801554E-01 1 .900384E-03

1.

1 .900300E-0:

2 5.560535E-01 9.247824E-02 2 1.446001E+0C 1 .868009E+00 3.551264E+00 1.158831E-05 1.158831E-05 3.479977E+00 9.670610OE-02 2.11 9398E-04 6.629966E+00 7.073123E+00 1.304762E+01 6.651 572E+00 1.137812E+01 I1 9.017084E-02 I 1.289410E+01 132 3.816212E-01 I 5.227550E+00 132 1.854373E-01

2. 756378E-04 S1.027530E+00

2. 907587E-06 I 2.395366E+00 3 2.907459E-06 2.303174E+03

3. 830449E+02

)5.989335E+03 7.737294E+03 1 .470933E+04 4.799878E-02 4.799878E-02 1 .441406E+04 4.005567E+02 8.778545E-01 2.7461 32E+04 2. 929688E+04 5.404322E+04 2.755081 E+04 4.7128 19E+04 5.340735E+04 1 .580675E+03

2. 165251 E+04 7.68081 3E+02 1. 141692E+00 4.256031 E+03 1 .204322E-02 9.921605E+03 1 .204269E-02 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 137 35 CS 138 36 BAI137M 78 Piant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Ttle: NEI 99-01 Rev 6 EAL Calculations l6 of 19 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)I 2 3 4 5 6 7 8 9 10 3.000000E-01 4.000000E-01 5.000000E-01 6.000000E-01 8.000000OE-01 1 .000000E+00 1 .500000E+00 2.000000E+00 3.000000E+00 4.000000E+-00

9. 166370E+04 1.001 597E+I05 3.621505E+04 2.766909E+05 4.763061 E+05 4.956754E+05

7. 378778E+05 3.999907E+05 3.3536 18E+05 0.000000E+00 1 .265570E+

15 1. 037153E+ 15 3.000054E+

14 1. 91 0089E+15 2.466075E+-15

2. 053088E+ 15 2. 037527E+ 15 8.283807E+

14 4.630229E+14 0.000000E+00 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment C 'Title: NEI 99-01 Rev 6 EAL Calculations 17 of 19 I Heere Abeet Us Products & Service. -teduetnes

& AapOce4less.

-TrelsIna flhesuras Yes, wrettere:

tltn I > Ta p. Sub Saturae Water Regie.Sub Saturated Water Region -Steam Table Atany pressare water belowissaturation tem atre issaid tobelsna sub saturated state Fur example. wate, at a iressure aft1 atmosphere aod temperature below Ohe saturated tempsteratur'e at iS sub saturated Water at a pressare sf 10 atutospheres has a saturation temperature of 1l0"¢, antd so mater betow Ohis terepeatare is also sub saturated Learn mere about steam et our tutorial -Wf.atItoomLL Set your ofioots fur" these steam tables Note: -You cannot use commas (,) as decimal points.Please use periods (.)Example: 1.02 not 1,02!Feature* Tratutt Out lt en~*tcfsaeta 0 lnputs Pressure Temperature Pressure and Temperature El~ StogIe Value Table put abratute El El E~fl Z~Z Z~Z Vapout Pressure Saturation Teomperature Specdflc Enthralpy of Watet (ho)D~ensiy of WaterVolume of Water (o)Specdl Envopy of Water (sr)trith JflcgK El El I, Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Atcmn Title: NEI 99-0 1 Rev 6 EAL Calculations 18tchofnt19, itl n lsite for Spir you ace ttere(: t Etem5 .I S~perthnated Steam Resists Superheated Steam Region -Steam Table The superheated steam regon depics steam at a temperature trigter than its saturation temprature Should saturated steam he heated at constant pressure, its temperature wit rise. producing sarperheated steam Learn mere about steam In our tutorial -Wa sSem Set your toogt for these steam tablesI Note: -YOU caritot use commas (,} as decimal points.Please use periods (0 Example: 1.02 not 1,02 Output t Sigl Vatue Table Supertheat Temperature F r-Pv Feature.L Treater 0 atofe Saturation Temperature Degrees Superheat SpctcEnstralpy at Water (Irs)Specific Enthalpy ottEnaporation (N)Specitic Enttalpy of Superhseated Steam (h)Density of Steam Specific Vohene of Steam (s)Jag JAro jag 555 Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment C Title: NEI 99-01 Rev 6 EAL Calculations 19 of 19 Table 6.1.-I. Units used to express S tabulated in relation to the units used to express________ _______S__________

Point source Line source Surface. Volume o0 as a factor of to. S SOUrce source S 0 S. S S y-photons em 2 s?-photons cm~s MeV fim 2 s MeV cm~s MeV cm 2 s rads h rads h R h R R R?'-photons mCi (milhicuries) m$mCi V-photons MeV V-photons MeV mCi rag-equiv, Ra y-photons cms mCi cm MeV oms mci cm mCi cm V-photons¢ms MeV cins V-photons cms MeV cms mCi cm y,-photons cm 2 s mCi cm 2 MeV cm2S mCi cm 2 mCi cm 2 y-photons cm's MeV cm 2 s v-photons cm 2 s MeV cm~s mCi y-photons cm 8 s mCi MeV cm8$mCi cm 8 mCi V-photons M~eY cmas y-photons cm3 s MleV mCi.70 disintegrations v-photons 3.1~smCi ndisintegration 3,7" 10'disintegrations V -photons MeV smCi disintegration ,gphoton 4?MeV 4.Jcmn~smCi Me V 1.6 ~ ergs _ cm 2 s 100 ergsl(g~rad) 1 610 ergs cm 2 s 100 ergs/(g, rad)eV 1.6io ergs _cm2 60 87.7 ergs/(gR)c rs m 2 s M.. e- -V g .-g- Iio B"7.7 ergsl (gR)Rcm 2 43r. Kr hmCi Rcm 2 4~84hmg-equiv.Ra mg-equiv.

Ra mg-equiv.

Ra Img-equiv.

Rn cm cm 1 cm 3 Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment Dl Title: NEI 99-01 Rev 6 EAL Calculations 1 of 11j Purpose: The purpose of this appendix is to calculate radiation readings for Other Indications (Fuel Clad Barrier Potential Loss 5.A).a. Off-Gas Post Treatment Radiation Monitor D11K615 Procedure 64C1-OCB-004-1/2.

b. Fission Product Monitor D11K630 Procedure 64C1-OCB-005-1/2.

Defining Units that MathCAD will understand MWt := 10 6-W I Ci:.- *Bq and 2.7.10 Rad := 0.01iGy, Per FGR-11 page 219 IItCi := 10-6 Ci Rem := 0.01Sv Rem mRem :=-1000 Per FGR-11 page 219 cc := mL 1 eps := -sec 1 cpm :=-min The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used with the post treatment radiation monitors.

These isotopes are listed in 64CI-OCB-004-1/2 Page 56.Isotopes: "Ar-41I""Kr-85m""Kr-85""Kr-87""Kr-88""Kr-89""Kr-90""Xe-13 lm"'"Xe-133m""Xe- 133""Xe-135m""Xe-135""Xe-137" S"Xe-138" Core._Inventory

0 3.30E+03 3 .78E+02 6.92E+03 1 .32E+04 1 .86E+04 2.26E+04 3.03E+02 5.27E+04 1.58E+03 1 .89E+04 1 .09E+04 4.81E+04 4.52E+04 Ci MWt Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment 0 Title: NEI 99-01 Rev 6 EAL Calculations 2 of 11 CorePower

= 2818.MWt RCS inv := 9965 ft 3 F Release := 0.05 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.To determine the RCS inventory mass the water specific volume needs to found as follows: Ul: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4°F+/-+

532.0 °F TAvgu 1 := 2 462.2.OF U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 °F + 535.0 °F T Avgu 2 := 2= 480.35.°F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 URCS water := 0.0199161-

-Ibm According to http :/l/www. s piraxs arco. c om/res ources /st eam-tab les.asp the specific volume for 106Opsia and 480.35 F conditions is 0.0199161ft^3/Ibm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv 18.g RCS mass .- -2.27 x 1 i VRCS water Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 3 of 11 The concentration of isotopes in RCS is calculated as follows: FDEI300 .- gm -0.0 13-~2.38-104 gm This fraction was determined in Attachment C.This is the fraction that will be applied isotopes concentrations to adjust to RCS DEl 300 uCi/gm equivalent as described in NEI 99-01.CoreInventory-CorePower.

F_Release.

F_DEI3O00 Conc_DEI 3 0 0 RCS isotopes : RCS mass augment(Isotopes, Conc_DEJ300 RCS_isotopes)=

o: "Ar-41" 0E+000 i. "Kr-85m" 2.582E+001 2 "Kr-85" 2.958E+000

-3 "Kr-87" 5.415E+001 14 "Kr-88" 1.033E+002 S51 "Kr-89" 1.456E+002 6 "'Kr-90" 1.769E+002 7 "Xe-131m" 2.371E+000 8 "Xe-133m" 4.124E+002

9. "Xe-133" 1.236E+001 10, "Xe-135m" 1.479E+002 1i1 "Xe-135" 8.53E+001 12 "Xe-137" 3.764E+002

'13 "Xe-138" 3.537E+002 I-Ci gm

[Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 4 of 11 RCS_mass.

Conc_DEI 3 0 0 RCS isotopes Rateoff Gas*: 1lhr The gaseous activity corresponding to 300uCi/gm DEl would evolve from the RCS water to main steam. Assuming dilution in one hour of main steam flow and release through the offgas.augment(Isotopes, Rateoff Gas)o 1 0! "Ar-41" OE+000 1 "Kr-85m" 1.628E+006 2 "Kr-85" 1.865E+005 3 "Kr-87" 3.414E+006 4 "Kr-88" 6.512E+006 5 "Kr-89" 9.176E+006 6 "Kr-90" 1.115E+007 7 "Xe-131lm" 1.495E+005 8 "Xe-133m" 2.6E+007 9 "Xe-133" 7.795E+005 10 "Xe-135m" 9.324E+006 11 "Xe-135" 5.377E+006 12 "Xe-137" 2.373E+007 13 "Xe-138" 2.23E+007 IltCi sec Total-RateoffGas

= Rateoff -a =1.197 >< 108. tC-- sec Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 5 of 11 RadM~nDose Factors : (7.34E-4'8.25E-5 1 .26E-6 4.40E-4 1 .09E-3 9.44E-4 7.00E-4 1 .68E-6 I1.29E-5 1.3 8E-5 2.42E-4 1.33E-4 9.55E-5 ,6.16E-4, The radiation monitor Dose Factors were obtained from 64CI-OCB-004-1/2 Page 56.toRemo sec plCi~yr DoseRate := (RadMOnDoseFactors.RateoffGas))

augment(Isotopes, Dose Rate) =0 1 0 "Ar-41" 0E+000 1 "Kr-85m" 1.343E+002 2 "Kr-85" 2.35E-001 3 "Kr-87" 1.502E+003 4 "Kr-~88" 7.098E+003 5 "Kr-89" 8.662E+003 6 "Kr-90" 7.805E+i003 7 "Xe-131m" 2.511E-001 8 "Xe-133m" 3.354E+002 9 "Xe-133" 1.076E+001 10 "Xe-135m" 2.256E+003 11 "Xe- 135" 7.152E+002 12 "Xe-137" 2.266E+003 13 "Xe-138" 1.374E+004 mRem yr Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 6 of 11 Dose_RatetotaI

= DoseRate Summing all rows of the Dose_Rate matrix.Dose atetotal ; 4.452 x 0 ' em Calculating the ratio of this release to ODCM allowed release: LimitODCM

=500 m-- The ODOM limit is listed in ODCM Section 3.1.2 yr and in 64CI-OCB-004-1/2 Page 56.Factorallocation

= 0.225 The factor is listed in 64Cl-OCB-004-1/2 Page 56.Factorsafety

= 0.5 The factor is listed in 64C1-OCB-004-1/2 Page 56.DoseRate total RatioReleaseODCM LiiOC.Fatrioato=sft

-LimtODCM FactralloationFactorsft LRatioReleaseODCM

= 791.516J Calculating the detector response as follows: Total-RateoffGas

=1.197 x 108 C-- sec 14 cc Flowofa := 200cfm =9.439 x 10 -Ofgssee Factorefficiency

= 3.8.105 p liCi cc calculated above.The efficiency factor is approximately 3.ESEcps/uCi/cc per attached below e-mail.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 7 of 11 TotalRateoff Gas" Factoreicny Reading~cs DEI300 : Flowoffgas Reading~cs DEI300 =4.82 x 108.cp]According to S$16039 pg 21 and SX27520 pg 14 the range of this instrument is 1E-1 to 1E6 cps. The reading will exceed the range of this monitor, but a reading at the high end of the monitor range will still be indicative of fuel damage.Part B of FCEAL4 The fission product monitor for the containment air is DilP010 per SX18062 pg 32 and SX27520 pg 32. The iodine-noble gas sample panel is designed to extract a representative sample of gaseous and particulate effluent, transport the sample to: the iodine monitor where the iodine is retained on a filter for 1-131 gamma radiation detection and the gas sampler where it is continuously monitored for gross gamma and beta radioactivity (SX29455 section 6, S19207 section 7, and $30523 section 1-5). Since Iodine's and Noble-Gases are separately measured, only noble-gases will be evaluated in this attachment.

Also, noble-gases provide real time indication.

gal Leakage~wHP

= 3900---day gal Leakage~wLP

= 700 --day The High Purity Waste Stream leakage inside the DryWell is obtained from SMNH-93-029 page 6.The Low Purity Waste Stream leakage inside the DryWell is obtained from SMNH-93-029 page 6.Leakage~wtot

=Leakage~wHP

+ Leakage 0 w LP =4.6 x .0 3 ga-_ day DWVolu1 := 146010ft 3 SPVOlu 1:= 112900ft 3 The dry well free volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol_Containmentu1

= DW_Voluj 1 + SPVolu1 = 2.589 x 105.ft3 3The dry well free volume is listed in NL-06-1 637 Enclosure DW-V°Iu 2:= 146266ft3 1 page 47 table 34.

SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13--021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 8 of 11 SP_VoIu 2:= 109800ft3 Sne suppression pooi voiume Is iistea in INL-UO-Ilbi(Enclosure 1 page 47 table 34.Vol_Containmentu 2:= DWVolu 2+ SPVolu 2=2.561 >( 105.ft3 ContainmentVol Avg : VolContainm entul + Vol Containmentu 2 53= 2.575 x 105.ft3 2 ft 3 a~water STP :=0.016714-

-Ibm Specific volume at STP per ASME international Steam Tables page 174 1 0.958. gm PwaterSTP

"- teTPm The isotopes concentration in the containment (DryWell and Suppression Pool) after the 1st hour is calculated as follows: C_IsotopeScontainment

1 hr Leakage~w -tot" Pwaterona met STp 'Conc -DEI 3 0 0 RCSvo v isotopes i augment(Isotopes, CIsotopeScontainment)

=0 0 "Ar-41" 0 1 "Kr-85m"'

2.463"10-3 2 "Kr-85" 2.821"10-4 3 "Kr-87" 5.164"10-3 4 "Kr-88" 9.851"10-3 5 "Kr-89" 0.014 6 "Kr-90" 0.017 7 "Xe-131m" 2.261"10-4 8 "Xe-133m" 0.039 9 "Xe- 133" 1.179" 0-3 10 "Xe-135m" 0.014 12 "Xe-135" 0.13"0-3 13 "Xe-137" 0.034 mL Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 9 of 11 F Responsexel 3 3 :=~ 2.7.10 cpm mL F ResponseKr 8 5 := 2.29-10o8 cpm liCi mL The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number P0l0.The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number Polo.Since the D11P010 gas detector is G-M counter ($30523 pdf page 76 section 1-4) the Xe133 response factor will be applied to all isotopes of Xenon, while the Kr85 response factor will be applied to all isotopes of Krypton. Because GM will detect all gammas as long as they are above the threshold for the detector.Isotopes =K "Ar-4 1""Kr-85Sm""Kr-8 5""Kr-87""Kr-8 8""Kr-89""Kr-90""Xe-131lm""Xe-133m""Xe-133""Xe-135m""Xe- 135""Xe-137" S"Xe-138" 0 FResponseKr 8 5 F_ResponseKr 8 5 F_ResponseKr 8 5 F_ResponseKr 8 5 F_ReSPOnseCKr 8 5 F_ResponseKr 8 5 N FResponse

F_Responsexe1 3 3 FResponsexe1 3 3 FResponsexe 1 3 3 F_Responsexel 3 3 F_Responsexe1 3 3 I, FResponsexel 3 3 F Responsexe1 3 3)

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment ID Title: NEI 99-01 Rev 6 EAL Calculations 1l of 11 The response of the detector D11P010 is calculated as follows: the response factor for the detector is multiplied by the concentration of isotopes in the containment.

Responsep 0 1 0:= (FResponse.CIsotopeScontainment) augment(Isotopes, Responsep 0 1 0)=0 1 0 "'Ar-41" 0 1 "Kr-85m" 5.64"105 2 "Kr-85" 6.46"104 3 "Kr-87" 1.183.106 4 "Kr-88" 2.256" 106 5 "Kr-89" 3.179"106 6 "Kr-90" 3.862.106 -cpm 7 "Xe-131lm" 6.106"103 8 "Xe-133m" 1.062" 106 9 "Xe-133" 3.184"104 10 "Xe-135m" 3.808"105 11 "Xe-135" 2.196'105 12 "Xe-137" 9.692'105 13 "Xe-138" 9.108'105 Tot_Responsep 0 1 0:= 2Responsep 0 1 0=1.469 x This exceeds the range of the monitor 10 to 10^6 cpm. The reading at the high end of the monitor range will still be indicative of fuel damage.U2: Per SX29455 Table 6-1 the recorder for P010 is D11-R630.

Per SX27520 page 31 the recorder range is 10 -1 0^6 cpm.UI: Per $19207 Table 7-1 the recorder is MPL D11-R630.

Per SX18062 page 31 the recorded data sheet is A16462 sheet D11B. Per this data sheet the recorder range 10-10^6 cpm.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment D Title: NEI 99-01 Rev 6 EAL Calculations 11 of 11 From: Hostetter, Dwight A.Sent: Thursday, March 31, 2005 8:50 AM To: Wehrenberg, James A.

Subject:

Efficiency Factors Andy, Hatch Chemistry gave me the following efficiency factors for the instruments listed: Plant Service Water Monitor (Procedure 64C1-OCB-008)

U-i: 1.78 E+7 cps/uCi/ml U-2:1I.53E+7 cps/uCi/mI Liquid Radwaste Monitor (Procedure 64CI-OCB-009)

U-i: 3.136 E+6 cpsiuClIml U-2: 4.84 E+6 cps/uClfml Dwight Offgas Post-treatment .Monitor Response Factor From: Hodgins, S. R.Sent: Tuesday, April 26, 2005 2:01 PM To: Wehrenberg, James A.

Subject:

RE: Calibration set points Sorry for the incomplete information.

The number is alilthat is listed on our set point record. In 1994 we contracted Ultrapure Watel: Technology Inc to determine efficiency factors for our process radiation monitors.

We.do not verify these subsequent to the initial mock up determinations that they did then. We do 24 month source checks to verify that monitor response has not degraded appreciably., went to the Ultrapure report and the unit for the efficiency factor Is cps/(uci/cc).

---

Original Message-From: Hostetter, Dwight A.Sent: Friday, April 22, 2006 2:19 PM To: Hodgins, S. R.; Wehrenberg, James A.

Subject:

RE: Calibration set points Thanks Steve.Dwight From: Hodgins, S. R.Sent: Friday, April 22, 2005 2:17 PM To: Hostetter, Dwight A.

Subject:

RE: Calibration set points To my knowledge we do not determine an~eff fac for FPM NG monitors.

For most of our monitors, we correlate release rates to monitor response by comparing isotopic analysis of a process sample to monitor response.

The FPM noble gas samples we pull typically are clean (no activity).

Unit I Post Treat A and B =3.85E5: Unit 2 Post Treat A and B = 3.75E5.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 1 of 16 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors (ROS Barrier Loss Threshold 4.A). This attachment is some what repetitive of FC EAL3 (Attachment C) except of the DEI131 here is equal to TS instead of DEII31 of 300uCi/gin.Defining Units that MathCAD will understand MWt := 106-W Mega Watt Thermal is same as watts.Ci: .-*Bq 2.7.10 1 and Rad := 0.01Gy Per FGR-11 page 219:= 10-6 Ci The source terms for noble gases and iodine's are provided in NL-06-1637 Enclosure1 page 16 and are shown below.Isotopes : "Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 1m""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-13 8""I-131""1-132""1-13 3""1-13 4""I-135" Corelnventory

(3.30OE+03

" 3 .78E+02 6.92E+03 I1.32E+04 1 .86E+04 2.26E+04 3.03E+02 5 .27E+04 1 .58E+03 I1.89E+04 1.09E+04 4.81IE+04 4.5 2E+04 2.72E+04 3 .93E+04 5 .52E+04 6.05E+04 5.16E+04 Ci MWt 9 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 2 of 16 CorePower

= 2818-MWt RCS inv := 9965ft3 FRelease := 0.05 DWVolu1 := 146010ft3 The core power is provided in NL-06-1 637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1 637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases and halogens (i.e. Iodine's) is 0.05 per RG-1.183 page 13.The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.SP_Vo1u 1 := 112900ft3 The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.Vol_Containmentui

= DWVOl~uI + SPVolu1 2.589 x 10 5.ft 3 3 The dry well free volume is listed in NL-06-1 637 DW-V°Iu 2 := 146266ft3 Enclosure 1 page 47 table 34.SP_Volu 2 := 109800ft3 The suppression pool volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.Vol_Containmentu 2 := DW_Volu 2 + SPVolu 2 =2.56 1 x< 105.ft 3 VolContainmentul

+ Vol-Containmentu 2_3 ContainmentVol Avg := 2.575 x lO-.ft-2 To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 0 F TAvguI : = 462.2.OF U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 0 F T Avgu 2:= = 480.35.°F Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 3 of 16 The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 V°RCS water := 0.0199161-

-Ibm According to htt p://www.s piraxs arco. com/resources/steam-tab les.asp the specific volume for 1060psia and.480.35 F conditions is 0.0199161ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv8 RCS mass >- --2.27 x 10 8gin"-RCS water The concentration of isotopes in RCS is calculated as follows: CoreInventory.

CorePower.

F_Release COnCRCS isotopes : RCS mass augment(lsotopes, COnCRCS isotopes)0 1 0 "Kr-83m" 2.049'10-3 1 "Kr-85" 2.347" 0-4 2 "Kr-85m" 4.296"10-3 3 "Kr-87" 8.195"10-3 4 "Kr-88" 0.012 5 " T Kr-89" 0.014 6 "Xe-131lm" 1.881"10-7 "Xe-133" 0.033 8 'Xe-133m" 9.809"10-9 "Xe-135" 0.012 10 "Xe-135m" 6,767'10-3 11 "Xe-137" 0.03 12 "Xe-138" 0.028 13 "I-131" 0.017 14 "1-132" 0.024 15 "1-133" 0.034 16 "I-134" 0.038 17 "I-135" 0.032 Ci gin tPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEl 99-0 1 Rev 6 FAL Calculations 4 of 16 The DEl for the RCS is calculated in Attachment C and was found to be 2.37E4uCi/gm.

gm___ -5 F DEJ TS.- g -8.439xx10

_ _ 2"714 I-tCi gm Per NEI 99-01 RCS Barrier EAL4 RCS should be of TS limit. According to HNP Tech Specs section 3.4.6 the DEIl31 limit is 2uCi/gm Amount of isotopes in the drywell is found as follows: In this calculation the core inventory is multiplied by core power, containment release fraction, and the DEl of 300uCi/gm fraction.

The amount of total curies released are distributed throughout the drywell and suppression pool. The concentration then is multiplied by the U1 DryWell volume. This way we have the total curies inside the DryWell only. The DryWell for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.Corelnventory.

CorePower.

F_Release.

F_DEl_TS-D WVolu 2 Isotopes_in_DW

augment(Isotopes, Isotopes inDW) =ContainmentVol Avg 0 h 1_ _"Kr-83m" 2.229E+001"Kr-85" 2,553E+000"Kr-85m" 4.674E+001"Kr-87" 8.916E+001"Kr-88" 1.256E+002"Kr-89" 1.526E+002"Xe-131m" 2.047E+000"Xe-133" 3.56E+002"Xe- 133mi" 1,067E+001"Xe-135" 1.277E+002"Xe-135m" 7.362E+001.Ci"Xe-137" 3.249E+002"Xe-138" 3.053E+002 "1-131" 1,837E+002 "1-132" 2.654E+002 "1-133" 3.728E+002"I-134" 4.086E+002"I-135" 3.485E+002 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 5 of 16 The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air density for pressure 44.7psia and temperature of 343F (NL-06-1637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

DW_Volu 2 = 4.142 x 109-mL Psteam'.-

-1.53 x10 , ft 3 mL 10.4690 --lb According to http://www.spiraxsarco.

corn/resources/steam-tab les.asp the specific volume for 44.7psia and 343 F conditions is 10.4690ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below. The drywell pressure and temperature were obtained from NL-06-1637 Enclosure 1 table 11 and table 12.The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.ergs -cm 2 3 0 S eVg h 100 ergs/(g .rad)2-6 ergs cm s 1.6.10 --.-~ 30 Rad_hr or~GeometriCfactor-Sv 100 ergs g. rad-6 1.6-10 x 3600 -5= 5.76 x 10 100 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEl 99-01 Rev 6 EAL Calculations 6 of 16 Defining Units that MathCAD can process.q:=1.60219.10-19coul eV := qe.volt MeV := 106.eV Fundamental charge of electron page 6 of Fundamental of Nuclear Science and Engineering.

defines the unit of charge.cc := mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

3 .00E-0 1 4.00E-0 1 5.00E-01 6.00E-0I1 8.00E-01 1.00E+00 1 .50E+00 2.00E+00 3.00E+00 4.00E+00 SMeVSv: r6.11E+02" 6.6 8E+02 2.41IE+02 1 .84E+03 3.18E+03 3 .30E+03 4.92E+03 2.67E+03 2.24E+03 0.00E+00 MeV cc.SeC Iabsair : 0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.0238 0.02 11194 2 cm gm Heightcyl

= 1.265.10J cm Calculated in Attachment C Calculated in Attachment C Radiationfield

SHighty, 1 (P.labsairlj 0 1 2 3 4 5 6 7 8 9 0 0.324 0.364 0.132 1.003 1.693 1.702 2.32 1.171 0.871 0 Rad hr The unit conversion was handled by MathCAD internally.

This unit conversion was explained above in this attachment.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NE! 99-01 Rev 6 EAL Calculations 7 of 16 SMRadiation field := £Radiationfil

= Rad-iedhr Radiationleak

= SUMRadiationfield

= 9.581- a-hr According to the $43177 Operator Manual Figure 1-4 the model number 877-1 detector has 1 to 1 ratio of Radiation present to radiation measures or shown. Therefore the detector would read 9.58 1 R/hr. However the detector is located in the area where some radiation is present during normal operations.

The following calculation will determine this contribution.

Per H16241 and H26417 the N003 detector location is at elevation of 156ft and approximately 27 ft from the center of RPV. According to (BH2-M-V999-0047 Table 2 and 3) the Solenoid Valves B21-AOV-F013 are located in approximately at elevation of 153.6 and radius between 17 to 22 feet.The radiation field from RPV at elevation 153.6 for Unit 1 and Unit 2 is listed below (BH2-M-V999-0047 Table 2 and 3).Solenoid Valves Solenoid Valves for U1 Rad/hr for U2 Rdh 1B21-AOV-F013A 6.44 2B21-AOV-F013A 2.46__1B21-AOV-F013B 4.41 2B21-AOV-F0138 9.96__1B21-AOV-FO13C 12.3 2B21-AOV-F013C 9.96__1B21-AOV-F013D 13 2B21-AOV-F013D 24 1B21-AOV-F013E 12.3 2B21-AOV-F013E 2.46 1B21-AOV-F013F 12.3 2B21-AOV-F013F 9.96 1B21-AOV-F013G 12.3 2B21-AOV-F013G 9.96 I1 B21-AOV-F013H 4.41 2B21-AOV-F013H 2.46 1B21-AOV-F013J 4.41 2B21-AOV-F013K 9.95 1821-AOV-F013K 12.3 2B21-AOV-F013L 9.96 I1B21-AOV-F013L 12.3 2B21-AOV-F013M 9.96 Ave rage = 9.7 _ ______ 7.2 9.7 + 7.2 Rad _85.Rad Tkn naeaebtentouis Rdainpv 2 hr hr Plant: H-NP UI & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 8 of 16 The radiation field from Main Steam Lines (BH2-M-V999-0047 Table 6 and 6).Solenoid Valves Solenoid Valves for Ul ,Rad/hr for U2 Rad/hr 1B21-AOV-F013A 17.57 2B21-AOV-F013A 17.39 1B21-AOV-F013B 17.28' 2B21-AOV-F013B 17.28 1B21-AOV-F013C 17.08 2B21-AOV-F013C 17.08 1B21-AOV-F013D 17.01 2B21-AOV-F013D 17.39 1B21-AOV-F013E 17.16 2B21-AOV-F013E 17.28 1B21-AOV-F013F 16.13 2B21-AOV-F013F 17.08 1B21-AOV-F013G 16.13 2B21-AOV-F013G 15.43 1B21-AOV-F013H 16.62 2B21-AOV-F013H 17.28 1B21-AOV-FO13J 16.33 2B21-AOV-FO13K 15.43 1B21-AOV-F013K 16 2B21-AOV-F013L 14.78 1B21-AOV-F013L 13.4 2B21-AOV-F013M 14.78 Average = 16.4 _________

16.5__6.4__+ 6. Rad 16.45- a Taking an average between two units.RadiationMsL 2.h-h The radiation field from Recirculation Lines ((BH2-M-V999-0047 Table 9 and 10).Solenoid Valves Solenoid Valves for Ut Rad/hr for U2 Rad/hr 1B21-AOV-F013A 7.5 2B21-AOV-F013A 11.97 1B21-AOV-F013B 8.23 2B21-AOV-FO13B 6.23 1B21-AOV-F013C 6.19 2B21-AOV-F013C 6.85 1B21-AOV-F013D 5.2 2B21-AOV-F013D 10.08 1B21-AOV-F013E 5.89 2B21-AOV-F013E 15.22 1B21-AOV-F013F 5.15 2821-AOV-F013F 6.19 1B21-AOV-F013G 5.89 2B21-AOV-F013G 6.85 1B21-AOV-F013H 7.5 2B21-AOV-F013H 15.96 1B21-AOV-F013J 7.5 2B21-AOV-FO13 K 6.85 1B21-AOV-F013K 6.22 2B21-AOV-F013L 7.8 1B21-AOV-F013L 6.22 2821-AOV-F013M 7.8 Average = 6.5 9.3 6.5 + 9.3 Rad Rad RadiationRL, .- -7..2 hr hr Taking an average between two units.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NE! 99-01 Rev 6 EAL Calculations 9 of 16 2 RadiationtotaI

= Radiationleak

+ Radiation~ppV

+ RadiationMsL

+ RadiationRL, = 0.763 --.-R ea K* s2 hr Rad Radiationtotai 42.381 hr This can be rounded to 40R/hr. The range of this instrument is 1-10^7 R/hr (established in attachment C).GRODEC Input 1 18 60,2.229e+

1,0 61 ,2.553e+0,0 62,4.674e+1

,0 63,8.916e+1

,0 64,1 .256e+2,0 65,1 .526e+2,0 141,1.837e+2,0 142,2.654e+2,0 143, 3.728e+2, 0 144,4.086e+2, 0 145,3.485e+2,0 146,2.047e+-0,0 147,3.560e+2,0 148, 1. 067e+- 1,0 149,1 .277e+2,0 150, 7. 362e+ 1,0 151,3.249e+2,0 152, 3. 053e+2,0 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E l Title: NE! 99-01 Rev 6 EAL 10oaf 16 Calculations OUTPUT OF GRODEC CALCULATION OPERATING IN MODE DATA FILE NAME:RCSEAL4 1 INPUT DATA LISTING ISOTOPE INITIAL ADDITION ACTIVITY(CI)

RATE(CI/HR)

KR-83M KR-85 KR-85M KR-87 KR-88 KR-89 1-13 1 1-132 1-133 I-134 1-135 XE-131M XE-133 XE-1 33M XE-135 XE-i135M XE-137 XE-i138 2.229E+01 2.553E+-00 4.674E+01 8.916E+01 1 .256E+02 1.526E+02 1 .837E+02 2. 654E+02 3.728E+02 4.086E+-02 3.485E+02 2. 047E+00 3.560E+02 1 .067E+01 1 .277E+02 7.362E+01 3. 249 E +02 3.053E+02 0.000E+00 0. 000E+00 0.000E+00 0.O00E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0. 000E+00 0.000E+00 0.000OE+00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

=1.000000 HOURS TOTAL ACTIVITY = 2155.917000 CURI ES ISOTOPE ISOTOPE NUMBER NAME INITIAL ADDITION ACT(CI) RATE(CI/HR)

ACTIVITY (CI)60 61 62 KR-83M KR-85 KR-85M 2.229E+01 2.553E+00 4. 674E+01 0.000E+00 1.536E+01 0.O0OE+00 2.553E+00 0.000E+00 3.993E+01 Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL Calculations 11 of 16 63 KR-87 8.916E+01 0.OOOE+00 5.160E+01 64 KR-88 1.256E+02 0.000E+00 9.806E+01 65 KR-89 1.526E+02 0.000E+00 3.199E-04 75 RB-87 0.000E+00 0.000E+i00 1.131E-13 76 RB-88 0.000E+00 0.O00E+00 9.609E+01 77 RB-89 0.000E+00 0.O00E+00 2.669E+00 84 SR-89 0.000E+00 0.000E+00 5.850E-03 141 1-131 1.837E+i02 0.000E+00 1.830E+02 142 1-132 2.654E+02 0.000E+00 1.953E+02 143 1-133 3.728E+02 0.OO0E+00 3.603E+02 144 1-134 4.086E+02 0.000E+00 1.837E+02 145 1-135 3.485E+02 0.OOOE+00 3.142E+02 146 XE-131M 2.047E+00 O.000E+00 2.490E+O0 147 XE-133 3.560E+02 0.000E+00 3.561E+02 148 XE-133M 1.067E+01 0.00OE+00 1.053E+01 149 XE-135 1.277E+02 0.000E+00 1.444E+-02 150 XE-135M 7.362E+01 0.000E+00 5.122E+00 151 XE-137 3.249E+02 0.O00E+00 7.612E-03 152 XE-138 3.053E+02 0.000E+00 2.838E+01 161 CS-135 0.O00E+00 0.O00E+00 3.603E-09 163 CS-137 0.000E+00 0.000E+00 8.030E-05 164 CS-i138 0.000E+00 0.000E+00 6.615E+-01 171 BA-137M 0.000E+00 0.000E+00 8.029E-05 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME ACTIVITY (CI)RB-87 1.130985E-13 CS-I135 3.603327E-09 WHAT IS THE SOURCE VOLUME (CC)4.142000E+-09 WHAT IS THE SOURCE DENSITY (GM/CC)1 .530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.THE NUMBER OF ENERGY GROUPS INPUT: 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 3.0O0000E-01 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 12 of 16 Calculations 2 3 4 5 6 7 8 9 10 4.000000E-01 5.000000E-01 6.000000E-01 8.000000E-01 1 .000000E+00 1 .500000E+00

2. 000000E+00 3.000000E+00 4.000000E+00 MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO.UC/GM SOURCE STRENGTH UC/CC Cl KR 83M 45 KR 85 47 KR 85M 46 KR 87 48 KR 88 49 KR 89 50 KR 89 51 RB 88 67 RB 89 68 SR 89 21 I 131 4 I 132 5 I 133 6 I 134 7 I 135 8 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 137 35 CS 138 36 BAI137M 78 2.423253E+00 3.707578E-03 1 .535679E+01I 4.029027E-01 6.164411 E-04 2. 553299E+-00 6.300656E+00 9.640004E-03

3. 992890E+01I 8.141909E+00 1.245712E-02 5.159739E+01 1 .547392E+01 2.367509E-02

9. 806223E+01 5.047167E-05 7.722166E-08 3.198521 E-04 5.047167E-05 7.722166E-08 3.198521 E-04 1.516330E+01 2.319985E-02 9.609376E+01 4.211933E-01 6.444258E-04 2.669212E+00 9.23081 6E-04 1.41231 5E-06 5. 849808E-03 2.888350E+'01 4.419176E-02 1.830423E+02 3.081971E+01 4.715415E-02 1.953125E+02 5.685235E+01 8.698410OE-02 3.602881 E+02 2.898289E+01 4.434381E-02 1 .836721E+02 4.957314E+01 7.584691E-02 3.141579E+02 3.929409E-01 6.011 996E-04 2.4901 69E+00 5.619389E+01 8.597666E-02 3.561153E+02 1.66231 8E+00 2.543346E-03 1 .053454E+01 2.278258E+01 3.485735E-02 I1.443792E+02 8.082252E-01 I1.236585E-03 5.121 933E+00 1.2011 59E-03 1. 837774E-06 7.61 2058E-03 4.477746E+00 6.850951E-03 2.837664E+O1 1 .267052E-05 I .938590E-08 8.029641E-05 1.043846E+01 1.597085E-02 6.615126E+01 1 .266997E-05 1 .938505E-08 8.029286E-05 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)1 2 3.000000E-01 4.000000E-01

6. 111697E+02 6.676899E+02 8.438217E+12 6.913930E+12 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 13 of 16 SCalculations 3 4 5 6 7 8 9 10 5.000000E-01 6.000000E-01 8.000000E-01 1 .OQ0OO0E+00 1 .500000E+00 2.000000E+00 3.000000OE+00

4. 000000E+-00 2.414372E+02 1.844615E+03 3.175361 E+03 3.304496E+i03 4.91 8952E-+03 2.666496E+03 2.235936E+03

0. 000000E+-00 2.000066E+i12 1 .273399E+

13 1.644043E+

13 1.368722E+13

1. 358287E+ 13 5.5223 14E+ 12 3.087083E+-12 0.000000OE+00 SPlant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment E Title: NEI 99-01 Rev 6 EAL 14 of 16 Calculations

[Yes are here: h'4rSe WTs S Srib Saturaed Water Region Sub Saturated Water Region -Steam Table At an~y pressure, water belew its saturatesn temperature is said It be ir a sub Cronwa ea saturated 1t atrsasph~ers base sa turaton temperature at 180°C, and so water Irelowr the temperature is sub saturated em Learn mare abaut steam in aorr tutmriatl-What is Steam?eaueSat youre ptitae tor these steam tale~Note: -You cannot us commas (,) as decimal points.Trannr that,,. Please use perods (.)tartat mstrt Example: 1.02 not 1,02Single Vatue Table Temperature F EJ Temperature

-: ...... +::[Speciltc Votume of Watw+(v) : :::::: s:::: I,

[Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 AttachmentiteNE9-0Rv6EACaclios1 Ef1 I 1 5 o nIsefo prf 0.5 AI~set Us ., Products & tamtes *, teatmoles

& Awkatons -TraegM You me here: I, th I I t Sseerp ae SlIeaseReOel.

Superheated Steam Region -Steam Table The supeseied sleam region diepicts steam at a tempr~eature higher than Its satturataons temperatu re Shtould saturated steam be heated at cntn presue tu its tnperature wI rise. preoducktg superheater steam Learn more about steam is our tuora -Whi m tem Set your for these steam tabltes Note: -You cantnot sue commas (,} as dlecimal points.Please use petlods (.)Example: 1.02 not 1,02 Feaur Er L e t hat 0 Restore and S.serteat Teerperatore

• StirgIe Value Table yu absolute Pressur~e Supedtsat Temperature liii Saturation Temperaturesuperheat Spedific Enhls of Water (ho)SpcfcEnthalpy of Fvaporatlorr (hl Specific Enthalpy of Superheated Steam (h)O~eomst of Steam f[I JArg Specifi Volume ot Steam (v)

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021AtahetE Title: NEI 99-01 Rev 6 EAL Calculations AtaC6of161 Table 6.1.-I. Units used to express S tabulated in relation to the units used to express _________S Point source Line source Surface. Volume co as a factor of ~ o" S source source S 0 t.S Sv7 p-photons 7-photons V-photons y-photons y-photons cm 2 s s cms cm~s cm~s 7-photons mCi mCi mCi mCi ,70 disintegrations

.p-photons cm, s (millicuries) 7m cm 2 cm 2 7O s mCi disintegration MeY. MeV MeV MeV MeV cm~s s cms cm~s cmS s MeY ~ mCi mCi mCi 37 0 cm~s cm cmz cm* disintegrations, n--photons MeV smCi "disintegration photon MeV mCi mCi mCi MeV cms mCi cm mncm 2 z J msmCi EMeV -eg m3600-rads 7-photons 7-photons y-photons 7-photons 7-hoo M.eV0 2hg h s cms cm~s cm8 s 100 ergs/(g.rad)

Me MY .6l0~ ergs cm,3600 s rads MeV MeV MeVMV 1..1- ha h s cmns cm~s cm 3 s 100 ergsftg ,rad)EMeV _erscm"- s R 7-photons v-photons 7-photons 7-photons y-pho.ton

-.1a ' h60-hs cms cmas cm~s 87.7 ergs/(gR)1 6 06ergs -cm 2.00s R MeV MeV MeV MeV 1.e.1 &-eee.a---

h hs cmns cross cm~s 87.7 ergs! (gR)R mCi mCi mCi Rcm 2-- mCi cmct cm 3 4 ' R mg-equiv.

Ra mg-equiv.

Ra mg-eqluiv.

Ra .Rcm 2"--h- rag-equiv.

Ra cm cm 1 cm 3 h 8. ra g-equiv. Ra Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations I of 6 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Fission Product Monitoring sensors (RCS Barrier Loss Threshold 5.A). The normal RCS liquid activity is shown in FSAR Table 11.3-1. Noble gases are not typically retained in the RCS water, but are continuously released via offgas system. Offgas releases are shown in FSAR Table 11.3-1. Assuming 2% of the halogens are volatile per NUREG-0016 Table 2-4. According to the table the ratio of concentration in reactor steam to the concentration in reactor water for halogens is 0.015 or 1.5% we assumed a higher value of 2%. It is assumed that 100% of noble gases leaves the solution.

All other nuclides 0.1% are assumed to be airborne per table 2-4 NUREG-0016.

Fission Product Monitor D11K630 Procedure 64CI-OCB-005-1/2.

Defining Units that MathCAD will understand MWt:= 106.W Ci:= .Bq 2.7.10[LCi:= I0-6 Ci Per FGR-11 page 219 I cpm :=rai RCS inv := 9965ff3 The RCS inventory is page 47 Table 34.provided in NL-06-1637 Enclosure 1 DW Volu1: 146010ft 3 SP_Vo1u 1:= 112900ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.Vol Containmentu1

= DW_Volu1 + SPVolu1i 2.589 x 105.ft3 DWVolu 2 := 146266ft 3 SP_Vo1u 2:=109800ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 2 of6 Vol_Containmentu 2:=DW Volu 2+ SP_Volu 2= 2.56 1 x 105.ft 3 Containment Vol Avg : Vol_Containmentul

+ VolContainmentu 2 3= 2.575 x 105.ft3 2 To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 0 F TAvgu 1 := = 462.2.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 °F TAvgu 2 := -480.35. F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 Acc"-'RCS water := 0.0199161

--Ibm http480 web of tU atta The RCS mass is calculated as follows: RCS inv 108.i RCS mass .- -2.27 x 0 i'0 RCS water*ording to:Ilwww. spiraxsarco.

com/resources/steam-tab asp the specific volume for 1060psia and.35 F conditions is 0.0199161ftA3/Ibm.

This)site was validated in Attachment H. The copy he webpage and the provided information is ched to this attachment below.RCS mass = 5.003 x 105.Ibm Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 3 of 6 The concentration of isotopes in the containment is calculated as follows and using MathCAD internal conversion converted into cpm via fission response factor : Now we need to figure out noble gas concentration in the containment as follows: The noble gas isotopes Release is listed in FSAR Table 11.1-1: "Kr-83m""Kr-85m""Kr-8 5""Kr-87""Kr-8 8""Kr-89""Kr-9 0""Kr-9 1""Kr-92""Kr-93""Kr-94""Kr-95""Kr-97""Xe-13 lm""Xe-133m""Xe-133""Xe-135m""Xe-135""Xe-137""Xe- 138""Xe-139""Xe-140""Xe-141""Xe- 142""Xe- 143""Xe- 144" ReleaSeNoble

(3.40E+03" 6.10E+03 10 2.00E+04 2.00E+04 1 .30E+05 2.80E+05 3 .30E+05 3.30E+05 9.90E+04 2.30E+04 2.10E+03 1 .40E+01 1.50E+01 2.90E+02 8.20E+03 2.60E+04 2.20E+04 1 .50E+05 8.90E+04 2.80E+05 3.00E+05 2.40E+05 7.30E+04 1 .20E+04 5.60OE+02 IlCi s I The detector responses listed below were applied as follows: the Xe133 response was applied to all isotopes of xenon, while the Kr85 response factor was applied to all isotopes of krypton.Because GM will detect all gammas as long as they are above the threshold for the detector.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment 6Fl Title: NEI 99-01 Rev 6 EAL Calculations 4 of 6 FResponsexe1 3 3 := 2.7-107 -p mL FResponseKr 8 5 : 2.29. 108.p hmL NolIsotop 0"Kr-83m""Kr-85m""Kr-8 5""Kr-87""Kr-88""Kr-89""Kr-90""Kr-91""Kr-92""Kr-93""Kr-94""Kr-95""Kr-97""Xe-131mi""Xe-133m""Xe- 133'"Xe-135m""Xe- 135""Xe- 137""Xe- 138""Xe- 139""Xe- 140""Xe- 141""Xe- 142""Xe- 143""Xe- 144" The response factor was obtained from SX18062 pg32 and SX27520 pg 32 for equipment tag number P0l0.The response factor was obtained from SX1 8062 pg32 and SX27520 pg 32 for equipment tag number P010.FResponseKr 8 5 FResponse Kr85 FResponseKr 8 5 FResponseKr85 FResponseKr 8 5 F ResponseKr 8 5 FResponseKr 8 5 FResponseKr 8 5 FResponseKr 8 5 F ResponseKr 8 5 F ResponseKr 8 5 F ResponseKr 8 5 FResponseKr 8 5 Responsefactor

F-Responsexel 3 3 F-Rep~nexe33 F-Responsexei 33 F-Rep~nsxe 33 FResponsexe1 3 3 F-Responsexe1 3 3 FRspos ee 33 F-Responsexe 133 F-Rep~nexe133 FResponsexel133 FResponsexe1 3 3 F-Responsexe l 3 F-Responsexe 1 3 3 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 5 of 6 FlOWcoreUl

= 78.5.106 hrm 7716 Ibm Fl°WcoreU2

= "7 " hr-The flows can be found in GE-NE-O000-0003-0634-01 Figures Ia & lb.Flowore~ + FowcoeU07 Ibm FlOWav :=F°cr~ lw~e2=7.775 x 10 ag2 hr The concentration of noble gases isotopes in the containment and activity is found as follows. In this equation RCS mass is divided by RCS flow rate gives time that can be multiplied by the release rate of noble gases. The resulting amount of isotopes is divided by the containment volume and multiplied by the detector response factor.iFe v( RCmssl,owavg .ReleaSeNobleRePnfao]

A~tV~tnole LContainmentVolAvg 0 "Kr-83m" 2.474"103 1 "Kr-85m" 4.439'103 2 "Kr-85" 7.276 3 "Kr-87" 1.455"104 14 "Kr-88" 1.455'104 5: "Kr-89" 9.459"104 61 "Kr-90" 2.037'105~

7 "Kr-91" 2.401"105 i8 "Kr-92" 2.401"105~

9 "Kr-93" 7.204" 104.10 "Kr-94" 1.674"104 12 "Kr-97" 10.187 1i3 "Xe-131m" 1.287 14 "Xe-133m" 24.879 15 "Xe-133" 703.479 augment(Noblelsotop, ActivitYnoble)

=*cpm 16"Xe-135m" 2.231"103 1=1 _____

Piant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment F Title: NEI 99-01 Rev 6 EAL Calculations 6 of 6 17"Xe- 135" l.887" 10: 18 "Xe-137" 1.287' 04 19 "Xe-138" 7.635"103 20 "Xe- 139" 2.402"104 21 "Xe-140" 2.574" 104 22 "Xe-141" 2.059"104 23~ "Xe-142" 6.263" 103"Xe-143" 1.029"103 25 "Xe- 144" 48.042 ActivitYnoble~tot

= Ac'tivitYnoble

=1.008 x 10 6.cpm~er SX18062 page 34 the monitor K630 range is 10 to 10A6 cpm. This reading of 1E6 cpm ~tthe upper limit of the detector scale. This reading may be difficult to distinguish from!he RCS Barrier Potential Loss 5.A in attachment D.RP5: No radiation monitors capable of indicating a potential loss of the RCS dentified.

we/

lPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G]Title: NEI 99-01 Rev 6 EAL Calculations 1 of 121 Purpose: The purpose of this appendix is to calculate radiation readings for Drywell Radiation Monitoring sensors ( Primary Containment Barrier Potential Loss Threshold 4.A).DWRRM is D11K621 NMP-EP-110-GL02 pg 66.Defining Units that MathCAD will understand MWt := 106.W Mega Watt Thermal is same as watts.Ci :- *Bq 2.7.10 and Rad := 0.01Gy Per FGR-11 page 219 IJ.Ci := l0-6 Ci The source terms for noble gases and iodine's are provided in NL-06-1 637 and are shown below. We will only consider the isotopes that have release fraction above 0 as described in RG1.183 Table 1 and Table 5."1-129""I-130""I-131""I-132""I-133""I-134""1-135""I- 136""I-137""1-138""Kr-83m""Kr-8 5""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m" 1.23 E-03 I1.05E+03 2.72E+04 3 .93E+04 5,.52E+04 6.05E+04 5.1 6E+04 2.45E+04 2.39E+04 1.18E+04 3 .30E+03 3 .78E+02 6.92E+03 I .32E+04 1 .86E+04 2.26E+04 3 .03E+02 5 .27E+04 1 .58E+03 Isotopes : Corelnventory

Ci M tt Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 2 of 12"Xe-135m""Xe-137""Xe- 138""Rb-86""Rb-88""Rb-89""Rb-90""Cs- 134""Cs-134m""Cs-135""Cs- 136""Cs-137""Cs- 13 8""Cs- 139" 1 .09E+04 4.81 E+04 4.52E+I04 7.06E+01 1 .89E+04 2.42E+04 2.34E+04 6.83E+03 1 .65E+03 2.3 5E-02 2.18E+03 4.14E+03 5.02E+04 4.75E+04)I CorePower

= 2818. MWt The core power is Enclosure 1.provided in NL-06-1637 page 14 RCS inv := 9965ft 3 FRelease := 0.05 DWVoluI := 146010ft 3 SP_Vo1u 1:= 112900ft 3 The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.The release fraction into the containment for noble gases, halogens (i.e. Iodine's), and Alkali Metals is 0.05 per RG-1.183 page 13. GAP Release.The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.VolContainmentul

= DWVoIuI + SPVolu1 2.589 x 105.ft DW Volu 2 := 146266ft3 SP_Volu 2:= 109800ft 3 The dry well free volume is listed in NL-06-1 637 Enclosure 1 page 47 table 34.The suppression pool volume is listed in NL-06-1637 Enclosure 1 page 47 table 34.

Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 3 of 12 VolContainmentu 2:= DW Volu 2+ SPVolu 2=2.561 x 105.ft3 Containment_VolAvg

VolContainmentul

+ VolContainmentu 2 2.7 10ft 2 We find the amount of curies in the drywell as follows: FCore := 20%The 20% is the core fraction gap release as described in NEI 99-01 for the Containment Barrier.FRelease = 0.05 CorePower

= 2.8 18 x 10 3.MWt The concentration then is multiplied by the UI DryWell volume. This way we have the total curies inside the DryWell only. The DryVA~ll for the unit 2 was used because it is larger volume thus giving us smaller reading to initiate the EAL.DW_Volu 2 = 1.463 x 105.ft3 Containment Vol Av .7 0.t Coreinventory.

CorePower.F_Core.F_Release.

DWVolu 2 Isotopes in DW : Containment Vol Avg 0 1 0 "1-129' 1.969E-002 1 "I-130" 1.681E+004 2 "1-131" 4.354E+005 3 "1-132" 6.291E+005 4 "I-133" 8.836E+005 5 "I-134" 9.685E+005 6 "1-135" 8.26E+005 7 "1-136" 3.922E+005 8"I-137" 3.826E+005 B "1-137" 3.826E+005 I

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment~l:NI9-1Rv6ELCluain G] 1 FF r 9"I-138" 1.889E+005 augment(Isotopes, Isotopes in DW)10 "Kr-83m" 5.283E+004 11 "Kr-85" 6.051E+003 12: "Kr-85m" 1.108E+005 13 "Kr-87" 2.113E+005 P14 "Kr-88" 2.977E+005 15 "Kr-89" 3.618E+005 16 "Xe-131m" 4.85E+003 17, "Xe-133" 8.436E+005 18 "Xe-133m" 2.529E+004 19 "Xe-135" 3,025E+005 20 "Xe-135m" 1,745E+005 21 "Xe-137" 7.7E+005 22 "Xe-138" 7.235E+005 23 "Rb-86" 1.13E+003.24 "Rb-88" 3.025E+005 25 "Rb-89" 3.874E+005 26 "Rb-90" 3.746E+005 27- "Cs-134" 1.093E+005 28 "Cs-134m" 2.641E+004 29 "Cs-135" 3.762E-001 30 "Cs-136" 3.49E+004 31 "Cs-137" 6.627E+004 32 "Cs-138" 8.036E+005 33 "Cs-139" 7.604E+005.Ci The activities in the above table were entered into the GRODEC computer program (CALC F-86-03) to convert the activity to specific energy groups which are used to estimate the detector response.

The volume of 4.142E9cc and steam/air density for pressure 44.7psia and temperature of 343F (NL-06-1637 Tables 11 &12) is 1.53E-3gm/cc was used in GRODEC. Note GRODEC was installed on a Computer DELL SN#CYC7LS1 that was running Windows XP. To verify the program proper operation nine test cases were executed and output results were matched to the verification files listed on pages G1-G26 CALC F-86-03. The GRODEC input and output files can be found in GRODEC section of this calculation.

The isotopes of 1-129, 1-136, 1-137, and 1-138 are excluded from GRODEC because they are not supported, i.e. these isotopes are not part of CALC F-86-03 (pdf pages 105-112) and thus have no way of making appropriate entrees.

Plant: HNP U1 & U2 SNC CALCULATION SMNH.-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 5 of 12 The GRODEC results are manipulated below. The GRODEC input and output text files are listed below. The air absorbed dose is in table 1.3-2 on page 13 of Engineering Compendium of Radiation Shielding Volume 1. Also on page 366 Table 6.1-1 shows how to convert the flux into air absorbed dose in R/hr. The copy of this table was attached to this appendix.

The calculation for the dose is shown below.1.6.0-6ergs 1.5.-V-cm 2 s* -3600 gT" " 100 ergs/(g .rad)2 161- 6 ergs cm ...s 1-.6.1-m'0°°° Rad-hr Geometricfactor.

Sv 100 ergs gr ad or-6 1.6.10- x3600 -5=5.76 x 10 100 Defining Units that MathCAD can process.q =1.60219.10-1 9-coul eV := qe-volt MeV := 106.eV Fundamental charge of electron page 6 of Fundamental of Nuclear Science and Engineering.

defines the unit of charge.co := mL The Source Strength for various energy groups was obtained from the GRODEC results.Energygrp

"3.00E-01

" 4.00E-0 1 5.00E-01I 6.00E-01 8.00E-01 1.00E+00 1.50E+00 2.00E+00 3 .00E+00 4.00E+O0**MeV Sv: r1.5 1 .64E+06 7.81 E+05 4.46E+06 9.75E+06 8. 12E+06 I1.50E+i07 6.42E+06 6.72E+06 5 .22E+00 MeV cc sec lx~absair

0.0296 0.0297 0.0296 0.0289 0.028 0.0256 0.023 8 0.02 11 0,.0194)2 cm gm Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 6 of 12 3 HeightcyI

= 1.265-103cm qo =.505 Calculated in Attachment C Calculated in Attachment C Radiationfield

/Sv~,.Hehtcy 1~P4 saii)0 1 2 3 4 5 6 7 8 9 0 801. 194 894.343 427.342 2.432"103 5.191"103 4.189" 103I 7.075" 103 2.815"103 2.612"103 1.866" 10-3 The unit conversion was handled by Rad MathCAD internally.

hr This unit conversion was explained above in this attachment.

SMRadiationf ied := ~'-"Radiationfil

= 2.644 x10Ra-c ie dhr~ccording to the $43 177 Operator Manual Figure 1-4 the model number 877-1 detector~is 1 to 1 ratio of Radiation present to radiation measures or shown. Therefore the~itector would read 2.644E4 R/hr. The range of this instrument is 1-10^7 R/hr established in attachment C).

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 7 of 12 GRODEC Input 1 30 60,5 .283e+4, 0 61 ,6.051e+3,0 62,1 .108e+5,0 63,2.11 3e+5,0 64,2.977e+-5,0 65,3.618e+5,0 73,1.130e+3,0 76,3.025e+5,0 77,3.874e+i5,0 78,3.746e+5,0 140,1.681e+4,0 141 ,4.354e+-5,0 142,6.291e+-5,0 143,8.836e+5,0 144,9.685e+5,0 145,8.260e+5,0 146,4.850e+3,0 147,8.436e+5,0 148,2.529e+I4,0 149,3.025e+5,0 150,1 .745e+5, 0 151 ,7.700e+5,0 152,7.235e+5,0 159,1.093e+5,0 160,2.641e+4,0 161 ,3.762e-1,0 162,3.490e+'4,0 163,6.627e+4,0 164,8. 036e+-5, 0 165,7.604e+5,0 OUTPUT OF GRODEC CALCULATION OPERATING IN MODE 1 DATA FILE NAME:BCEAL4 INPUT DATA LISTING I.qATAPI::

INIITIAI 13rllTI NNI SPlant: HNP Ul & U2 SNC CALCULATION SN-301Atcmn Title: NEI 99-01 Rev 6 EAL Calculations 8 of 12 ACTIVITY(CI)

RATE(CI/HR)

KR-83M 5.283E+04 0.000E+00 KR-85 6.051E+03 0.000E+'00 KR-85M 1.108E+05 0.000E+00 KR-87 2.113E+-05 0.000E+00 KR-88 2.977E+05 0.000E+00 KR-89 3.618E+05 0.000E+00 RB-86 1.130E+03 0.000E+00 RB-88 3.025E+05 0.000E+00 RB-89 3.874E+05 0.000E+00 RB-90 3.746E+05 0.000E+-00 1-130 1.681E+-04 0.O00E+00 1-131 4.354E+05 0.000E+00 1-132 6.291E+05 O.000E+-00 1-133 8.836E+05 0.000E+00 1-134 9.685E+05 0.000E+00 1-135 8.260E+05 0.000E+-00 XE-131M 4.850E+03 0.000E+00 XE-i133 8.436E+05 0.000E+00 XE-133M 2.529E+04 0.000E+00 XE-135 3.025E+05 0.000E+00 XE-135M 1.745E+05 0.000E+00 XE-i137 7.700E+05 0.000E+00 XE-138 7.235E+'05 0.000E+00 CS-134 1.093E+05 0.000E+00 CS-134M 2.641E+04 0.000E+00 CS-135 3.762E-01 0.000E+'00 CS-136 3.490E+04 0.000E+00 CS-137 6.627E+04 0.000E+00 CS-138 8.036E+05 0.000E+00 CS-139 7.604E+-05 0.000E+00 WHAT ARE THE START, STOP, AND INTERVAL TIMES 1.00 1.00 1.00 TIME THIS INCREMENT

= 1.000000 HOURS TOTALACTIVITY

= 5803093.000000 CURIES ISOTOPE ISOTOPE INITIAL ADDITION ACTIVITY NUMBER NAME ACT(CI) RATE(CI/HR) (CI)60 KR-83M 60 R-3M 5.283E+04 0.000E+00 3.640E+04 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 9 of 12 b1 1K1-t-i5 b5.Ubll:+UJ U.UUUL=+UU b5.Ub2ZI:+U3 62 KR-85M 1,108E+05 0,000E+00 9,465E+04 63 KR-87 2.113E+05 0.000E+00 1.223E+05 64 KR-88 2.977E+05 0.000E+00 2.324E+'05 65 KR-89 3.618E+05 0.000E+00 7,583E-01 73 RB-86 1.130E+03 0.000E+00 1.128E+03 75 RB-87 0.O00E+00 0.O00E+00 2.680E-10 76 RB-88 3.025E+05 0.000E+00 2.570E+05 77 RB-89 3.874E+05 0.000E+00 3.237E+04 78 RB-90 3.746E+05 0.000E+00 2.334E-01 84 SR-89 0.000E+00 0.000E+00 8.717E+i01 86 SR-90 0.000E-+00 0.O00E+00 7.483E-02 92 Y-90 0.000E+00 0.000E+00 7.497E-04 140 1-130 1.681E+04 0.000E+00 1.589E+04 141 1-131 4,354E+05 0,000E+0O 4,338E+'05 142 1-132 6.291E+05 0.000E+00 4.630E+05 143 1-133 8.836E+05 0.000E+00 8.539E+05 144 1-134 9.685E+05 0.000E+00 4.354E+05 145 1-135 8.260E+05 0.000E+00 7.446E+05 146 XE-131M 4.850E+'03 0.000E+00 5,900E+03 147 XE-133 8.436E+05 0.000E+00 8.439E+05 148 XE-133M 2.529E+-04 0.000E+00 2.497E+04 149 XE-135 3.025E+05 0.O00E+00 3.420E+-05 150 XE-135M 1.745E+05 0.000E+00 1.214E+04 151 XE-137 7.700E+05 0.000E+00 1.804E+01 152 XE-138 7.235E+05 0.O00E+00 6.725E+04 159 CS-134 1.093E+05 0,000E+00 1.093E+i05 160 CS-134M 2.641E+04 0.000E+00 2.079E+04 161 CS-135 3.762E-01 0.000E+00 3.762E-01 162 CS-136 3.490E+04 0.000E+00 3.483E+04 163 CS-137 6.627E+04 0.000E+00 6.627E+04 164 CS-138 8.036E+05 0.O00E+00 3.777E+05 165 CS-139 7.604E+05 0.000E+00 9.546E+03 170 BA-136M 0.000E+00 0,000E+00 3.483E+04 171 BA-137M 0.O00E+00 0,000E+00 6.627E+04 172 BA-139 0.000E+00 0.000E+00 5.834E+04 START OF MESS RUN ISOTOPES NOT INCLUDED IN MESS RUN NAME ACTIVITY (CI)RB-86 1128.253000 RB-87 2.680317E-10 S R-90 7.48271 9E-02 Y-90 7.497340E-04 I-130 15889.090000 CS- 134M 20787.760000 CS-i135 3.762085E-01 BA-I136M 34826.530000 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 10 of 12 WHAT IS THE SOURCE VOLUME (CC)4. 142000E+09 WHAT IS THE SOURCE DENSITY (GM/CC)1.530000E-03 START EXECUTION OF THE MESS SUBROUTINE, ID NUMBERS ARE MESS ID NUMBERS,NOT MAIN PROGRAM IDS.THE NUMBER OF ENERGY GROUPS INPUT: 10 THE MAXIMUM ENERGY OF EACH GROUP: ENERGY MAXIMUM GROUP ENERGY 1 3.000000E-01 2 4.000000E-01 3 5.000000E-01 4 6.000000E-01 5 8.000000E-01 6 1.000000E+00 7 1.500000E+00 8 2.000000E+00 9 3.000000E+00 10 4.000000E+00 MAX GAMMA ENERGY OF BR 90 IS GREATER THAN THE MAXIMUM GROUP ENERGY STANDARD FIXUP: MAXIMUM GROUP ENERGY IS SET TO THE HIGHEST ENERGY YET ENCOUNTERED+'0.01 MEV NEW MAX ENERGY = 5.210000 MEV MESS INPUT DATA: SOURCE DENSITY (GM/CC) = 1.530000E-03 SOURCE VOLUME (CC) = 4.142000E+09 ISOTOPE ID NO, SOURCE STRENGTH UC/GM UC/CC CI KR 83M 45 5.743404E+03 8.787408E+00 3.639745E+'04 KR 85 47 9.549409E+02 1.461060E+00 6.051709E+03 KR 85M 46 1.493609E+04 2.285221E+01 9.465387E+04 KR 87 48 1.929548E+04 2.952209E+01 1.222805E+05 KR 88 49 3.667663E+04 5.611524E+01 2.324293E+05 KR 89 50 1.196635E-01 1.830852E-04 7.583388E-01 KR 89 51 1.196635E-01 I.830852E-04 7.583388E-01 RB 88 67 4.055830E+04 6.205420E+01 2.570285E+05 RB 89 68 5.108358E+03 7.815787E+00 3.237299E+04 BR 90 69 3.682816E-02 5.634709E-05 2.333896E-01 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEi 99-01 Rev 6 EAL Calculations 11 of 12 1 131 I 132 I 133 I 134 I 135 4 5 6 7 8 XE 131M 53 XE 133 54 XE 133M 55 XE 135 56 XE 137M 57 XE 137 58 XE 138 59 CS 134 33 CS 136 34 CS 137 35 CS 138 36 CS 139 76 BA 137M 78 BA 139 37 6. 845877E+04 7.305456E+'04 1 .347498E+05 6.869780E+04

1. 174962E+05 9.310631E+0 1 .331608E+05 3.940021 E+0 5.397361 E+04 1 .915720E+0J 2.846699E+00 1 .061136E+04 1 .724668E+04 5.495517E+03 1.045720E+04 5.960024E+04 1 .506267E+03 I1.045720E+0 9.206075E+03 C... l-trtij OL_.1 .-u 1.047419E+02 1.117735E+02 2.061673E+02 1.05 1076E+02 1. 79769 1E+02)2 1.424527E+00 S2.037359E+02

'3 6.028232E+00

• 8.257963E+01

'3 2.931051E+00

• 4.355450E-03 1 .623539E+01

• , 2.638742E+01 S8.408140E+00

• , 1.599952E+01

• , 9.118837E+01

• 2.304588E+00 4 1.599952E+01 1 .408530E+01 U,., I i IJ1 l ,JU -, UL. IJ 4.338410E+05 4.629658E+05 8.539448E+05 4.353559E+05 7,446038E+05 5.900389E+-03 8.438743E+05 2.496894E+-04 3.420448E+05 1 .214042E+04 1 .804027E+01 6.724697E+04 1 .092967E+-05 3.482652E+

04 6.627002E+04 3.777023E+05 9.545605E+-03 6.627002E+04 5.8341 30E+04 MESS OUTPUT DATA: ENERGY MAXIMUM SOURCE STRENGTH GROUP ENERGY (MEV/CC-SEC) (GAMMAS PER SEC)1 2 3 4 5 6 7 8 9 10 3.000000QE-01 4.000000E-01 5.000000OE-01 6.000000E-01 8.000000E-01 I1.000000E+00

1. 500000E+00 2.000000E+00 3.000000OE+00 5.21 0000E+00 1.51 0080E+06 1 .638563E+06 7.8 10211E+05 4.458929E+06 9.754498E+06 8.118191E+06 I1.504968E+07 6.422278E+06 6.717616E+06 5.222530E+00

• 2.084916E+16 1.696731E+16 6.469979E+15 3.078147E+16

5. 05039IE+ 16 3. 362555E+ 16 4.155719E+16 1 .330054E+16 9.274789E+

15 4.151962E+09 Table 6.1,-I. Units used to express S tabulated in relation to the units used to express q q ~Point source Line source Surface

• Volume cv as a factor of W o S SOUrCe source So S *S S y-photons cmn 2 s 7-photons cm 2 s MeV-cm 2 s_ MeV~y-photons 5 m~i (mfilicuries)

MeV 5 7-photons oms mCi cm MeV oms mCi y,-photons cm 2 s mCi cm 2 MeV mQi y-photons cm 8 sI mCi cm 8 MfeV cm~s m~i 3.7. 1Ov disintegrationis y-photons smCi *ndisintegration Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment G Title: NEI 99-01 Rev 6 EAL Calculations 12 of 12 cm~s cm cm 2 cms disintegrations V-photons E MeV smCi ndisintegration photon-'-MeY mCi mCi mCi M~eV cm 2 s mCi cm cm 2 cm 3 4" cm 2 smCi_ MeV ...Beg cm 2 s rads V-photons y-photons y-photons y'-photons 7-photon ' /a 3 0 0 h s cms cmzs cma s 100 ergs/(g.rad) ergs _cera 6-s rads MeV MeV MeV MeV 1.6' 10-8 eeV"a "g'-'0 h-h1 s cms cmts cross 100 ergs](g rad)Me 1 6 0 ergs -cm2 360 R V-photons V-photons V-photons V-photons E-hoo M.e10- ht-. 60-h s cms cross cmas 87.7 ergs/(gR)ergs *cm--2 a 36 R MeV MeV MeY MeV 1..0aM--e"V g s cms cm~s cm~s 87.7 ergs/(gR)R mCi mCi mCi 2"h- mCi cm cm 2 cm 2 2VK hinCi R mg-equiv.R mg-equiv.

Ra mg-equiv.

Ra 4 , Rcm 2-- rag-equiv.R1a cm cm 2 cm 8 a 8 h mag-equiv.

Ra Southern Nuclear Design Calculation

[Plant: HNP Unit: 1&2 [Calculation Number: SMNH-13-021

[Sheet: H-I ATTACHMENT H -VALIDATION OF SPIRAX SARCO ON-LINE STEAM TABLES Rather than interpolate from the ASME steam tables, an on-line set of steam tables was used to determine the specific volume of the reactor coolant at normal operating conditions.

Spirax Sarco, a global provider of products for the control and efficient use of steam, provides on-line steam tables at their company website, http://www.spiraxsarco .com/resources/steam-tables. asp.Reactor Coolant Normal Operating Conditions To verify that the Spirax Sarco steam tables provide accurate results, the specific volumes of subcooled water at 1000 and 1100 psia and 470 and 490 F (see sheets below) are compared below to the ASME steam table values.Units ASME Spirax Sarco Delta*ASME Spirax Sarco Delta*P psia 1000 1000 1100 1100 ____T F 470 470 470 470 ____SV cu ft/Ibm 0.019722 0.0197175

-0.02% 0.01 9704 0.01 96999 -0.02%T F 490 490 490 490 SV cuft/lbm 0.020140 0.0201348

-0.03% 0.020119 0.0201141

-0.02%* Delta =[(Spirax Sarco -ASME)/ASME]

X 100%The Spirax Sarco steam tables agree extremely well with the ASME steam tables.The linearly interpolated results from the ASME Steam Tables would likely be less accurate than using the on-line steam tables because specific volume is a non-linear function of pressure and temperature.

Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 ICalculation Number: SMNH-13-021 Isheet: H-2 Ho.me Aboat U. Pradect & 50w. -tmdsflaek

& r'r.nag Rammo -Contsct Sub Saturated Water Region -Steam Table Al any txeosse, water bedow tis saturalkon temperature ts said to be in a sub saturated state For exaopte' water at a reesure of I atmospfhere and a temperature betow I fIte saturated temperature of Is sub saturated.

Water at a pressure of 10 atroosptteres has a saturaion temp4erature of 180"C, and so water below5 fits temrperature Is also sub saturated.

Learn more about steam to ottutr ioal -jters.Set your for thseso steam tables Note: -You cannot use commas 4,) as decimal points.Pleas, use periods (.)Example: 1.02 not 1,02 Feature* Tra~let reat Swets tlamt 0 Preosora aid tamrtratere 4~ Sto~e Value '~ Table urn tie atoojot.Pressure TemeratEl Vapou PressuJre Saturation Temperature Specifc Entlialpy of Waler (hi)Oerelt of Water Specific Voftame of Water (s)Specifi Entrop of Water (si)so jAra k~n9 J~5 5 K Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13-021 Isheet: H-3 I-- e onr Spia Sxa.o In Hsqre AttstUs is Preoferts&

Seyi.

• tidtslbes

&Aepij~oans

• Trattme is Contest V Yes a-e her: Homee I) Seantt Sten Thl Suil Saturated Water Region Sub Saturated Water Region -Steam Table AJ1 any p~ewawe, water below its saturation teenperature is said to be to a su saturated state orexample water at a pressure of 1 arnospttere and a temlperature below tesaturated temperature of lt000C Is sub saturated.

Water at a pressure of 1atmosphers has a saturation temperature of 180"C, arnd so water below Urstemperature Is also sub saturated.

i Learn more about steam is our tutortat -Set your io..tS for theKse steam tables.Note: -You cannot use commas (,) as decimal points.Please use periods (.)Example: 1.02 not 1,02 Feature L Tratng ba~ltos 0* Single Value ©Tabl~e El Pressure pat atiselat."temperature Vapour Pressure ear gauge Saturation Temperature El Specifc Enthalpy of Water (hi)Denstit of Water Specifc Volume of Water (v)Specifi Entropy of Water (is)Jigs JAg K El El Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13-021 Isheet: HA4 r intiffii~nai site for Spirax Sarco Ve.t yew eltonat Seea~c Sane Cebsita tieme Abe..t Us Predart. & Serewes

• todasbios

£ Apptmabee.

-Trarnere Rewarm Coetad -Yoet tw a : Htom I> Resomy Sit Sarta Wate Regai Sub Saturated Water Region -Steam Table MIany prossee water below its saturation temperature ts said to be itt a sub saatated state For example, water at a preemie of I atmnosphere and a temperature below Ite saturated temperature of lt00'C is tub saturated, Water at a pressure of 10 atmospheres has a satura-=iton temperature of and so water below hits temperature Is also sub satuated.eant merm abou Seam le ourtutolal Set your for these steam tables Note: -You cannot us. commas I,) as decimal points.Please us. periods (.)Example: 1,02 not 1,02 Feature* Treein ma 0 Pressure aind Tanlperature

• Sn~ge Veto. Table Pressure EJ Temp-=ertture Vapour Pressure Saturation Tenmterattae Specific Enthatpy of Water (he)Density of Water Specific Volume of Water (v)Specifi Entrop of Water (Si)Jag eWe'JAg K i~J Southern Nuclear Design Calculation SPlant: HNP Unit: 1&2 Calculation Number: SMNH-13"021 ISheet: H-5 I~%ar-Intrntonal site for Spirax Sarco Homee Absst U. -P Proa..ct & Seswcea -lndealie.

& Apetir~atees

• ;ionineg Raelweca

• Coetast ,-Yeu m ere:ar Home ), k T ite Sue Set~ratad Water Regis..Sub Saturated Water Region -Steam Table Al any pressure, water below fits satuion ateperature is said to be to a suab saturated state Fer examine', water at a pressure of I atmosphere and a temperature below3 Ihe saturated temperature of 1l0"C is sub saturated Water at a pressure ofE 10 atmospheres has a saturation temperature of 180°C and so water below thds tenperature is also sot, saturated Learn more about steam is our tutorial -istSean?Set yoer for these steam tale Note: -You cannot use commas (,) as decimal points.Please us. periods (.)Example: 1.02 not 1,02 Feature LI Tratleile that s:ea °"p Eo we r sse" a er t 0 Pressure TemperatureStogie Value © Table jI.. P.absd E]Vapour Pressure Saturation Temperature Spcii Erthlp of Water (he)Denslt of Water Specifi volume of Water (v)Specifc Entropy of Wate (se)Oar eases'C kg/o9 JAg K Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 1 of 25 Purpose: The purpose of this appendix is to calculate radiation readings for AS1 (RS1) EALl.Defining Units that MathCAD will understand MWt := 10 6-W 1 Ci: .-*Bq 2.7.10 Mega Watt Thermal is same as watts.Per FGR-11 page 219 tlCi := 10-6 Ci Rem := 0.01lSv Rem mRem:= -1000 Per FGR-11 page 219 cc := mL 1 cps := --sec 1 cpm := -" mm The source terms for isotopes are provided in NL-06-1 637 and are shown below. We are only looking at the isotopes that are used in HNP FSAR Chapter 15 evaluations.

Such as Iodine's and noble gases Table 15.3-4. It is assumed that only iodine's and noble gases needs to be considered, because particulates will be retained in the primary containment water.Isotopes : "1-131""1-132""1-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 lm""Xe-133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe-138" Core__Inventory

(2.72E+04" 3.93E+04 5.52E+04 6.05E+04 5.16E+04 3 .30E+03 3.78E+02 6.92E+403 1.32E+04 1 .86E+04 2.26E+-04 3.03 E+02 5.27E+04 1.58E+03 1 .89E+04 1 .09E+04 4.8 1E+04 4.52E+04 Ci MWt'I Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEl 99-01 Rev 6 EAL Calculations 2 of 25 CorePower

= 2818.MWt RCS inv := 9965 ft3 The core power is provided in NL-06-1637 page 14 Enclosure 1.The RCS inventory is provided in NL-06-1637 Enclosure 1 page 47 Table 34.To determine the RCS inventory mass the water specific volume needs to found as follows: UI: The Main feed water temperature is 392.4 F and core recirculation temperature is 532.0 F. Per GE-NE-0000-0003-0634-01 pg 15.392.4 0 F + 532.0 °F T Avgu 1 := = 4.622E+002.°F U2: The Main feed water temperature is 425.7 F and the core recirculation temperature is 535.0 F. Per GE-NE-0000-0003-0634-01 pg 16.425.7 0 F + 535.0 0 F T Avgu 2 := 2= 4.804E+002.

0 F The higher temperature will give smaller mass and therefore the higher temperature was used to determine the specific volume. According to GE-NE-0000-0003-0634-01 page 19 the reactor pressure is 1060psia.ft 3 VRCS water : 0.0199161-

-Ibm* According to http://www.spiraxsarco.

com/resources/steam-tab les.asp the specific volume for 1060psia and 480.35 F conditions is 0.0199161ft^3/lbm.

This website was validated in Attachment H. The copy of the webpage and the provided information is attached to this attachment below.The RCS mass is calculated as follows: RCS inv RCS mass .- -2.27E+008-gm

'URCS water Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 3 of 25 The Core Release Fractions are taken from RG1.183 Table 1. This fraction describes release of isotopes to RCS water.Isotopes 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83mi""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-138" F_Core_Release

(0.3 10.3 0.3 0.3 0.3 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEl 99-01 Rev 6 EAL Calculations 4 of 25 The concentration of isotopes in RCS is calculated as follows: Conc_IsotopeRCS

Core Inventory.Core

_Power-F FCore Release RCS mass Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""1-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe-138" Cone_IsotopeRcS 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.013E+005 1.464E+005 2.056E+005 2.254E+005 1.922E+005 4.097E+004 4.693E+003 8.592E+004 1.639E+005 2.309E+005 2.806E+005 3.762E+003 6.544E+005 1 .962E+004 2.347E+005 1 .353E+005 5.972E+005 5.612E+005 gm The RCS will have some equilibrium noble gases and some iodine in the RCS water from normally operating reactor. According to HNP FSAR U2 Table 11.1-2 the levels of iodine's is on the order of 1 E-1 uCi/g which is negligible to calculated above order of magnitude of 1 E5 uCi/g.The reason these isotopes are in low concentrations because they are continually removed in the power plant steam condensers.

Thus initial equilibrium noble gases and iodine's will be neglected.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 5 of 25 The Partition Coefficient is the ratio of the concentration of a nuclide in the gas phase to the concentration of that nuclide in the liquid phase when the liquid and gas are at equilibrium.

It is assumed that 100% Noble gases are released into the steam. According to NUREG-0016 Table 2-7 page 2-13 the Iodine's partition coefficient is 0.004.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe-138" FPartCoefficient

0.004 0.004 0.004 0.004 0.004 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 6 of 25 The release concentration of isotopes is calculated by multiplying the concentration of isotopes in RCS by the partition factor and an arbitrary density of release fluid. The arbitrary density of fluid will cancel out. For ease of math calculations it was chosen to be 1lgm/cc.Prls c=c.g XL =(ConcIsotopeRCS-FPartCoeffcient-Pris)>Isotopes =0 1 2 3 4 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-8 5""Kr-85m""Kr-87""Kr-88""Kr-89""Xe- 13 lin""Xe- 133""Xe-133rn""Xe-135""Xe-135mn""Xe-137""Xe-138" XRL~S 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 4.053E+002 5.856E+002 8.225E+002 9.014E+002 7.688E+002 4.097E+004 4.693E+003 8.592E+004 1.639E+005 2.309E+005 2.806E+005 3.762E+003 6.544E+005 1.962E+004 2.347E+005 1.353E+005 5.972E+005 5.612E+005 IICi cc Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 7 of 25 Reactor Building Vent TEDE part of the calculation:

According to HNP ODCM page 3-14 the release from reactor building vent is considered to be a ground level release.-6 sec__X~gnd := 8.37.10--3 m Table 3-4 of HNP ODCM page 3-17 Q~rlS~x1d

= 1.42.10 mL = 3.009E+005.cfm Table 3-4 of HNP ODCM page 3-17 sec The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEABRxBLD

= (Q~rlspxld'XQgnd-XRj~s)

Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""1-133""1-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe- 131mi""Xe-.133""Xe-133m""Xe-135""Xe-135m""Xe-137""Xe-138" XEAB RxBLD 7-0 1 2 3 4 5 6 8 9 10 11 12 13 14 15 16 17 a 4.817E-001 6.96E-001 9.775E-001 1.071E+000 9.138E-001 4.87E+001 5.578E+000 1.021E+002 1 .948E+002 2.745E+002 3.335E+002 4.472E+000 7.777E+002 2.332E+00 1 2.789E+002 1.609E+002 7.098E+002 6.67E+002 tlCi cc Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 AttachmentI Titie: NEl 99-01 Rev 6 EAL Calculations 8 of 25 The Dose Conversion Factor (DCF) for Effective Dose Equivalent (EDE) was taken from FGR12'Effective Column" of Table 111.1. The unit conversion was performed with MathCAD internal features.Isotopes 0"I-13 1""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe- 135m""Xe- 137""Xe-138" DCFEDE :

1.12E-13 2.94E-14 1.30OE-13 7.98E-14 1.50E-18 1.19E-16 7.48E-15 4.12E-14 1.02E-13 0 3.89E-16 1.56E-15 1.37E-15 1.1 9E- 14 2.04E-14 0 5.77E- 14)Sv" Bq. sec 0 2.427E+005 1.493E+006 3.92E+005 1.733E+006 1.064E+006 2E+001 1.587E+003 9.973E+004 5.493E+005 1.36E+006 OE+O000 5. 187E+003 2.08E+004 1.827E+004 1.587E+005 2.72E+005 OE+O000 7.693E+005 mRem. cc hr. 3 Sv. m mRem.cc-1.333E+019.

B q. sec hr. MathCAD internal conversion feature.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 EAL Rev 6 Calculations 9 of 25 The Dose Conversion Factor (DCF) for Committed Effective Dose Equivalent (CEDE) was obtained from FGRI1 Table 2.1 column labeled "Effective".

The unit conversion was performed with MatchCAD internal features.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"1-131""1-132""I-133""I-134""1-135""Kr-83m"~"Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe-137""Xe- 138" DCFcEDE: S8.89E_09" 1.03E-10 1 .58E-09 3.55E-11 3.32E-10 0 0 0 0 0 0 0 0 0 0 0 0 0 Sv 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.293E+001 3.815E-00 1 5.852E+000 1.3 15E-001 1.23E+000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O OE+O000 OE+O000 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 mRem p.Ci Sv mRem= 3.704E+009.-

Bq liCi MathCAD internal conversion feature. Same conversion factor is available on page 121 of FGR-11.

SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 AttachmentiteNE9-0Rv6EACaclios1 If2 The EDE, CEDE, and TEDE are calculated as follows: The exposure time is provided in NEI 99-01.texp := lhr 3-4m BR := 3.5.10 -sec The breathing rate for persons offsite is listed in section 4.1.3 page 1.183-16 of RG-1.183.EDE~xBLD :=(DCFEDE'XEABRxLD'texp)f Effective dose Equivalent during a release from Reactor Building Vent.,

Effective Dose Equivalent from CEDERxBLD

= (DCFcEDE'XEAB_RxBLD'*texp BR)nhalation during a release from Reactor Building Vent.0 1 13 i5 16.7 i *0' 1. 169E+005 1.039E+006 3.832E+005 1.857E+006 9.723 E+005 9.74E+002 8.851E+003 1.019E+007 1.07E+008 3,733E+008 0E+000 2.319E+004 1.618E+007 4.259E+005 4.426E+007 4.375E+007 0E+000 5. 132E+008*mRem CEDE~xBLD

-41:7 9'9'10 12 13 14 15;17 1.998E+007 3.345E+005 7.208E+006 1.775E+005 1.416E+006 0E+000 0E+000 OE+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+i000 0E+000 0E+000-mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

TEDERxB3LD

- ZEDEpxBLD

+ ECEDEPxBLD

= 1.142E+009.mRem Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 11 of 25 The reactor building monitor indication related to the exclusion area boundary TEDE is calculated as follows: X100_RxBLD

S100 mRem X The 100 mRem comes from NEI 99-01.Isotopes=2'4 6'8 11 12 14 15 17"I-131""1-132""I-133Y"I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Xe- 13 im""Xe- 133""Xe-133m""e15""Xe-135m""Xe- 137""Xe-138" Xlo0 RxBLD =(0f 2-5,¸6¸: 7 8BI 9 10 11 12 15 16 17* ,.. 0 " 3.549E-005

5. 128E-005 7. 203 E-005 7.895E-005 6.733E-005 3.589 E-003 4.111E-004 7.525E-003 1.435E-002 2.023E-002 2.458E-002~

3.295E-004 5.73 1E-002 1.718E-003 2.055E-002

1. 185E-002 5.23 1E-002 4.915E-002]

IlCi cc Since the detector only respond to noble gases according to HNP FSAR Table 7.5-1 (Sheet 31 of 34) note 14 and pg 3 of Doc ID RE203727981, thus only noble gas isotopes will be summed.17 XTEDE_100_RXBLD

=i=5 This is summation of rows that correspond to Xl00_RxBLDinoble gases only. These rows are 5 through 17 of the matrix above.XTEDE_100_RXBLD

= 2.639E-001.

c Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NE! 99-01 Rev 6 EAL Calculations 12 of 25 CDE part of the calculation:

The Thyroid Committed Dose Equivalent (CDE) from inhalation was obtained from FGR 11 Table 2.1 column labeled "Thyroid".

The unit conversion was done via MathCAD internal unit conversion feature.Isotopes =0.2 3 5.6 7 8 10 11.12 13 15 17 0"I-131""I-132""1-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe- 135m""Xe-137""Xe- 138" DCFcDE : r2.92E_07"N 1 .74E-09 4.86E-08 2.88E-10 8.46E-09 0 0 0 0 0 0 0 0 0 0 0 0 0 Sv.0 2.3 4 6 7 8 9.10 11 12 13!5 16 17 0 1.081E+003 6.444E+0001.067E+000

3. 133E+001 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 mRem liCi 1 Sv- 3.704E+009.

• e Bq p~tCi MathCAD internal conversion.

Corresponds with FGR-11 page 121.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations l3 of 25 The CDE exposure due to the Reactor Building Vent release is calculated as follows: 3 BR = 3.5E-004.m sec texp = 1E+000.hr Provided above In tis quaton he DE dse onvrsinforvise mulipledbistpscneraont exclusion area boundary, and breathing rate, and time of exposure.IsotopoCSDERxBLD

= (DCFCDE.XEAB_RxLD'BR'texp))

IsotopesCDE~xLD=

2,"5 6 17 81 i-i 12 14'5 16 17',i "0. ..6.564E+008 5.651E+006 2.217E+008 1.44E+006 3.608E+007 OE+O00 OE+O000 OE+O000 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 OE+O000 OE+O000*mRem CDE~xLD := ZlsotopesCDERxB1LD

= 9.213E+008.mRem Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 14 of 25 The reactor building monitor indication related to the exclusion area boundary CDE is calculated as follows: ( 500mRem~5 OO RxBLD :=-y CDERXBLD) .XRLS The 500 mRem comes from NEI 99-01.Isotopes=0 1 2 3 4 5 6.7 8.9 10 11 12 13 14 15 16 17 0"1-131""I-132""I-133""1-134""I-135""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe-138" X 5 00 RxBLD=0 2 3 6 7 8 9 10 111 12 14 15 16 17 0 2.2E-004 3. 178E-004 4.464E-004 4.892E-004

4. 173E-004 2.224E-002 2.547E-003 4.663E-002 8.895E-002 1.253E-001 1.523E-001 2.042E-003 3.551E-001 1.065E-002 1.274E-001 7.345E-002 3.241E-001 3.046E-001 cc 17 XCDE_500_RXBLD

= E X500-RxBLDi i= 5 XCDE_500 RxBLD =_l.635E+000-l~This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.

Piant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 15 of 25 XTEDE_100 RxBLD = 2.639E-001.Ixi cc This was calculated above and since it is more limiting should be used as the indication for this EAL.Main Stack TEDE Part of the calculation:

According to HNP ODCM page 3-14 the release from Main Stack is considered to be a Elevated release.-8 sec X Qelev := 4.10-10 .-3 m Q~rlSMain

= 9.44.10 mL- 2E+004.cfm sec Table 3-4 of HNP ODOM page 3-17 Table 3-4 of HNP ODCM page 3-17 The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEABMain :=(QrlSMain"XQelev'XRWs))

Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"I-131""I-132""I-133""I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe-133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe-138" XEABMain 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.569E-004 2.266E-004 3.183E-004 3.489E-004 2.976E-004 1.586E-002 1.817E-003 3.326E-002 6.344E-002 8.939E-002 1.086E-001 1 .456E-003 2.533E-001 7.593E-003 9.083E-002 5.238E-002 2.312E-001

2. 172E-001 liCi CC Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 16 of 25 The dose conversion factors were provided above in this calculation and are used in the same manner. The EDE, CEDE, and TEDE are calculated as follows: tex = 1E+000.hr 3 m BR = 3.5E-004.-

sec Provided above Provided above EDEMain := (DCFEDE'XEAB_Main'texp))

Effective dose Equivalent during a release from Main Stack.Committed Effective Dose Equivalent from inhalation during a release from Main Stack.CEDEMain := (DCFcEDE-XEABMain~texp.BR)

EDEMain=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.806E+001 3.384E+002 1.248E+002 6.048E+002

3. 166E+002 3.172E-001 2.882E+000 3.3 17E+003 3.485E+004 1.216E+005 0E+000 7.552E+000 5.268E+003 1.387E+002 1.441E+004 1.425E+004 0E+000 1.671E+005.mRem CEDEMain =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 6.508E+003 1 .089E+002 2.347E+003 5.78E+001 4.61E+002 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000 0E+000*mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 17 of 25 TEDEMain := EEDEMain + Main =3.718E+005.mRem The Main Stack monitor indication related to the exclusion area boundary TEDE is calculated as follows: Xl0Main := 100T~en)'XRLS The 100 mRem comes from NEI 99-01.Isotopes =o 1 2 3 4 7..8 9'10 11l 1i2 13 16 17"I-131""I-132""I-133""1-134""I-135""Kr-83m""Kr-85""Kr-87""Kr-88"*"Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe-135""Xe-135m""Xe- 137""Xe- 138" X100 Main 0 1.09E-001 1.575E-001 2.212E-001 2.424E-001 2.068E-001 1.102E+001 1.262E+000 2.311E+001 4.408E+001 6.211E+001 7.547E+001 1.012E+000 1.76E+002 5.276E+000 6.311E+001 3.64E+001 1 .606E+002 1.509E+002 IlCi ee Since the detector can discriminate noble gases according to pg 3 RE203727981, thus only noble gas isotopes are summed. The P DMS connects 1D11N055 and 1D11N056 to 1D11P006 and manual $57925.17 XTEDE_100-Main

= E Xl0Main.i= 5 This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.

Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 18 of 25 XTEDE 100 Main = 8.104E+002*-~

--cc CDE part of the calculation:

The ODE exposure due to the Main Stack release is calculated as follows: 3 m BR = 3.5E-004.-

sec texp =1E+OOO0hr Provided above In tis euatin th ODEdoseconvrsinforvise mulipledbistpscneraont exclusion area boundary, and breathing rate, and time of exposure.IsotopesCDEMain

= (DcCcDE.XEAB_Main.BR~texp))

IsotopesCDEMain

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 2. 137E+005 1.84E+003 7.22E+004 4.689E+002

1. 175E+004 OE+O00 OE+O00 OE+O000 OE+O000 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00 OE+O00.toRero Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 19 of 25 CDE Main := lsotopesCDEMain

= 3E+005.mRem The Main Stack monitor indication related to the exclusion area boundary CDE is calculated as follows: X50Main := *XDEai)RLS The 500 mRem comes from NEI 99-01.Isotopes = 1, 2.5 16 7 9 i10 11 12 13 14:15 16 ,17"I-131""I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131m""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe- 138" X500_RxBLD

='2" 7 8 9 10 11 12 13 14 17 0 2.2E-004 3. 178E-004 4.464E-004 4.892E-004

4. 173E-004 2.224E-002

2. 547E-003 4.663E-002 8.895E-002 1.253E-001 1.523E-001 2.042E-003 3.551E-001 1 .065E-002 1.274E-001 7.345E-002 3.241E-001 3.046E-001 cc XCDE_500 Main:=i= 5 X50Maini This is summation of only noble gases, rows 5 through 17.I pCi I XCDE 500 Main 5.022E+003 I -cc Plant: HNP U1 & U2 SNC CALCULATION SMNH-13--021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 20 of 25 IXTEDE_100 Main = 8"I04E+002" liCi[cc This was calculated above and since it is more limiting should be used as the indication for this EALU Recombiner Building Vent TEDE part of the calculation:

According to HNP 0DCM page 3-14 the release from recombiner building vent is considered to be a ground level release.Sec XQgnd =8.37E-006.--

35m m Provided above

d. := 2.36.105 ..- = 5.001E+002.cfm Table 3-4 of HNP ODCM page 3-17 The radio nuclide concentration at Exclusion Area Boundary is calculated as follows: XEAB~aeB~d

-- (OrIsReCBld.X gfnd-Xjus))

Isotopes ="I-131"1"I-132""I-133""I-134""I-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-13 im""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe- 138" XEAB RecBld 0 I.2.3 4 5..6.7.8.9 12 14 15 16 17 8.006E-004 1.157E-003 1.625E-003 1.781E-003 1.519E-003 8.094E-002 9.,271E-003 1.697E-001 3.238E-001 4.562E-001 5.543E-001 7.432E-003 1.293E+000 3.875E-002 4.636E-001 2.673E-001

1. 18E+O000 1.109E+000 litCi cc Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment II Title: NEI 99-01 Rev 6 EAL Calculations 21 of 25 The EDE, CEDE, and TEDE are calculated as follows: texp = IE+000.hr 3 m BR = 3.5E-004.--

see Provided above Provided above EDERecBld

= (DCFEDE'XEAB_RecBld'texp)r Effective dose Equivalent during a release from Reactor Building Vent./ ,>Committed Effective Dose Equivalent from CEDERcCBId

= I[DCFcEDE'XEABRecBld*texp'.BR) inhalation during a release from Reactor Building Vent.EDERecBId=

0 1.943E+002 1.727E+003 6.369E+002 3.086E+003 1.616E+003 1.619E+000 1.471 E+00 1 1.693E+004 1.778E+005 6.204E+005 OE+O000 3.855E+001 2.689E+004 7.079E+002 7.355E+004 7.272E+004 0E+000 8.529E+005.mRem CEDERecBId

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 3.321E+004 5.56E+002 1. 198E+004 2.95E+002 2.353E+003 OE+O000 OE+O000 OE+O00 OE+O000 OE+O000 0E+000 0E+000 0E+000 0E+000 OE+O000 OE+O00 0E+000 0E+000-mRem Per section 4.1.1 of RG 1.183 TEDE is a sum of CEDE and Deep Dose Equivalent (DDE).However, section 4.1.4 of RG 1.183 says that DDE and EDE are equivalent.

Plant: HNP UI & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 22 of 25 TEDERecBId

= ZEDERecB~d

+ ZCEDERecBId

= 1. 898E+006-mRem The recombiner building monitor indication related to the exclusion area boundary TEDE is calculated as follows: Xl00 RecBld : C 00rnRern STEDERecBldI "XRLs The 100 mRem comes from NEI 99-01.Isotopes =0 0 "I-131" 1 ."1-132" 2 "I-133" 3 "I-134" 4 "1-135" 5 "Kr-83m" 6 "Kr-85" 7 "Kr-85m" 8 "Kr-87" 9 "Kr-88" 10 "Kr-89" 11 "Xe-131lm" 12 1"Xe-133" 13 "Xe-133m" 14 "Xe-135" 15 "Xe-135m" 16 "Xe-137" 17 "Xe-138" XlO0_RecBlid-0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 2. 136E-002 3.086E-002 4.334E-002 4.75E-002 4.051E-002

2. 159E+000 2.473E-001 4.528E+000

8. 637E+000 1.217E+001 1.479E+001 1.983E-001 3 .448E+00 1 1.034E+000 1,237E+001 7, 132E+000 3,147E+001 2.957E+001 iiCi CC Since the detector only respond to noble gases according to pg 3 of DoclD: RE203727981, thus only noble gas isotopes will be summed.17 XTEDE_100_RecB~d

= X100-RecBldi i= 5 This is summation of rows that correspond to noble gases only. These rows are 5 through 17 of the matrix above.XTEDE_100_RecBld

=1'588E+002-lCi cc i Plant: HNP Ul & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 23 of 25 CDE part of the calculation:

The ODE exposure due to the Main Stack release is calculated as follows: BR = 3.5E-004.m sec texp = 1E+000.hr Provided above Provided above In this equation the CDE dose conversion factor is multiplied by isotopes concentration at exclusion area boundary, and breathing rate, and time of exposure.IsotopesCDERcCBld

= (DCFcDE' XEAB_RecBld.BR-texp))

IsotopesCDERecBld

=0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.091E+006 9.392E+003 3.685E+005 2.393E+003 5.996E+004 OE+O00 0E+000 OE+O00 OE+O000 OE+O000 OE+000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000 OE+O000.tRero Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 24 of 25 CDERcCB~d

= LlsotopesCDERecBld

= 1.53 1E+006-mRem The Recombiner Building monitor indication related to the exclusion area boundary CDE is calculated as follows: ( 500mRernm X50RecBld

= ) L The 500 mRem comes from NEI 99-01.Isotopes =0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0"1-131""1-132""1-133""I-134""1-135""Kr-83m""Kr-85""Kr-85m""Kr-87""Kr-88""Kr-89""Xe-131lm""Xe- 133""Xe-133m""Xe- 135""Xe-135m""Xe- 137""Xe-138" X500_RecBld 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 0 1.323E-001

1. 912E-00 1 2.686E-001 2.944E-001 2.511E-001 1.338E+001 1.533E+000 2.806E+001 5.352E+001 7.542E+001 9.164E+001 1.229E+000 2.137E+002 6.407E+000 7.664E+001 4.42E+00 1 1.95E+f002 1 .833E+002 kiCi cc 17 XCDE_500_RecBld

= X500-RecBld.

i= 5 XCDE_500 Re =l 9.84E+002.

l~-~cc This is summation of only noble gases, rows 5 through 17.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment I Title: NEI 99-01 Rev 6 EAL Calculations 25 of 25 XTEDE_100_RecBld

= 1.58+/-0 ~_ cci This was calculated above and since it is more limiting should be used as the indication for this EAL.FCaIR 7 6 3 := 4.80-107 cpm cceD The calibration factor for 1011R763A/B was obtained from DoclD RE203186522.

XTEDEI 00_RecBld counts :=XTBDE_100_RecBld'F-Ca 1 R 7 6 3= 7.622E+009.cpm XTEDE 100 RecBld counts Purpose: The purpose of this appendix is to calculate expected radiation field at Drywell Radiation Monitoring sensors (CG1).As water level in the RPV lowers, the dose rate above the core will increase.

The dose rate due to this core shine should result in up-scaled Containment High Range Monitor indication and possible alarm. Containment Challenge Table calculations should be performed to conservatively estimate a site specific dose rate set point indicative of core uncovery (ie., level at TOAF). Additionally, post-TM I studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

Defining Units that MathCAD will understand MWt :=106.W Mega Watt Thermal is same as watts.1 Ci .- Bq Per FGR-11 page 219-11 2.7.10:= 10-6 Ci Rem := 0.01Sv Per FGR-11 page 219 Rem 1 1 mRem := -cc := mL cps :=-- cpm := -" 1000 sec mm The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top Of Active Fuel. The calculation pdf page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Sense the RPV core is assumed to be columized source the same radiation level is assumed to be at the edge of the RPV. This is conservative and according to the BH2-M-V999-0048 pdf page 39 column 2 on the edge of the SFP the radiation level is 2.29E5R/hr.

As can be seen the radiation level does not very much from the middle of the core to the edge of the core. The page B-9 of BH2-M-V999-0048 provides the gamma source strength broken down by energy groups. It can be seen that the maximum gamma strength is spread between 0.4 to 1.8 MeV, therefore this calculation will use linear attenuations that are associated with 1MeV gamma rays.RSFP := 2.68-105 R BH2-M-V999-0048 last page of the caic column 3.hr The next step is to calculate the plant elevation level that corresponds to the TOAF for vessel instrument zero.TOAFvz := -158.44in TS Bases 2.1.1.3 for 150 inch long fuel.VIZ := 517in VIZ :=517inthe vessel instrument zero is in reference H26189 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment J Title: NEI 99-01 Rev 6 EAL Calculations 2 of 4 VZEL :=143ft + lin The elevation of the vessel zero is in reference H16032 TOAFEL :=VZEL + VIZ -TOAFvz =199.37.ft According to H16241 and H26417 the Drywell Wide Range detector is located at EL156ft and 27ft from the center of RPV.DetEL :=156ft DetR := 27ft When the water level reaches TOAF some radiation will travel from the edge of the core to through the RPV steel and Concrete Sacrificial shield as shown on the figure below.therefore the first step is to find the angle (line of site) from TOAF to the detector.

This will allow calculating the distance that gammas will travel through the shielding material (steel and concrete).

H16032 SPlant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment JI Title: NEI 99-01 Rev 6 EAL Calculations 3 of 4 Per FSAR U2 Table 5.2-6 the RPV material is steel. Per FSAR U2 Table 5.4-1 and S 15213 the RPV inside diameter is 218inches.

Per FSAR U2 Table 5.2-10 the wall thickness is 5.38 inches.DIA~pV:= 218in Thickwall

= 5.38in The next step will determine the angle of the line of sight from the TOAF RPV edge to the detector.AY :=TOAFEL -DetEL = 43.37.ft Difference in elevation.

DIA~pV RPVR 2- -9.083.ft Determining radius from diameter DetR = 27.ft AX := DetR -RPVR =17.9 17-fi ta~)=opposite adj asent Difference in their radiuses to determine deltaX Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.(AX'\Q:= atani -I = 22.446.deg

~ AY,.J This is the angle for a shortest line of site travel for a gamma ray to the detector.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment4J Title: NEI 99-01 Rev 6 EAL Calculations 4 of4 This step will determine linear distance that a gamma ray will have to travel trough steel vessel at the angle determined above.ino)= opposite hypotenuse Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.Thickwl d steel .- wal -14.09l1in sin(ca)or dsteeI = 35.79.cm This step will determine radiation level gamma radiation by the RPV steel.1'steel := 0.460.-cm I(t) = Io-e-t'R RSFP = 2.68E+005.-

hr dsteeI = 3.579E+001.cm on the other side of the RPV wall due to attenuation of Linear attenuation for steel at 1 MeV from page 178 of Engineering Compendium of Radiation Shielding Vol 1.Radiation intensity equation due to material attenuation page 170 of Engineering Compendium of Radiation Shielding Vol 1.determined above determined above-P~steel'dsteel

-IRpV := RSFp.e hr.19-The resulting value is much smaller than the lower range of the Drywell Radiation Monitoring sensor of I to 10A7 R/hr (Attachment C). There is no point to further account for the attenuation that the concrete sacrificial shielding would introduce, since it will lower the value even further.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 1 of 15 Purpose: The purpose of this appendix is to calculate radiation readings expected at Area Radiation Monitors due to core shine. This is done to satisfy initiating condition CG1 EAL2b "Other indications".

Defining Units that MathCAD will understand MWt := 106-W Ci .- .Bq Per FGR-11 page 219 2.7-10 ltCi := 10-6 Ci Remn:= 0.01Sv Per FGR-11 page 219 Rem 1 1 mRem .- cc := mL cps := -- cpm :=-1000 sec mai The calculation BH2-M-V999-0048 provides dose rate for Spent Fuel Pool with a full core with water level at Top Of Active Fuel. The calculation pdg page 39 column 3 calculated this radiation level at the center of the core as 2.68E5 R/hr. Since the RPV core is assumed to be a columnized source, the same radiation level is assumed to be at the edge of the RPV.DRsFP := 2.68.105.Ri BH2-M-V999-0048 last page of the caic column 3.hr The main body of this calculation (Section AU2 EALIlb) provides a list of Area Radiation Monitors and their distance to the edge of the drywell. The furthest monitor is Refueling Floor Stairway and is approximately 95 feet away from the center of the drywell. None of the listed monitors have a direct view of the reactor core. However, they do "see" gammas that reflect off the refueling area ceiling. The operating deck dose rate due to these reflected gammas is given by the following from Davisson "Gamma Ray Dose Albedos" (copy provided below in this attachment).

Area DRmon = DRsfp.COS(9).

.Rmonitor 2 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 2 of 16 The first step is to find the angle of the building roof reflection to the Refueling Floor Stairway monitor and the distance from the roof to the detector.AX := 95ft EL roof := 281ft + 9in =281.75.ft ELdet := 233ft AY :=ELroof -ELdet =4 8.7 5-ft tan(0)-= opposite adjacent This is the horizontal distance from the drywell center to the detector.Bottom elevation of the roof DWG H25963 Section 3.The detector elevation is given in the table for AU2 initiating condition within the main body of this calculation.

Right angle triangle equation from page E2 of Gieck"Engineering Formulas" 7th edition.Rmonitor := JAX 2 + AY 2 = 106.778.ft This is hypotenuse from the secondary containment ceiling to the radiation detector.cx := 0.5099%cx= 0.005099 or The dose albedo for 1 MeV gamma for iron with incident angle of 0deg and emerging angle of between 55.2 to 64.6 degrees. This albedo is obtained from the tables listed below in this attachment.

DWG H25694 shows a metal roof deck. 1MeV gamma is chosen because most of the radiation from the SFP falls in the range of 1MeV per BH2-M-V999-048 page B-9.The drywell radius is given in DWG H25570 RDrywell := 18ft+/- +l1in = 18.833.ft Ae:=r.Drywell Area Rem DRmon :=DRSFPD-cos(0).o

--cx = 60.975.-Rmonitor 2 h or DRmon = 6.098E+004.

mRem hr The area radiation monitor range is 1 -1E4 mR/hr as described in main calculation forAU2 initiating condition.

It can be seen that the radiation at the furthest Area Radiation Monitor due to reflection from secondary containment roof is off scale for the detector and thus all other monitors that are closer to the drywell would also be off scale.

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K Title: NEI 99-01 Rev 6 EAL Calculations 3 of 15 A145/SD-7$/1-4 A HANDBOOK OF RADIATION SHIELDING DATA J. C. COURTNEY, EDITOR Nuclear Science C~enter Louisiana State Universityv Baton Rouge and Shielding and Dosimetry Division American Nuclear Seclety JULY 1 1978 Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment K]Title: NEI 99-01 Rev 6 EAL Calculations 4 of 15/5-27 Gamma Bay Dose Albedos C. M. Davisson U. S. Naval PReserch Laboratoryi The dose rate reflected from a surface as deduced from Reference 1 through 4 may be represented as: P.R. =D.R,° cos ° a(Eo,98, , r where D.R. -Reflected dose rate D.R.° Dose rate incident on surface at 60 A -Reflecting area r Distance from center of reflecting area to receptor CA and r 1 must be in the saine units)tt(E 6wo, 6, *) = Dose al~bedo le abeod ll(E~ , c,8,>, for gaimas incident on water. concrete.

iron and techniques in an extension of the original work by Theus and Beech 6.The albedos are given for incident game energies of 0.2) 0.662) 1.0, p2.5 and 6.13 MIeV and for incident angles with respect to the normal of 00, 220O* 44°, 660 and 880, as well as for point sources on the surface of the materials.

The emerging polar angles, 8i, as well as the emerging sectors or directions into which the emerging gammas were divided are shown in Fig. 5.13. The values of the polar angles, O., and of the azimuthal angles 4, defining the emerging directions, are given on each page of Table 5.8.Note: The dose albedo values have statistical eirrors that range from 402 or 50Z at very small albedo values to 5% or 10% at large albedo values.Ref erences' Reactor Shielding Design Mdanual, T. Rockwell IllI, editor, rIID-7004 (larch 1956)p. 334.2 D. 3. Raso, "Ilonte Carlo Calculations on the Reflection and Transmission of Scattered Gaimna Rays," Nuzcl. Sol. and Eung. 17, 411 (1963). This report has a good discussion of the meaning of various terms and derived quantities.

The lose albedos given here are those which he described in quotes, as "dose" albedos.s W. E. Seiph, "Neutrons and Ga~mma-lay Albedos," DASA-189 2-2 (May 1967), OBNL-BSIC-21 (February 1968), or Chapter 4 of Weapons Radiation Shielding Handbook (NTIS No.AD-816 092). The dose albedos given here arc those defined as cni in this report.'9 R. L. French. and 1!. B. Wells, "An-Angle-Dapondent Albedo foi Fas 2--Reflactior Calculations," Nuel. Sci. and 1mg. 19, 441 (1964).5C. N. Dayvissom and-L. "Gamma-Ray Albedos of iron," NRL Quarterly on Nud. Scd. and Tech. (January 1, 1960), p. 43; and private communication.

SR. B. Theus and L. A. Beach, "Gamma-Ray Albedo," NRL Quarterly on tued. Sci. and Tech. (July-September 1955).

Plant: HNP U1 & U2 SNC CALCULATION SMNH-13-021 Attachment KI Title: NEI 99-01 Rev 6 EAL Calculations 5 of 15 5-28 F~iSure 5.13 Geometry arnd Solid Angle Divisions OAII{A 615 lOS 052120 Po1.r 300001000 0.02 03 0.66 IsV ASile 6,ioldos.

i1 Point

• incident Cl Point A EL 016 11 61 261 11. 66 861 lootcO 61 03 1.4 661 03 Soorce 0.-1.t 1 32.0-160.3 3.7217 5.0279 6.59 r.,g .1sg6 6.7109 6.5177 1,9555 5.O01.0 2.3331 3.1226 2.3036 2.9521 0-5.. 2 0.0- 20.0 t. 0580 6.0560 6.7552 6.8551 2.050o *.0107 ,.050. 2. 1272 0. 3769 3. 1216 2.2009 6.15053 0.2-160.0 5.622 5.655-9 6.17119 7.0712 6.7605 6.2620 1.6923 1.0022 2.5+975 2.95229 2.6011 1.2.6 0.0- 112.0 6. 1556 5.6935 6.1797 6.5515 7.21.53 ,,oo 2~J .0361 2., 222 3.500 1.5959 9.i565 6 5.5/2 6 6.1701 3.0213 1,76.3 0.1723 3.6~5+21.6-51..

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5 3.5258 5.7106 5.8529 51.718 1.2016 1.7152 57 112.5-1.5.0

5. 552 2. 5657 1.82 2.0o25. 9 1.137 1.1367 6.7217 2.6232 32 15.0- 67.5 5.11.92 1.50o02 6.6516 6.2112 1.31.32 1.7870 5.5Z061 6.1.255 31 20.5- 5.o 5. 3066 712 7.4570 15. 57571.o .L 1076 52 0.2. 20.5 2.07 5IT .5837 0.0211. 15.6 1.7 1. 506, 3.2151 1.00 5, 0.7556.9 35 165.0-020.0 2.631. 2.1966 3.5050 .6611. .0537 1.555 1.2397 34 155.0-163,0 3.251 2.6515 3.51515 5,5015 .6372; .6555 6.1556 1.6204 35 135.2-155.0 2.2259 2..1537 5.5257 3.1516 .711l.1 .6135 6.1151 1.93253 36 122.0-155.0 3.160 5.3.162 3.31.50 5.71..10 36 120 202 27 955.0-122.0

2,0710 2.0599 2.3.214 3.54 0.5769 .7615 .2157 1.17 3.35,O7 6, 38 20.0-103.0 6k.0159 2.2706 2. 7706 5.6s1o1 14.sox .72553 .7550 1.5095 1.5537 2.7010 60 .6-77.6 59 75.0- 90.0 2.1.57 2.6761 1 91 6.1131 6

.0112 .7835 6.1200 6.016 3.723. :k.coTo 10 60.0- 75.0 2.1t719 3z. 606 1. o16 7.5955 1.0125 6.1.012 2.1133 1.5050 41 1~o 062.1670 5.1039 5.025T 9.5597 .9793 1,53-90 2.9164 6.8533 S3.3 , 22.- 1.0 ~ 2.s601 2.6170 6.3645 12.11255 .9302 1.6891 1.030 51.9030 S15,05.0 30.2 2.03 7.05.10 15.0022 .0866 6.658 1/.0537 1751 II 2..- 15.0 2.5059 3.7160 7.5061 1"7.2589 1.1235 5.21.2 6.1 5.2097 1. 10.0-165.0

.6556 .761.3 1.2862 2.2572 .0535 .2.57 .3727 3.5970 08 90.0.105.0

.7116 .6125 .0167 1.5501 5.0618 2.0502 .2536 .1201 .-11935 .6755 1.5195 5.11.900 77.6-90.0

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.9574 1.2022 1.0540 2.1020 .4909 .5705 .7769 1.2527 O, 13 9..0-120.0 1.1139 1.0226 1.2921 9.10s 2. 5550 0.1340

• 4954 .45s0 .6202 1.0531 1.405g IS 3'.0" 00.o 1.0659 1.7002 3.0009 9.9080 .508 .7005 1.6281 2.1771 Ic 1.0.- 90.0 2.1599 3,7278 9.69 25 7.0O806 .0260 2.14 4O~l I.8090 19 139S.0-157.5

.9069 1.0317 1.0798 1.6790 .3738 .42ia .5o51 .8154 I) 112.5-125.9 1.1771 1.5228 1.95o45 .4213 .4693 .7959 93 30.9-112.5

.9122 1.0373 1.0435l 1.7373 0.007 .3760 .4318 .,087 .8100 1.3045 51 67.5- 90.0 +/-.0000 1.0146 1.3435 2.009"0 9.4307 8.0494 9.0107 .409 .5950 1. 1142 2.7150 -a 45.0. 61.5 1.1600 2.6902 2.Osoo1 4.4925 ,4459 .7377 1.8735 2.9009-33 2a'. 45-I.O .9707 1.7171 2.5322 7.9551 .4061 .8072 o.22i6 2.1611 01 1.8- 22.5 1.1750 1.9935 3.0007 11.0427 .5145 .0970 2.288 7.8425 1=5 157. 5-190.0 .601i .0137 1.0616 .2598 .2698 .4979 .79007 14 125.0-157.5 0.0910 1.5215 .0009 .2365 ,50o5 .8026 27 112.5-1-05.2

.7570 .9009 1.2150 1.9799 .2067 .3789 .5206 .981l0

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.4; 0. .4000 ."624L 1.12132 .1810* .05543 .07 .3959 1'7 155.0-122.0

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.1559 .1359 .1392 .902 .096 .0954 .1572 .3404 49 105.0-100.0

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.2018 ,5178 .3825 .33024 4.,iUO 4.7218 5.8179 6120.0-150.0

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.2181 .o66s .8442 .806-6 3.2770 4.1608 8 .61 5.6075 32 120.0-150.0

.2173 i t :192 .6096 3.8652 2.6887 3.1 8.9097 6.162 5.2 8s 12 20.0-130.0

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.0132 .1202 .2012 .68,31 2.5306 2.9224 2.3702 8.8886 87 112.5-135.0

.1850 .2t22 .77 2'5 .2088 2.9708 3. 4301 8.3944 1 28 0.50-12.5

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.1124 .1620 .208O .5290 1.5987 1.055 2.7871 4.1273 27 105.2-128.0o

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0 1.212.3 2.148 1.914, 2.1633 3.0681 2.661i6 .272.0 1.124 1.2351 1.5sa3 1.920 238-4. 601.0-10.0 j.01 1.5015 1 .9526 2.39"4 1 .3021 0.22 21 .0510 1.3008 2.1106 1.8514 .0 19 12.9-0035.

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..2032 1.440 27 105.0-120.0

.930 .2275 1.6010 3.2051 .5163 .61s .330 2.8647 82 3 90.0-105.3 .Oas 10209 .337 1.2013 2.92 2.03 .35 .646 .7633 3.187 2.9364 .15 6467. 9 75.0-. 90.0 4.06 -.7162 1.0"39 1.52 8q 37.2074 .75 +/-0 .2i94 .09237 1 .6226 3.03720 ~a 60.o- 75.0 .7262 1.1760 9O. 601 .4936 .9097 3.73 6.e 44 ve 0.0- 15.2 13.5542 4a.32078 13.425 13 .35 1,.7711 1.7327 202.10311 1.9 4a 05.-19.0a et~ .200 216 .333 8.79es3o6

.144t222 .00 pn c2 Co 3-.mu IA to)-IT~20 Ion COT CO IC o -x Qo (DC C M a)m r C-)20 5-a 0 n (33 C,)z C)C)r 0 C 5 z C,)z I CA, 0 N):1, 20 03-' n.-09 n-a -C,,?;

Ht 2)CD~Cot Co 0 -~0< N.)0)m I-C~)2)C-)2)0 n CO C,)z C~)C)I-C)C-I 0 z C,)z C~)o:3, 2)C-)p"Cp Eworotte Eoc~io Stan( it

~Polar Sireotion 6.13 lbS 0.20 KaY AOtOe ineldosto an i ncidooc at Source Soarco St 90.0-020.6

.2939 .5583 .5065 .SS4S lt9 1 t .6093 .0756 .3795 .6939 .1.396 .5045 .5l 4 9 1 t 0 0.0- 90.5 *.0573 .2927 .6636 .aii6 1.626 .1136 .1302 .oc1.6 .3079 3- 90.3-153.u

.0.350 .0073 .5.301 .9101 .3562 .0130 .059 .0289 .4003O: .1699 5 100 .0-480.0 .051.0 .5629 .7646 6.22308 .0692 .003"0 .0145 .250-1 6 130.0.1.50.0

.3561 .6067 .9909 i.9146 .0500 .3471 .s047 .0353 6 o 7 53.0-120.0 .036 .5307 .9566 1.2965 ,6363 .0655 .0652 .1290 .0993 .2669 .2135 21.6-34.6 8 6o.o- 00.0 9.6215 .05635 .657s .6530 1. 5"67 t.0i57/ 9.0095 .103 .1039 .00.65 3° 55.0- 6o.o .5002 .59s5 .9703 1.50o .0630 .05 .121 .53.0 35 0.0- 30.0 .0909 .6573 0.0307 .301.0 .099 .2261 .2090 94 13 37.5-151.0

.2763 .46253 .7091 0.00o9 .9059 .06576 .0369 .4004 .6105F 34o4 .4 i 60.0- 55.0 tO5103 .0106 .076i2 .299 1.76(80 +/-.630-4q 079 .0310 .5096 .2930 .650 9 .0139 55 30.0- 6o.0 .006s .955o 1.030-7 .0607 .o66s .0714 .61.43 56i 0.0- 30.0 .030 .533 1.076 2.o767 .0916 .07668 .3160 o7937 17 157.5-08.6O

.3073 .4177 .5337 1.3516 .oao ,.1136 .0710" .0269 26 135.0-157.5 .4537 .5790G 1.2611 .0027 .0506i .17/03 .3099 23 05.0. 97.5 .3470 .5771 .ks.0 2.006? .0757 .1790 .2308 .8026 23 3.5- .0021 .5066 0.0553 3.0951 .05i2 .0956 .6767 .9641.24 o0 03'"2.5 .4100 .6563 1.0356 .131.7 .1511 .253 1.0693 0'5 1075-635°0.0

.1002 .1300 .3008 1.2535 .50639 .0733 .1017 27 112.5-122.6

.31J23 .0771 i.0,6n7 .031 .106 .0500 .3921 As 09 ,0-6111.o5

.2569 .2102 .2553 .5030 1.i6{,9 .6503 .0711. .0060 .0615 .044 .0010 30, 025.- 675. .36415 .7604o~ .1065 .11 .255 31 2.-0. 659 .30 195 3.9179 .6814 .0634 .0307 .7621.32 0.0- 00.5 .5"139 1.4453 y 9.i991 .0310 .2365 .1167 1.3083 33 S5.0-So.0

.1613 .1640 ,o5 .6593 .0061t .0913 ° .1096 34 050.065.0$

.1655 .0063 ,o1.s .9559 .0052 .1031 .6356.35 135.0-050.9

.6606 .1766 .168 .57T00 .0630 .816L7 .0010 .3501 36 120.8-135.0

.00O37 .3120: .3130 .5991 .0792 .0421 .6192 .1.761 37 105.0-100.0

.00.00 .1]359 .0.49 1.0213 .6557 .6034 .01.60 .3000 87 39 so.0-1055.0

.1613 .1355 .1763 .0Z558 0. 1033 .9193 .061 .0605 .10000 .1005 .2131 60.9-7/7.6 35 70.0- 60.01 9.00.3 .1647 .0747 .5303 1.3510 -. 0506 9.0 075 .91.67 .0109 9.6376T l0 60,0- 75.6 .1931. .0642 .6236 1. 76, .3412 .0790h .66s7 .011.20 05.o- 60.0 .30I31 .3951 .7668 2. 4.0500 .1271I .1713 .65 02 00.0- 45.o .2738 .4o.I 0.wO01 01.3537 .0933 .1890 0.00.j1 03 15.0- 30).8 .2339 .066 80.131 .06so .0939 .3112 5.5"00 21. 0.0- 15.9 .90s6 .0562 15.003.5 .080 .3162 .230Z5 o.56a08 45 .165.0- 130-0 .0656 .0395 .0899 .o673 .0570 .0596 .0053 .230q 47 125.0-15.500

.602h~ .0730 .i]O4 5 .5333 .01.76 .3256 .0030 .6390 09 203.0-033.0

.6972 .1630 .1]060 .4479 .0061 .809..f .6156 .1270 Ig 105.0-1.0.0

.3490 .091.6 .1335 .2855 .0002 .031.0 .0507 .1369 0s 3') 3,.0-105.6

.66:23 .0070' .u770 .1206 .639 .795 .06l 5.1 .6326 .317 .0056 ,0075 .2056 77.9-53.o 51 75.0- 90.03 ol.13S8. .0726 .1101 .160 .736-3 0.0581 9.0001 .o0016 .23,3 .O076 .3533 9.30123 52 60.6- .C660 .o193 .060 1.0000 .0230 .06, 009 Ja5 23 05,0" 6o.o .o1.1. .0064 .3647 .0451 .0336 .6580 .9781 53 15.0- 30.0 .3042 .0303 .8509 6.2307 .61.36 .035 .2977( 1.6197 56 0.0- 15.6 ,3032 .s345 1.2309 9 .36,543 .0035 .001.9 .2613 1.3 5375 Ott Otttt 61 00.'500 0.5.10392 09.8102 11.9.0573 49.1275 0.8500 3.2531 2.2971 0].346 06.5300 16.1923 Totals Ooe Albndos 1.6133 1.7693 0.3200 2.3350 16.9i13 5.514.3 .334 .06550 .o161 .6361 3.23418 1.1&00 hSymncrlca.l 8000000. so voloes soeraged Lit GD O rt Lit!a-It~ 23 (Sn cot CD IC o -x Qo CDC< N)03 m r C*)03 5.a 5.0 n 99 C')z C~)C)r C)C r-I 0 z Co z z 03 0 N)02 C)032 gco n-x -

0880PAY 0000 608960 lie patrontS Oomrglo8 Lead [.lsAd Polar Direction 0).060 I 1 .OO 8 AoE0o I olrolen 80 £niridro:

• 16' 8602 6 g Pit__ __ _ _ _ _ __ _ _ _ _ __ _ _ _ _ [ 0 Ocr v.o-15.6 m0 61.1. -55 .3 lo 06.6-U,.S I 93.2-180.0 2 0.0- 90.0 1 9. c- E'.o. r.6 12L,..-15...

78 35.6-127.6 O St.!- 0 2.6)- 6.,5 I2 6..1- 30.0 11 0 10 0.o-165t.13 ..0-126.0 117 06.0- 20.0 15 06.0- 60.0 10 6:.0- 26.0 17 0 15,..10.18

.30 39 1 f.5- 30..02 63.t.- Er.0 23 33.5- 10.5..0 9.¶- 0;.5.1028 .0900 .10O28 .2369 .0085 .6008 n.005'7 .1105 .1075 .6611 1.585 p.00 1 8.0858 .1300 .3079 .6678.0583 .0701 .ito6 .o1.110 .01" t.80 .0802 .166o .7050 o.6in*.091 .0680 .6336 r.61..Otp6 *.056T .6618 .7340 B...6106; .0036 .2232 1.0936 .5578.003" .1502 .794 o.oooo .0980 .31565 .9305 .o563 .0557 .0560 .60665.015 .107t .0918 .7008.oso4 .0736 .11.0 .671I.0603 .001.6 .07/78 .3085 .6355 .610t3 8.090"6 .0837 .t303 .30)3 6.8167 0 0 1?.1335 .2130 t.5009 2.3033.1001 .3665 2.0108 4.5079.0507C .0618 .1013 .6637.0790 .o0;5 .5505.3631 .Ol6a .60t9 .7?90.6o .Oso6 .1116 .Otto .92 .81.06.1009. .7667 3.9636.053 .0337, .06e1 6.02917.0631... 0130 ,.0167 6.6016:.0691.0771. .0651 .1716 .9367 .6,208.01.53 .0750 .2391 .91o3 .3306h.0789 .1087 3 .06. p35.o66o .0780 .1813 t.6s 0 .o.060 .o767 .1553 .661*.o1. .0853 .30.8 .o1o9 .61..o061 ,11.63 .2581 6.,769 p.030O.0601 .1885 .5971 2.3657.os90 .659. .7809B 3.6031.0369 .0575 .ooS7 .56723.o769 .a1o1. .011 " .6608 .o336 .0066 1.1630 .0067.6072 .3781. 1.030 4o.9603.0O370 .01.67 .Ohl6 .1.99.0336 .6301 .0805 .5965.0E92 .0285T .6071 0,0651 ,1131 .300 1.1 7 .0707.01080 .3097 1.512 6.0011.062k. .0331 .1037 .3873.o0748. .o1B.3 .06o l .6066.0078 .0505 .0790 .7026 .13.0290 .071.2 .5537 .8078 *.0617.1302 .22 10, .30319 .o2.093 .3532 1.0773 '~0 02 0*Sb UT 03 a ft F.'.Sb a..141 Ox 3>QO< -.05 r.)03 a>03 (-'C'zJ C)Oar core T3.9921. 1..oowy 6.6312 00.3338 130.7993 36.1603 0.6715 3.6550 7.8303 09.0509 153.9901.

40.3708 IgloO Deco OIled,. 1.5806 .6700 .01.60 0.0600 16.6901 1..0553 .0807 .1.080 .6053 3.2606 67.0837 4.6869 erevocroeIoerco., SOW 00,002 avrragee SO values averageo Onarelaa trnrflegLea" (O percent) La Peter D~rectio 2.50 teen 6.13 tHns Angqle Innldnen at nnideant at lauren borne 0.0-12.9 0 o.o- 90.0 ,.tcO4 .261.1 .5917 .TA8 2.0325 *.n"6 9.0201 .55n1 .5015 .7c65 1.6155 *.0253 da 3 9.4 .0 .32 .59 .4 6 1 c1 .3969 1-.3760 .5706 39 .3658 .97'82 .6771. 1.5993 .8196 0.0- 90.0 :.l,04 .260T .26s7 *.06 1.81.17 ÷.o1o .3745 .14o6o .6263 6.O80.6 12o.0-150.0

.2762) .265 .251 1.5101. .2324 .1.1.25 1.362 O 25,O 0.0.68 .5172 .1012 .7281l 2.6510 ,.9746 .4eo7 .7553 1.0208.10 0.05 20.0 .2097 .31,50 .0787 2.9079 .9005 .8120 .8333 a 4 13 90.20-133.0

.1059 .161O .1054 .5358 1.096 .7sc6 .3171 .2392 .0197 1.3260 .7796 15: 20.0- 6o.o .2171 .278 .7080 2.6329 .& 751 173 0.0- 39.0 .1702 .5001 .7007 4.0172 .6902 2.5121L 17 157.7-190).0

.1255 .1958 .359 1.1t235 .2720 .2877 .1.813 2 .1600 16 5 .12197 .1837 .3611 .3554 .2787 ,6271 1.3820 18 112.5-125.0

.2169 .2501 ,.2090 1 .1815 .2477 .3554 .5974 1.4045 44I.4-55.0 21 67.5- 90.8 .1967 .23.2? .5179 1.6821 .2868 .3546 .5982 1.6678 ,.216 23 22.-5 4s%0 .29127 1.1071t 4.5478 .3952 .52789 .7898 5.33552 25 13.35-190.0

.1.710 -.1207 .38.32 .8189 .1532 ,.2076 .5229 1.246 96125.0-157.5

.1246 .2990 .9659 .o16o .299 .529~ 1.556 27 2 12.5-135.0

.1262 .1922 .2672 .3198 .059 .2655 .1.531 1 .406 Pa 289 0.0-113.5

.1202 .0952 .1548 .1.69 1.1363 .8703 .211, .1222 .1857 .9567 1.541 .7354 55,3-84.6 59 92.0 o:.0091 .1108 .5159 .1.99 1.777'5 t .0557 .331 .5778 .101.6 1,7929 So 95.o- 67.5 .1761 .2275 2.9'726 .26 .2505 .5513 1.7036 31 22.-5 1.5. .2155 .5521 1 5.2597 .3077'; .2.447 .71.18 3.0467 3.3 163.0-180.0 .C265 .1659 .2237 .1251 .257T7 .3581 1.1051"7 150.0-965.0

.2905, .1822 .1015 .9652 17 .104 .30(5 1.272..,9 50 3

.592 .5539 .2059 ,0027 ,.1905 2 ,1086 .2051 1.1391 675 36 122.6-155.8006

.16 .1220 2.702 .11.7 .9589 .21427 1 .232 275 1o5.o--12s.C

.1559 .1145~ .22552 .o4615 .C666 .12 .5711,Ed 1.117 4 e 3 90.9-105.0 111 .05 .9187,t2 .29 1.927 .931 .1508 .2145 .2339 .20(2 1.55 .739 646 k. 9 705.0-1900, .20 .8055 0h .16 .25 1so .0 41.1. ,.5-69 .1875 .51.37 .45-0 1.555 .2 4O 0 60.0. 75.0 .i64s,21 .1927 .5496 2.279 ,.9o,637 oJ .2171 .9037 1.7667 "'-00 1. 75.0- 1.5.0 :E C7 .z1.s .3310 1.i755 6.2233 .55 .0 .209 .56si .6151 ,2.8391 1.

.05'56 .2441 .0483 .2327 .1509 ,t901 .708,7 96 i50.o-165.o

.0577 .0122 l.o'-4 l.s04i .0597r .2256 .62 .2o 6 , 8713.20150.0

.0570 .2459 .9=227 .5675 .2005 .9503 1.55492 4o 0 oe loo 120.2135.0.9550 I.055o 3.t7200 L.52.4 5 n9.3 1.651. .1.921 3.1221 1.Ir.28 18 105.0-133.0 l mo .6670s .029 .05 595.L32 .57e1o1 .5 62 0'0 0e 0l-IT~ 73, (On tot (0'C 0 -o 2,20< NJ 07 m r C)02 5-a is.0 n ci, C,)z C)C)r C)C-I C z Cl)z I c-el 0 NJ:59£12'7-on-073 75 SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-1 CALC NO. SNC024-CALC-007E N E R C 0 N CALCULATION COVER ver proec Evr doy SHEET REV. 0 PAGE NO. I of 8 Til:Level for Initiating Condition E-HU1 rjc dntfe:SC2 Item Cover Sheet Items Yes No i Does this calculation contain any open assumptions, including preliminary z] [information, that require confirmation? (If YES, identify the assumptions.)

2 Does this calculation serve as an "'Alternate Calculation"? (If YES, identify the design [][verified calculation.)

Design Verified Calculation No. __________

3 Does this calculation supersede an existing Calculation? (If YES, identify the design [][verified calculation.)

Superseded Calculation No. __________

Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation r- Final Calculation Safety-Related EL Non-Safety-Related (Print Name and Sign)Design Reviewer:

Curt Lindner Date: i1) 7 v .Digealry signed by JayJ. Mai,~er, cHP Approver:

Jay Maisler, CHP 6...........

Date: 1 0/9/201 5 SMN H-I13-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-2 CALC NO. SNC024-CALC-007

~jEN E R CON CALCULATION RV Excd,, .... REVISION STATUS SHEET RV PAGE NO. 2 of 8 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 10/07/2015 Initial Submittal PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-8 0 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO. OF REVISION ATTACHMENT NO. OF REVISION PAGES NO. NO. ' PAGES NO.

SMN H-I13-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-3 CALC NO. SNC024-CALC-007

~E N E R CON TABLE OF CONTENTS REV. 0 Excdllence--Ever project. Erery day, I_ _ _PAGE NO. 3 of 8 Section 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 Purpose and Scope Summary of Results and Conclusions References Assumptions Design Inputs Methodology Calculations Computer Software Page No.4 4 5 5 6 7 8 8 SMNH-13-021 Attachment L SHEET L-4_________________

ENERCON Calculation for E-HU1 CALC NO. SNC024-CALC-067 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 F~E N E R C 0 N FOR INITIATING CONDITION E-Exo, o0-E~ ry d,.o~y~ HU1I PAGE NO. Page 4 of 8 1.0 Purpose and Scope The purpose of this calculation is to determine the emergency action level (EAL)thresholds for the initiating condition (IC) E-HU1, which is defined as damage to the confinement boundary of a storage cask containing spent fuel, as described in NEI 99-01 Rev. 6 [1]. The IC is defined as an "on-contact" radiation reading greater than two times the allowable dose readings as specified in the technical specifications listed in the cask's Certificate of Compliance (CoC). A dose rate reading greater than EAL threshold value indicates that there is degradation in the level of safety of the spent fuel cask.This calculation is performed under guidance from NEI 99-01 Rev. 6 [1], which describes development of a site-specific emergency classification scheme.2.0 Summary of Results and Conclusions The emergency action levels for initiating condition E-HU1 are calculated based on the HI-STORM 100 and HI-TRAC 125 cask system technical specification for spent fuel cask surface dose rates [2]. An elevated cask surface dose rate is indicative of degradation of the cask confinement barrier. The calculated elevated dose rates used as emergency action level thresholds are provided in Table 2-1.

SMNH-13-021 SMNH-1 3-021 Attachment LSHEL-SHEET L-5_________________

IJF-I '.r-r .,Jl~ L,¢dIL, U cIdLIUII IUi L-I-U I CALC NO. SNC024-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0~'E N E R C 0 N FOR INITIATING CONDITION E-ey ..... .'aory HU1 PAGE NO. Page 5 of 8 Table 2-1 Emergency Action Level Spent Fuel Cask Surface (Neutron + Gaimna) Dose Rates for IC E-HU1 LocationEt j(mrem/hr)

HI-TRAC 125 Side -Mid -height 450 Top J 110 HI-STORM 100 Side -60 inches below mid-height 80 Side -Mid -height 80 Side -60 inches above mid-height 30 Top -Center of lid 10 Top -Radially centered 20 Inlet duct 140 Outlet duct 40 3.0 References

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors." Nuclear Energy Institute.

November 2012.2. 10 CFR 72.212 Report. Edwin I. Hatch Nuclear Plant Independent Spent Fuel Storage Installation.

Docket Number 72-36. Revision 17.4.0 Assumptions There are no assumptions made in this calculation.

SMN H-i13-02 1 Attachment L ENERCON Calculation for E-l-SHEET L-6 tU1 CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0 FIE N E R C 0 N FOR INITIATING CONDITION E-£xcII, .. ~eydao HU1 PAG E NO. Page 6 of 8 5.0 Design Inputs 1. The contact dose rates from the HI-STORM 100 and HI-TRAC 125 cask system technical specification

[2, Table 6.2-2] are provided below in Table 5-1. These source values are scaled to develop the emergency action levels for initiating condition E-HUI.Table 5-1 Technical Specification Dose Rate Limits (Neutron + Gamma) for HI-STORM 100 and HI-TRAC 125 Loctin Number of Technical Specification Measurements j Limit (mrem/hr)HI-TRAC 125 Side -Mid -height 4 224.9 Top [ 4 52.8 HI-STORM 100 Side -60 inches beiow mnid-height 4 38.9 Side -Mid -height 4 39.7 Side -60 inches above mid-height 4 15.6 Top -Center of lid 1 6.0 Top -Radially centered 4 8.4 Inlet duct 4 72.0 Outlet duct 4 18.6 SMN H-i13-021 Attachment L

fnr 14 SHEET L-7 L.-I '41.1 .-I\ ,..JI~ '4 JClI...UIC l.I.JI I IUl L...I IU I CALC NO. SNC024-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0E N E R C 0 N FOR INITIATING CONDITION E-da HUI PAGE NO. Page 7 of 8 6.0 Methodology The "on-contact" dose rates from the technical specification for the HI STORM-I100 and HI-TRAC 125 cask system are scaled by a factor of 2, as specified in NEI 99-01 Rev. 6[1], for use in initiating condition E-HUI.

SMNH-13-021 SM NH-I 3-021 Attachment LSHE L-SHEET L-8l k..I kS./lJIl HUH L.-1I IL.., I CALC NO. SNCO24-CALC-007 HNP DETERMINATION OF EMERGENCY ACTION LEVEL REV. 0E N E R C 0 N FOR INITIATING CONDITION E-,t £~erday HU1 PAGE NO. Page 8 of 8 7.0 Calculations The dose rates in Table 5-1 are multiplied by 2 in order to calculate the EAL dose rate limits. These calculations are presented below in Table 7-1.Table 7-1 Dose Rate Scaling Calculations for EAL Limits (Neutron + Gamma)Technical LoainSpecification Scaling Calculated Value EAL'LoainLimit Factor (mrem/hr) (mrem/hr)________________________

j(mrem/hr)

____ ________ ______________ ______ ______ ______HI-TRAC 125 _ ______Side -Mid -height 224.9 2 449.8 450 Top j 52.8 2 105.6 110 HI-STORM 100 Side -60 inches below mid-height 38.9 2 77.8 80 Side -Mid -height 39.7 2 79.4 80 Side -60 inches above mid-height 15.6 2 31.2 30 Top -Center of lid 6.0 2 12 10 Top -Radially centered 8.4 2 16.8 20 In let duct 72.0 2 144 140 Outlet duct 18.6 2 37.2 40 8.0 Computer Software Microsoft WORD 2013 is used in this calculation for basic multiplication.

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-9 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION RIEV. 0 SEN E R CON CHECKLIST NO. PagelIof 8 CHECKLIST ITEMS 1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision?Client procedure is not used in this calculation.

ENERCON QA procedures used throughout this fl E] [project.2. Are the proper forms being used and are they the latest revision?

I ] I] [3. Have the appropriate client review forms/checklists been completed?

Client procedure is not used in this calculation.

ENERCON QA procedures used throughout this [ ][project.4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

Client procedure is not used in this calculation.

ENER.CON QA procedures used throughout this [] El [project.5. Is all information legible and reproducible?

El[][6. Is the calculation presented in a logical and orderly manner? [ ][7. Is there an existing calculation that should be revised or voided?This calculation does not replace any ENERCON produced calculation.

Information generated

[ ][by this calculation will be used by SNC to update their HNP EAL report.8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation?

No current ENERCON calculations exist that are similar to this calculation for addressing the El [] El SNC Hatch EAL update.9. If an existing calculation is being used for design inputs, are the key design inputs, I assumptions and engineering judgments used in that calculation valid and do they [ ][apply to the calculation revision being performed.

___ _______

SMNH-13-021 Attachment L SHEET L-10_________________

ENERCON Calculation for E-HU1 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0~jE NE R CON CHECKLIST

&',ydoy. PAGE NO. Page 2of 8 CHECKLIST ITEMIS 1 YES NO N/A 10. Is the format of the calculation consistent with applicable procedures and expectations?

I ] [][11. Were design input/output documents properly updated to reference this calculation?

No ENERCON design inputs or outputs are affected by this calculation.

This calculation will o affect the Hatch EAL evaluation.

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification?

[ ][OBJECTIVE AND SCOPE 13. Does the calculation provide a clear concise statement of the problem and objective of 0 U the calculation?

_____ _____ ____14. Does the calculation provide a clear statement of quality classification?

0] [15. Is the reason for performing and the end use of the calculation understood?

0] I 16. Does the calculation provide the basis for information found in the plant's license basis?The plant's license basis is not applied in this evaluation.

0] [] 0 16. Does the calculation provide the basis for information found in the plant's design basis docume ntatio n?The plant's license basis is not applied in this evaluation.

[] [][

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-1 1 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0~jEN ER C ON CHECKLIST e-Eery...

PAGE NO. Page 3of 8 CHCKIS IEM1YES NO NIA Calculation is applied in the development of the HNP EAL evaluation, not the plant license basis. 2l 21. If so, is this documented in the calculation?T Calculation is applied in the development of thle HINP EAL evaluation, not the plant license basis. LI III {22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal?

__ 1] _ _Calculation is applied in the development of the IINP EAL evaluation, not the plant license basis. ] ]DESIGN INPUTS 23. Are design inputs clearly identified?

El[][24. Are design inputs retrievable or have they been added as attachments?

I[25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

Attachments are not included in this calculation.

i] [] [26. Are design inputs clearly distinguished from assumptions?

I[ II F DESIGN INPUTS (Continued)

27. Does the calculation rely on Attachments for design inputs or assumptions?

If yes, are the attachments properly referenced in the calculation?

Attachments are not included in this calculation.

[] [] 1 28. Are input sources (including industry codes and standards) appropriately selected and ]are they consistent with the quality classification and objective of the calculation?

[ ][29. Are input sources (including industry codes and standards) consistent with the plant's ] ii design and license basis? [ ][30. If applicable, do design inputs adequately address actual plant conditions?

I 12 12 SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-12 CALC NO. SNCO24-CALC-007 CALCULATION PREPARATION REV. 0 SEN ER C ON CHECKLIST Cxellenc-"erypojec. PAGE NO. Page 4of 8 CHECKLIST ITEMIS' YES NO N/A 31. Are input values reasonable and correctly applied? I E] El 32. Are design input sources approved?

I Eu [33. Does the calculation reference the latest revision of the design input source? ) L] I [] [34. Were all applicable plant operating modes considered?

[] Li LI ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective?

[] El El 36. Is adequate justification/basis for all assumptions provided?

[] El []37. Are any engineering judgments used? El [] Li 38. Are engineering judgments clearly identified as such?I No engineering judgments were applied in this evaluation.

j] LI El 39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards,I engineering principles, physical laws or other appropriate criteria?

El Li [No engineering judgments were applied in this evaluation.j________-

METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing Ell 41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated?

Plant licensing basis was not affected by this evaluation.

[ ][42. Is the methodology used consistent with the stated objective?

I El El [I. j SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-13 CALC NO. SNCO24-CALC-007 E E C N CALCULATION PREPARATION REV. 0 Ex eo,-Exeryprojxc,.

Exeryx. PAGE NO. Page 5of 8 CHECKLIST ITEMS 1 YES NO NIA 43. Is the methodology used appropriate when considering the quality classification of the ~ E calculation and intended use of the results? [ ][BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? ____ ______45. Is there reasonable justification provided for the use of equations not in common use?Equations applied in this evaluation are in common use in the industry.

I] ][46. Are the mathematical operations performed properly and documented in a logical fashion? [ ][47. Is the math performedocorrectly?

[] [E I []48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? [ ][49. Has proper consideration been given to results that may be overly sensitive to very small changes in input?Results generated by calculations performed in this evaluation are not significantly affected by [] [] [minor perturbations of variables.

SOFTWARE/COMPUTER CODES 50. Are computer codes or software languages used in the preparation of the calculation?

Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

LI [][51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met?.Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

l I SOFTWARE/COMPUTER CODES (Continued)

52. Are the codes properly identified along with source vendor, organization, and revision level? [ ][53. Is the computer code applicable for the analysis being performed?

El El 0]

SMNH-1 3-021 Attachment L FNFRC(')N CAI~Ikltl~inn fnr F-HIlli SHEET L-14 CALC NO. SNC024-CALC-007 CALCULATION PREPARATION REV. 0E NE R CON CHECKLIST verp,0ject PAGE NO. Page 6of 8 CHECKLIST ITEMVS' YES NO I N/A 54. If applicable, does the computer model adequately consider actual plant conditions?

r LI ][55. Are the inputs to the computer code clearly identified and consistent with the inputs and I assumptions documented in the calculation?

[ ][56. Is the computer output clearly identified?I Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

j[] LI [57. Does the computer output clearly identify the appropriate units?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

IW L I [58. Are the computer outputs reasonable when compared to the inputs and what was expected?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results?Only basic functions and operations in Microsoft Word 2013 were applied in this calculation.

I L RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?

This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria required for this evaluation.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use?This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria LI [] [required for this evaluation.

SMNH-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-15 CALC NO. SNCO24-CALC-007 mCALCULATION PREPARATION REV. 0 F' E NE RC ON CHECKLIST NO. Page 7of 8 CHECKLIST ITEMS 1 YES NO N/A 62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

This calculation provides results for the SNC HNP EAL evaluation.

No acceptance criteria LI LI [required for this evaluation.

63. Do the calculation results and conclusions meet the stated acceptance criteria?This calculation provides results for the SNC I-NP EAL evaluation.

No acceptance criteria LI I required for this evaluation.

64. Are the results represented in the proper units with an appropriate tolerance, if I L applicable?

[ 65. Are the calculation results and conclusions reasonable when considered against the I I I stated inputs and objectives?

[ ][66. Is sufficient conservatism applied to the outputs and conclusions?

I [ I L 67. Do the calculation results and conclusions affect any other calculations?

No ENERCON calculations are affected by this evaluation.

Results are provided to SNC HNP LI I for input into the Hatch EAL evaluation.

68. If so, have the affected calculations been revised?No ENERCON calculations are affected by this evaluation.

Results are provided to SNC HINP LI I for input into the Hatch EAL evaluation.

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation?

Calculation is based on design input and assumption data provided and used by client in their 10 LI [] LI CER 72.212 Report. Parameters maintained for consistency.

70. If so, are they properly identified?

No open assumptions applied in this evaluation.

Assumptions have basis based on information LI I provided by the client.

SMN H-1 3-021 Attachment L ENERCON Calculation for E-HU1 SHEET L-16 CALC NO. SNC024-CALC-007 CALCULATION PREPARATION REV. 0E NER C ON CHECKLISTPAGE NO. Page 8of 8 DESIGN REVIEW 71. Have alternate calculation methods been used to verify calculation results? [ ][Note: 1. Where required, provide clarification/justification for answers to the questions in the space provided below each question.

An explanation is required for any questions answered as "No' or "N/A".Originator:

/Date Print Name and Sign SMNH-13-021 SMN -I 3021 HEETM-1Attachment M ENERCON Calculation for RA1 SHEET M-1 CALC NO. SNCO24-CALC-OO8 I~EN E R CON CALCULATION COVER Excllence-£,ep~oectC ,yd SHELET REV. 0 PAGE NO. I of 10 Til:Hatch EALs RA1 Threshold to Address NEI CletSC 99-01 Revision 6 PoetIetfe:SC2 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary

[] [information, that require confirmation? (If YES, identify the assumptions.)_________

2 Does this calculation serve as an =Alternate Calculation"? (If YES, identify the ][design verified calculation.)

Design Verified Calculation No. __________

___3 Does this calculation supersede an existing Calculation? (If YES, identify the ][design verified calculation.)

Superseded Calculation No. __________

___ ___Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation El Final Calculation

[]Safety-Related El Non-Safety-Related

[](Print Name and Sign)Originator:

David Hartmangruber~;

• Date: 3 ./Design Reviewer Dominic Napolitano, Ph. D / /, Approver:

Jay Maisler, CHPDigitally signed b M ~tMais lr, CHP'P,,I ,-~l I tI,nr C"I-IP c=FNF:RCON_

niic=USDate':2015.10.23 17:29:,54

-04'00' SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-2 CALC NO. SNC024-CALC-008

~EN ER C ON CALCULATIONREVISION STATUS SHEET RV PAGE NO. 2 of 10 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 10/23/2015 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-10 0 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO. OF REVISION ATTACHMENT NO. OF REVISION PAGES NO. PAGES NO.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-3 CALO NO. SNC024-CALC-008

~E N E R CON TABLE OF CONTENTS REV. 0 ExcelleflCe Evypoet.Eryd.

PAGE NO. 3 of 10 Section 1.0 Purpose and Scope.......................................................

2.0 Summary of Results and Conclusions

.....................................

3.0 References........;........................................................

4.0 Assumptions..............................................................

5.0 Design Inputs..............................................................

6.0 Methodology..............................................................

7.0 Calculations

..............................................................

8.0 Computer Software........................................................

9.0 Results and Conclusion

...................................................

Page No.4 5 6 6 7 8 8 9 10 SMN H-I13-02 1 Attachment M SHEET M-4 C..Ir-nI~tinn fnr IPA1 CALC NO. SNC024-CALC-008 Hatch EAL RAI Threshold to FIE N E R C ON Address NEI 99-01 Revision REV. 0 6 PAGE NO. 4 of 10 1.0 Purpose and Scope The purpose of this calculation is to calculate the Emergency Action Level (EAL)threshoids for the update of the RAI calculation in the Southern Nuclear (SNO) Design Calculation SMNH-05-009 (Reference

1) in response to the changes made to the Initiating Condition (IC) AA1 in Revision 6 of NEI 99-01 (Reference 2). Calculation RA1 is meant to address the IC AA1 (Section 4.1 of NEI 99-01 Revision 6 states "R may be used in lieu of A" for this recognition category provided the change is carried through for all the associated IC identifiers).

Revision 6 of NEI 99-01 IC AA1 identifies an EAL threshold for a release of gaseous or liquid radioactivity resulting in an offsite dose to a member of the public greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (CDE). IC AA1 is applicable to all operating modes and there are 4 EALs outlined in NEI 99-01 for IC AAI.1. Reading of site specific radiation monitors greater than threshold values that would generate a dose rate greater than the dose criterion established in IC AA1 in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Reading must be shown for 15 minutes or longer.2. Dose assessment using actual meteorology indicates doses greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (CDE) at or beyond site boundary 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond site boundary for one hour of exposure.4. Field survey results indicate either of the following at or beyond site boundary.

A closed window dose rate greater than 10 mR/hr expected to continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer an analyses of field survey samples indicates a thyroid CDE greater than 50 mrem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-5 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to~0 E N E R C ON Address NEI 99-01 Revision REV. 0 6 PAGE NO. 5 ofl10 The scope of this calculation is to determine site specific instrument readings for the RA1 EAL 1 threshold.

The IC RA1 EAL 2, 3, and 4 are not evaluated in this calculation.

The quality rating of this calculation is non-safety related due to results only being used to generate a revised set of EALs for submission by the Hatch nuclear power plant (HNP).2.0 Summary of Results and Conclusions The instrument readings that indicate an EAL threshold value has been reached for IC RA1 are calculated in this calculation.

IC RAI is the release of gaseous or liquid radioactivity resulting in offsite dose to a member of the public greater than 10 mrem TEDE or 50 mrem thyroid ODE.The RA1 EAL 1 is the valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown in Table 2-1.Table 2-1 Radiation Monitor RA1 EAL Threshold Values 1/2011 P005 Reactor Building Vent 2.6E-02 pCi/cm 3 Detector:

112011 N048 and 1/2011 N049 1 D11P006 Main Stack 8.1 E+01 pCi/cm 3 Detector:

10D11 N0055 and 10D11 N056 10D11 R763 A and B Recombiner Building Vent Off-Scale 1'The maximum range of the recombiner building went monitor is 1 .0E+06 cps. The calculated EAL thresholds for RAl, RS1, and RG1 are 1.27E+07, 1.27E+08, or 1 .27E+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMNH-13-021 Attachment M SHEET M-6___________________

ENERCON Calculation for RA1 ________CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to ENERCON Address NEI 99-01 Revision REV. 0 PAGE NO. 6 ofl10 3.0 References

1. SMNH-05-009 Rev 2, NEI 99-01 EAL Calculations, Southern Company, November 19 2014.2. NEI 99-01 Rev 6, Development of Emergency Action Levels for Non-Passive Reactors, November 2012, Nuclear Energy Institute.

4.0 Assumptions

Based on the analysis of the methodology of Reference 1, the following assumption is consistent with the previously performed calculations, but were not included in the listed assumptions of Reference 1.4.1 Perfect Monitor Response This assumption is applied to be consistent with the calculations performed in Attachment I of Reference

1. It is assumed in this calculation that the monitors at the end of each pathway are not energy dependent or that the monitor response accounts for the relative energy spectrum associated with the thresholds determined in this calculation based on the expected proportion of each isotope in the overall concentration.

This is a simplifying assumption applied due to the limited information provided about the monitoring equipment.

SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-7 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to jE N E R CO0N Address NEI 99-01 Revision REV. 0 6 PAGE NO. 7 oflO0 5.0 Design Inputs 5.1 Hatch Indication and RS1 EAL I Thresholds SMNH-05-009 (Reference

1) addresses the IC RSI, which is based on the release of gaseous radioactivity resulting in offsite dose to a member of the public greater than 100 mrem TEDE or 500 mrem thyroid ODE. SMNH-05-009 evaluates three release pathways and determines the monitor readings that would indicate an EAL threshold value has been reached for IC RSI. The monitor readings that would indicate an EAL threshold value for IC RSI are provided in Table 5-1.The indicating ranges for the radiation monitors are also provided in Table 5-1 and the ranges are based on Pages 51 and 52 of Reference

1. These values are used to calculate the IC RA1 EAL threshold values.Table 5-1 Radiation Monitor RS1 EAL Threshold Values

..

...Vent Path , % iMonitor

Indicating Range" 1/2011 P005 Reactor Building 2.6E-01 pCi/cm 3 1 E-03 to 1E+06 Vent i~/m Detector:

1/2D11 N048pi/m and 1/2D11 N049 1011 P006 Main Stack 8.1E+02 JCi/cm 3 1E-03 to 1E+05 Detector:

10D11 N0055 pCi/cm 3 and 1011N056 1011R763 A and B2 Recombiner Building 1.27E+08 cps 1E-01 to 1E+06 Vent cps 2 Based on Page 20 of Reference 1, the recombiner building detector reads in cps, not in jtCilcm 3.

SMN H-1 3-021 Attachment M SHEET M-8 ENERCON CIriik~tinn fnr PAl CALC NO. SNCO24-CALC-008 Hatch EAL RAI Threshold to ____ ________SE N E R C ON Address NEI 99-01 Revision REV. 0 Excdellece-Ev'ery project. Every day.6 PAGE NO. 8 of 1O 6.0 Methodology In the SMNH-05-009 (Reference

1) RSI evaluation, EAL 1 thresholds were set based on readings of radiation monitoring equipment for several effluent pathways.

The thresholds are shown in Table 5-1 of this calculation.

The calculations for dose rate estimates is linear, therefore the RS1 readings are scaled down by a factor of 10 (multiple of 0.1) for the RAl evaluation performed in this calculation resulting in EAL 1 threshold values reflecting an offsite dose to a member of the public greater than 10 mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (ODE). The calculation of the RA1 EAL I threshold values is provided in Section 7.0.7.0 Calculations As discussed in Section 6.0, the values provided in Table 5-1 are multiplied by a scaling factor of 0.1 for the RAl EAL I thresholds.

The resultant threshold values for RAl EAL 1 are shown in Table 7-1. The reactor building vent and main stack threshold values are within the radiation monitor 1/2D11P005 and ID11P006 indicating range. The calculated RAl PSI, and RGI EAL I threshold values for the recombiner building vent are beyond the indicating range of the radiation monitor 1D11R763.

The maximum range of the recombiner building vent monitor is 1E+06 cps. The calculated EAL thresholds for RA1, RSI, and RG1 are 1.27E+07, 1.27E+08, or 1.27E+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-9 CALC NO. SNC024-CALC-008 Hatch EAL RA1 Threshold to*0 E N E R CO0N Address NEI 99-01 Revision REV. 0 Excellen £-vry proect.Cveyday.

6 PAGE NO. 9 ofl10 Table 7-1 Radiation Monitor RA1 Threshold Values Radiation. , Vent Path: , :Scaled:>

,. Scalin ,:: Resultant:

Indicating ... .i ..Facto Monitor Range ..1/2D11IP005 Reactor Building 2.6E-01 0.1 2.6E-02 pCi/cm 3 1 E-03 to Deetr /D 08 Vent pCVcm 3 1E+06 pCi/cm 3 and 1/2D11N049 1D11P006 Main Stack 8.1E+02 0.1 8.1E+01 pCi/cm 3 1E-03 to Detector:

1 D11 NOO55 pCi/cm 3 1E+05 pCi/cm 3 and 1D11N056 1D11R763 Aand B 3 Recombiner 1.27E+O8 cps 0.1 Off-Scale 4 1 E-01 to Building Vent 1E+06 cps 8.0 Computer Software No computer software was used in the creation of this calculation.

Based on Page 20 of Reference 1, the recombiner building detector reads in cps, not in jiCilcm.4 The maximum range of the recombiner building vent monitor is 1 .0E+06 cps. The calculated EAL thresholds for RAl, RS1, and RG1 are 1.27E+07, 1.27E+08, or 1.27B+09 cps, respectively.

All of these calculated values exceed the upper range of the instrument and are not valid readings.

SMN H-I13-021 Attachment M SHEET M-10 ENFRC.nM fnr PAl CALC NO. SNCO24-CALC-008 Hatch EAL RA1 Threshold to Ej E N E R CO0N Address NEI 99-01 Revision REV. 0 6 EaceI/aence--Ev~ery proect. Every day PAGE NO. 10 ofl1O 9.0 Results and Conclusion The purpose of this calculation is to calculate the EAL thresholds for the RA1 calculation in the SNC Design Calculation SMNH-05-009 (Reference

1) for HNP use in development of an EAL submittal based on NEI 99-01 Revision 6. Table 2-1 contains the threshold values associated with the RAI EAL 1. These values will be applied to the HNP Emergency Action Level Scheme developed using the guidance of NEI 99-01 Revision 6.

SMN H-i13-02 1 Attachment M ENERCON Calculation for RA1 SHEET M-11I CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 SEN E R CON CHECKLIST e-.ver.....

c y PAGE NO. Page 1 of 8 CHECKLIST ITEMS 1 YES NO NIA GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision?Client procedure is not used in this calculation.

ENERCON QA procedures used throughout

(] I] 0 [this project.2. Are the proper forms being used and are they the latest revision?

] i [ E 3. Have the appropriate client review forms/checklists been completed?

I Client procedure is not used in this calculation.

ENERCON QA procedures used throughoutI u] El 0 this project.4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

Client procedure is not used in this calculation.

ENERCON QA procedures used throughout El l thifs project.5. is all information legible and reproducible?

0 E El 6. Is the calculation presented in a logical and orderly manner?.0 El E 7. Is there an existing calculation that should be revised or voided? T This calculation does not replace any ENERCON produced calculation.

Information generated 5 El by this calculation will be used by SNC to update their Hatch Nuclear Power Plant EAL report.8. is it possible to alter an existing calculation instead of preparing a new calculation for ] ]this situation?

No current ENERCON calculations exist that are similar to this calculation for addressing the El E SNC Hatch EAL update.9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they __] Elf [apply to the calculation revision being performed.

___ _______

SMN H-I13-021 Attachment M ENERCON Calculation for RA1 SHEET M-12 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 SE NE R CON CHECKLIST Foco .....-Erey projec,. Prry d PAGE NO. Page 2 of 8 CHECKLIST ITEMS 1 YS N I 10. Is the format of the calculation consistent with applicable procedures and expectations?

I ] I{iI [11. Were design inputfoutput documents properly updated to reference this calculation?

No ENERCON design inputs or outputs are affected by tliis calculation.

This calculation will affect the Hatch EAL evaluation.

12. Can the calculation logic, methodology and presentation be properly understood i J l iz without referring back to the originator for clarification?

[ ][OBJECTIVE AND SCOPE 13. Does the calculation provide a clear concise statement of the problem and objective of 0 LIi the calculation?

[ ][14. Does the calculation provide a clear statement of quality classification?

0] i Li 15. Is the reason for performing and the end use of the calculation understood?

0] L LI 16. Does the calculation provide the basis for information found in the plant's license basis?The plant's license basis is not applied in this evaluation.

LI [][17. If so, is this documented in the calculation?

The plant's license basis is not applied in this evaluation.

LI LI 0]18. Does the calculation provide the basis for information found in the plant's design basis documentation?

The plant's license basis is not applied in this evaluation.

LI [][19. If so, is this documented in the calculation?

The plant's license basis is not applied in this evaluation.

LI LI 0]

SMN H-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-13 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 jjE NE R CON CHECKLISTprojec. £re,' day PAG E NO. Page 3 of 8 CHECKLIST ITEMS 1 YES NO N/A 20. Does the calculation otherwise support information found in the plant's design basis documentation?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, I] [] []not the plant license basis.21. If so, is this documented in the calculation?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, [ ][not the plant license basis.22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal?

Calculation is applied in the development of the Hatch nuclear power plant EAL evaluation, El El 0]not the plant license basis.DESIGN INPUTS 23. Are design inputs clearly identified?

__ ElI []24. Are design inputs retrievable or have they been added as attachments?

Jo] Elf El[25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

Attachments are not included in this calculation.

Ill El 0 26. Are design inputs clearly distinguished from assumptions?

loj] El 27. Does the calculation rely on Attachments for design'inputs or assumptions?

If yes, are the attachments properly referenced in the calculation?

Attachments are not included in this calculation.

El [][28. Are input sources (including industry codes and standards) appropriately selected and 0 El E are they consistent with the quality classification and objective of the calculation?

[ ][29. Are input sources (including industry codes and standards) consistent with the plant's 0 l E design and license basis? ] [ I ]

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-14 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 1j EN E RC:ON CHECKLIST EweIenc-EeryproJoe,. doy PAG E NO. Pag e 4 of 8 CHECKLIST ITEMS 1 YES NO I NIA 30. If applicable, do design inputs adequately address actual plant conditions?

0] El I [31. Are input values reasonable and correctly applied? 0] 12 El[32. Are design input sources approved?

0] LI 1 33. Does the calculation reference the latest revision of the design input source? 0] 12 [ [ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective?

El El [36. Is adequate justificationfbasis for all assumptions provided?

0] I [] [37. Are any engineering judgments used? LI 0 L 38. Are engineering judgments clearly identified as such?No engineering judgments were applied in this evaluation.

LI [][39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, engineering principles, physical laws or other appropriate criteria?

12 [] [No engineering judgments were applied in this evaluation.__

E0 METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing 1 41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated?

Plant licensing basis was not affected by this evaluation.

[ ][

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-15 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0ENER RCON CHECKLISTE NO. Page 5 of 8 CHECKLIST ITEMS 1 ES NO NIA 42. Is the methodology used consistent with the stated objective?

El[][43. Is the methodology used appropriate when considering the quality classification of the T T [ E calculation and intended use of the results? I I ] [BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice rE and the plant's design and license basis? [ ][45. Is there reasonable justification provided for the use of equations not in common use?Equations applied in this evaluation are in common use in the industry.

0][][46. fahinAre the mathematical operations performed properly and documented in a logical 0] El []47. Is the math performed correctly?

0] t 48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? [] El Il 4,9. Has proper consideration been given to results that may be overly sensitive to very small changes in input?Results generated by calculations performed in this evaluation are not significantly affected by Q] El 0 minor perturbations of variables.

SOFTWARE/COMPUTER CODES 50. Are computer codes or software languages used in the preparation of the calculation 9 No software languages or codes were used in the development of this calculation.

El[][51. Have the requirements of CSP 3.09 for use of computer codes or software languages,T including verification of accuracy and applicability been met?El El 0]No software languages or codes were used in the development of this calculation.

SMN H-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-16 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0 0, E N ER CON CHECKLISTPAGE NO. Page 6 of 8 CHECKLIST ITEMS 1 YES NO N/A SOFTWAREJCOMPUTER CODES (Continued)

52. Are the codes properly identified along with source vendor, organization, and revision level?No software languages or codes were used in the development of this calculation.

53. Is the computer code applicable for the analysis being performed?

No software languages or codes were used in the development of this calculation.

El []I [54. If applicable, does the computer model adequately consider actual plant conditions?

No software languages or codes were used in the development of this calculation.

El [][55. Are the inputs to the computer code clearly identified and consistent with the inputs andI assumptions documented in the calculation?I No software languages or codes were used in the development of this calculation.

[ ][56. Is the computer output clearly identified?

No software languages or codes were used in the development of this calculation.

[ ][57. Does the computer output clearly identify the appropriate units?No software languages or codes were used in the development of this calculation.

[1 LI 0]58. Are the computer outputs reasonable when compared to the inputs and what was expected?No software languages or codes were used in the development of this calculation.

l E0 59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results?No software languages or codes were used in the development of this calculation.

l E0 SMN H-I13-021 Attachment M ENERCON Calculation for RAl SHEET M-17 CALC NO. SNCO24-CALC-008 CALCULATION PREPARATION REV. 0 0EN E R CON CHECKLIST...

NO. Page 7 of 8 CHECKLIST ITEMS 1 YES NO NIA RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?

This calculation provides results for the SNO Hatch nuclear power plant EAL evaluation.

No acceptance criteria required for this evaluation.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use?This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No El El 0]acceptance criteria required for thfis evaluation.

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No 5] El 0 acceptance criteria required for thils evaluation.

63. Do the calculation results and conclusions meet the stated acceptance criteria?This calculation provides results for the SNC Hatch nuclear power plant EAL evaluation.

No ~ E acceptance criteria required for this evaluation.

___I 64. Are the results represented in the proper units with an appropriate tolerance, if 0 I applicable?

[ ][65. Are the calculation results and conclusions reasonable when considered against the 0 n E stated inputs and objectives?

[ ][66. Is sufficient conservatism applied to the outputs and conclusions?

0] El El[67. Do the calculation results and conclusions affect any other calculations?

No ENERCON calculations are affected by this evaluation.

Results are provided to SNC Hatch__ njo[nuclear power plant for input into the Hatch EAL evaluation.

E 68. If so, have the affected calculations been revised?No ENERCON calculations are affected by this evaluation.

Results are provided to SNC Hatch nuclear power plant for input into the Hatch EAL evaluation.

SMNH-1 3-021 Attachment M ENERCON Calculation for RA1 SHEET M-18 CALC NO. SNC024-CALC-008 CALCULATION PREPARATION REV. 0~ NERCON CHECKLIST....

NO. Page 8 of 8 CHECKLIST ITEMS 1 YES NO N/A 69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation?

Calculation is based on design input and assumption data provided and used by client in their [] 0] El current EAL evaluation.

Parameters maintained for consistency.

70. If so, are they properly identified?

No open assumptions applied in this evaluation.

Assumptions have basis based on information I provided by the client.Note: 1. Where required, provide clarification/justification for answers to the questions in the space provided below each question.

An explanation is required for any questions answered as "No' or "N/A".Originator:

David Hartmangruber I Date Print Name and Sign Vogtle Electric Generating Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 3 Vogtle EAL Calculations

~ER~A 1~V'* ~~#rIZbd?Southern Nuclear Design Calculation SCalculation Number: X6CNAI5 Plant: Vootie Electric Generatino Plant Unit: fl 1 0] 2 [ 1 & 2 i Discio)line:

Mechanical Title: NEI 99-01 Rev 6 EAL Calculations

Subject:

Emergency IAction Level Setjpoints Purpose I Objective:

Document Emer eq.nc Action Level Values to support conversion to NEI 99-01 Rev 6 System or Equipment Tag Numbers: NIA Contents Topic Page "Attachments

1. of (Computer Printouts, Technical Papers, Sketches, Pages_____________Correspondence)

Purpose I A -SNC EP Concurrence 1 criteria 1 8- Reserved 0)Conclusion 3 C -References from X6CNA14 31 Design Inputs ...6 D0- TEDE and Thyroid CDE Calculations 18 Assumptions 18 E E- Shielding Calculations 29---References 22F -Evaluation of 52 psig Pressure on Mechanical 1 P.en.rtrations Method of Solutions 26 G -References from X6CNA16 Body of Calculation 31 H -VEGP 1, EAL Thresholds 7!- VEGP 2 EAL Thresholds

• 7i___________

___J -Reviewer Alternate Calculation 11"-'K -SNC024-CALC-005 Vogtle EALs RA1 thresholds 1 7______ to address NEI 99-01 rev 6 ______......_____L

-SNCQ24-CALC-004 VEGP EAL for IC E-HUI 16-'--Total # of Pages including cover sheet & 22_8 Attachments:

Nuclear Quality Level 0DSafety-Related E] Safety Significant 0l Non- Safety -Significant' erinRecord

__________________________

vmo origlnator' Revl~wer Approval I Approval2 No. IDescriptioun

_______ _______ _______" sudo~w. Wu C.T, Martin A.T Vieira DL.L Lamnbert____ ___________

Sgonfie Sign. on file Sign. on filegnonfl 2 :Complete Revision ' :.. \L. -.",, ___ r in Hawkinson Jay Maisler Keith Drudy Notes: 1. Additional work~ and changes to this calculation are required.

The calculation is "APPROVED PENDING" NRC review and acceptance of the Emergency Plan (EP) submittal.

2. The RTYPE for this document is CV805,015.

NMP-ES-039-F01 V4,0 NMP-ES-039-001 PURPOSE The purpose of this calculation is to provide values/data/curves and bases for use in development of the Vogtle Electric Generating Plant Emergency Action Levels (EAL) using NEI 99-01 Revision 6 guidelines.

This combined calculation document includes all unique calculations required to support emergency action level threshold as well as references to calculations used to create thresholds, but serve purposes beyond emergency action levels.The contents of this calculation are primarily an amalgamation of the calculations which supported the previous emergency action level scheme. The work performed in calculations X6CNA14 and X6CNA16 is directly transposed into this document and edited to account for the differences between NEI 99-01 Rev 4 and 6.Several emergency action level thresholds have been eliminated due to the new scheme. The thresholds previously calculated supporting ICs RU2, RA2, RA3, and SU4 have not been carried into this document.

Attachments K and L of this document contain calculations supporting new emergency action level thresholds and represent the only portion of this document which is not directly transposed from X6CNA14 and X6CNA16. The transposed material in this calculation has been further altered to reflect the new language and organization of NEI 99-01 Rev 6. These changes in language and organization are administrative in nature and have no impact on the calculation output.Criteria: The calculation performed will support the development of guidelines for NEI 99-01 EALs RU1, RA1, RSI, RO1, CAl, CS1, OGI, E-HU1, Fuel Clad Barrier Loss 3.A, RCS Barrier Loss 3.A, Containment Barrier Potential Loss 3.A and Containment Barrier Potential Loss 4.8.1. Declaration of an emergency, when such a declaration is not required, involves risk to the public as does the failure to make such a declaration, should one be warranted.

Therefore, this calculation shall develop a "best estimate" value for the dose rates or curie concentrations sensed at the monitors chosen for the Emergency Action Level (EAL) set points. When judgments are necessary, these judgments shall be as close to anticipated conditions as possible.2. If a particular monitor is to be used for an EAL, then the dose rate or curie concentration set point developed for that specific monitor shall be within the range of the monitor, or the monitor shall not be cited as applicable for the EAL.3. In accordance with the guidance of Regulatory Guide 1 .97 Revision 2, post-accident radiation monitors must read within a factor of 2 of actual radiation conditions.

Therefore, changes in the set points of this revision that are within a factor of 2 of the previous revision's set point for the same EAL do not invalidate the previous set point. It is up to the ultimate user of these calculations to determine if change to the EAL set point guidance document(s) is warranted.

4. Methods and Assumptions shall comply with the guidance of NEI 99-01 Revision 6.

Southern Nuclear Operating Company SOUTHERNE,.

Pln:VG Title: NEI 99-01 Rev 6 EAL Calculations XIN1 I COMPANY Unit: 1&2 SHEET 2 Note: NEI 99-01 Rev 6 states that the "A" Recognition Category designation may be changed to "R" provided the change is carried through for all of the associated IC identifiers.

As such, the Vogtle Electric Generating Plant Emergency Action Levels use the Recognition Category designation of "R" for the Abnormal Rad Levels / Radiological Effluent Recognition Category.

Southern Nuclear Operating Company SIm Plant: VEGP I X6CNA1 5 COMplant Unit: 1&2 Title: NEI 99-01 Rev 6 EAL CalculationsSHE3

==

Conclusions:==

The results of the calculations of the values / data required for the develoPment of the Vogtle Electric Generating Plant EALs are provided in the Body of the Calculations.

Initiating Condition RUl The liquid and gaseous release paths are continuous (per Tables 2-4 and 3-4 of VEGP ODCM), with release permits and alarm setpoints generated weekly.Thus, default setpoints are deleted from this version of the calculation.

Initiating Condition RA1 Greater than any of the following monitor readings, for 15 minutes or longer, serves as the threshold for EALI.Radiation Vent Path Reading Monitor RE-12444E Plant Vent -High Range 0.50 pCi/cc RE-I12839E Turbine Building Vent (SJAE) High 21 pCi/cc____ ___ ___ Range _ _ _ _ _ _Initiating Condition RS1 Greater than any of the following monitor readings, for 15 minutes or threshold for EALl.longer, serves as the Radiation Vent Path Reading Monitor RE-12444E Plant Vent -High Range 5 pCi/cc RE-12839E Turbine Building Vent (SJAE) -High Range 210 pCi/cc Southern Nuclear Operating Company SOT AN4 Plant: VEGP TteNE990Re6EACacliosX6CNA15 SOUTERA Unit: 1&2 Tite NE 90 e A acltosSHEET 4 Initiating Condition RGI Greater than any of the following monitor readings, for 15 minutes or longer, serves as the threshold for EALI.Radiation Vent Path Reading Monitor RE-12444E Plant Vent 50 jiCi/cc RE-12839E Turbine Building Vent (SJAE) -High 2100 Range Initiating Condition CA1 The following indication serves as a threshold for EAL 1 RPV water level less than 185'-10" Initiating Condition CSI The following indication serves as a threshold for EAL 1 .b.RPV water level less than 1 85'-4" The following indication serves as a threshold for EAL 2.b RPV level less than 181'-1 0" Greater than or equal to the following monitor reading serves as a threshold for EAL 3.b RE-005 or Containment Operating Deck High Range > 40 REM/hr RE-006 Initiating Condition CGI The following indication serves as a threshold for EAL 1l.b RPV level less than 181 '-1 0" Greater than or equal to the following monitor reading serves as a threshold for EAL 2.b RE-005 or Containment Operating Deck High Range > 40 REM/hr RE-006 Southern Nuclear Operating Company souTHl Pat:VG il:NE 90 e EAIacuain X6NA5 COMPANY Unit: 1&2 Til:HEEE-1Te ELCacla5n Initiating Condition E-HUI Greater than any of the following on-contact radiation readings (gamma + neutron) serve as the thresholds for EAL 1.LocationEA

______________________(m remlhr)HI-TRAC 125 Side -Mid -height 950 Top J 200 HI-STORM 100 Side -60 inches below mid- 170 Side -Mid -height 180 Side -60 inches above mid- 110 Top -Center of lid 50 Top -Radially centered 60 Inlet duct 360 Outlet duct 130 Fuel Clad Barrier Loss Threshold 3.A Greater than the following monitor readings serve as Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 > 2.6E+5 mR/hr RCS Barrier Loss Threshold 3.A Greater than the following monitor readings serve as Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 >= 8.7E+2 mR/hr Containment Barrier Potential Loss Threshold 3.A Greater than the following monitor readings serve as Potential Loss Threshold 3.A Containment radiation monitor RE-005 OR -006 >= 1 .3E+7 mR/hr.

Southern Nuclear Operating Company SOUHEA, Plant: VEGP Titl:NI9-1Rv6ELCluain X6CNA15 ISUHUnit:

1&2 TteNE990Re6EACacliosSHEET 6 DESIGN INPUTS Radiation Monitor System Parameters

1. Vogtle Radiation Monitoring System Operating Ranges Liquid Effluent Monitor Release Path Operating Range Page RE-0018 Liquid Radwaste Effluent Line 1.0E-06 to 1,0E-01 jiCi/cc 10 RE-0021 SIG Blowdown Effluent Line 4.0E-07 to 4.0E-02 pCi/cc 11 RE-0848 Turbine Building Drain Effluent Line 4.0E-07 to 4,0E-02 pCi/cc 12 Gaseous Effluent Monitor Release Path ARE-0014 Waste Gas Process Effluent Line 1.0E-01 to 1.0E+04 6 RE-I124420 Plant Vent 5.0E-07 to 5.0E-02 pCi/cc 3 RE-I124440 Plant Vent -Normal Range I1.0E-07 to 3.4E-03 pCi/cc 3 & 4 RE-I12444D Plant Vent -Mid Range 3.4E-03 to 4.0E-01 jiCi/cc 3 & 4 RE-I12444E Plant Vent -High Range 4.0E-0I to 5.8E+04 p, Ci/cc 3 & 4 RE-I128390 Steam Jet Air Ejector -Normal 1 .0E-07 to 3.4E-03 liCi/cc 1 RE-I12839D Steam Jet Air Ejector -Mid Range 3.4E-03 to 4.0E+01 #iCi/cc I RE-I12839E Steam Jet Air Ejector -High Range 4.0E+01 to 5.8E+04 pCi/cc I Area Radiation Monitors Monitor Release Path RE-0002 Containment Operating Deck I1.0E-01 to I1.2E+03 mREM/hr 4 & 5 RE-0003 Low Range RE-0004 Containment Access Hatch Area I1.0E-01 to I1.2E+03 mREM/hr 13 RE-0005 Containment Operating Deck 1 .0E-01 to 1 .2E+08 REM/hr 5 RE-0006 High Range RE-0008 Fuel Handling Building 1.0E-01 to I.2E+03 mREM/hr 13 RE-0011 Seal Table Room 1.0E-01 to I.2E+03 mREM/hr 13

Reference:

Attachment C, X6AZ01A.

2. Liquid Effluent Monitors' Alarm Setpoints Monitor Release Path Release High Alarm Reference Type* Setpoint (iC i/cc)RE-0018 Liquid Radwaste Effluent Line Batch Procedure 3431 1-C During Release Permit Scin Dependent 5.1 & 5.3 Between Releases 9.99E+20 RE-0021 SIG Blowdown Effluent Line Continuous Procedure Permit 34306-C During Release Section Dependent 5.1 & 5.3 No Activity 1 .00E-05 RE-0848 Turbine Building Drain Effluent Continuous Procedure Line 34310-C During Release Permit Scin____ ___ ____ ___ ____ ___ ____ ___ __ D pen ent 5.1 & 5.3 No Activity 1 .00E-05* Table 2-4, page 2-18, VEGP ODCM Refer to the following VEGP ODCM figures for the listed radiation monitors: Monitor VEGP 00CM Figure VEGP 00CM Page RE-0018 Figures 2-1 & 2-2 Pages 2-14 & 2-15 RE-0021 Figure 2-3 Page 2-16 RE-0848 Figure 2-3 Page 2-16 3. Gaseous Effluent Monitors' Alarm Setpoints Monitor Release Path Release High Alarm

Reference:

Type* Setpoint Procedure (pl~i/cc) 34333-C ARE-0014 Waste Gas Process Effluent Batch Sections Line 10.1 & 10.2 During Release Permit Dependent Between Releases 9.99E+20 RE-I12442C Plant Vent Continuous Permit Sections Dependent 7.1 & 7.3 RE-12444C Plant Vent Continuous Permit Sections Dependent 8.1 & 8.3 RE-I128390 Turbine Building Vent Continuous Sections (Steam Jet Air Ejector -9.1 & 9.3 Normal)______

No Confirmed Primary-to-Secondary 7.84E-04 Leakage Confirmed Primary-to-Secondary Leakage Permit_______________________________________

Dependent______

• Table 3-4, page 3-21, VEGP 00CM Refer to the following VEGP ODCM figures for the listed radiation monitors: Monitor VEGP ODCM Figure VEGP ODCM Page ARE-0014 Figure 3-1 Page 3-15 1IRE-12442C Figure 3-2 Page 3-16 2RE-1 24420 Figure 3-3 Page 3-17 RE-12839C Figure 3-4 Page 3-18 Plant Vent release type confirmed in 27AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager (copy in Attachment 03).The SJAE discharges to the environment via the Turbine Building Vent. As the SJAE must operate to maintain condenser vacuum while at power, it is a continuous release path.Turbine Building Vent release type confirmed in 28AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager (copy in Attachment 04).

Southern Nuclear Operating Company SO~a k Plant: VEGP X6CNA15*o~l Uit 12 Title: NEI 99-01 Rev 6 EAL Calculations SHEET9 Spent Fuel Pool Parameters

4. SFP Elevations Elevation Value Reference SFP Floor 1 79'-01/4" AX4DR023 Fuel Transfer Tube Centerline 186'-9%" AX4DR023, 1X2D48E007, 2X2 D48 E007 Top of spent fuel racks 1 93'-5" AX4DR023 Elevation of SFP Water Normal Level 21 8'-6"~ AX4DR023 Elevation of SFP Water Low Level 21 7'-0" AX4DR023 Containment Dimensions

5. Containment Elevations

& Dimensions Elevation/Dimension Value Reference Operating Deck Elevation 220'-0" 1x2D48E007, 1X2D48E008, 2X2D48E007, & 2X2D48E008 Containment Inside Diameter 140 ft 1X2D01A001

& 2X2D01A001 Operating deck thickness 2'-9" 1X2D48E007, 1X2D48E009, (above Seal Table Room) & 2X2D48E009 Top of inside Containment*

397'-.9" 1X2D01A001

& 2X2D01A001 Fuel Transfer Tube Centerline 186'-9%" AX4DR023, 1X2D48E007, 1 XD48E008, 2XD48E007 Elevation of RPV flange 1 94'-0" AX4DR023* Elevation of liner inner surface at top = spring line elevation 327'-9" + 70'-0" radius 6. Containment volume = 2.95E+06 cu ft

Reference:

Table 1 5A-1, VEGP FSAR Revision 20 (December 2015)7. Containment volume fraction above operating deck = 77.1%Unsprayed net free volume above operating deck = 68,900 cu ft Total net free volume below operating deck = 603,200 cu ft Volume fraction above operating deck = 1 -[(68,900 ft 3+ 603,200 ft 3)/(2,932,000 ft 3)]

Southern Nuclear Operating Company SOUTHERNA Ulnit: VEGPi Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 Volume fraction above operating deck = 0.771

Reference:

Table 6.5.2-1, VEGP FSAR Revision 19 (February 2014)8. Containment liner: 1/4" carbon steel

Reference:

VEGP FSAR sections 1.2.5, 6.2.7.2, & 6.5.3.1 and drawings 1X2D01A001

&2X20D01A001 Reactor Coolant System Parameters

9. Reactor Pressure Vessel & RCS Piping Dimensions Parameter Value Reference RPV Inside Diameter 173" VEGP FSAR Table 5.3.3-1 Hot Leg centerline elevation 187'-0" AX4DR023, IX4DL4A017-1, &(76% RVLIS) 2X4DL4A01 7-1 Cold Leg centerline elevation Hot Leg Nozzle Bottom 185'-91/2A" AX4DR023 Top of Active Fuel 181'-10" AX4DR023 (63% RVLIS)Cold Leg Pipe ID 27%" 1X4DL4A017-1

& 2X4DL4A017-1 Hot Leg Pipe ID 29" 1X4DL4A017-1

& 2X4DL4A017-1 RCS Coolant Parameters Parameter Value Reference Full Power Tavg 588.4 °F Table 2-1, page 2-3, WCAP-1 6736-P VEGP FSAR Table 15.0.3-3 RCS operating pressure 2250 psia Full power coolant mass 2.53E+08 g Page 3 of LTR-CRA-06-1 79 attached to WEC-SNC letter GP-1 8006 and Table 7.8-3 of WCAP-1 6736-P 10.11. Fuel Assembly outside dimensions

= 8.424" x 8.424"

Reference:

1 X6AN09-1 0000-2 & 2X6AN09-1 0000-0 12. Core effective diameter = 132.7 inches x 1 footl12 inches = 11.06 ft

Reference:

Table 5-1, page 5-4, 1/2X6AAI10-00095 Source Terms Southern Nuclear Operating Company SOUHERN A1A Plant: VEGP I X6CNA15 COMPANY Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SET1 13. Reg Guide 1.183 Core Release Fractions Noble Gases (Xe & Kr): 1.0 Halogens (I & Br): 0.4

Reference:

Tables 2 and 5, Reg Guide 1.183 14. Core Fission Product Radionuclide Inventory

& RCS Coolant Equilibrium Activity Isotope Core Inventory RCS Activity Kr-85 I1.04E:+06 8.37 Kr-85m 2.68E:+07 2.04 Kr-87 4.93E:+07 1.28 Kr-88 7.02E:+07 3.68 Xe-131m 7.13E:+05 2.02 Xe-133 2.12E:+08 256 Xe-133m 3.01E:+07 17.6 Xe-1 35 4.65E::+07 8.30 Xe-135m 4.18E:+07 0.56 Xe-138 1.69E:+08 0.74 1-131 1.03E+08 2.91 1-132 1.501E+08 2.96 I-133 2.10E+08 5.56 I-134 2.26E:+08 0.69 1-135 1.95E:+08 2.72 15. RCS Coolant Dose Equivalent 1-131 Activities Isotope 1.0 DE I- 60 iLCilg DE 1-131 131 Coolant Coolant Activity Activity (jltCi/g) (jiCilg)I-131 0.74 44.4 1-132 0.75 45.0 1-133 1.41 84.6 1-134 0.18 10.8 1-135 0.69 41.4

Reference:

Page 2 of LTR-CRA-06-179 RI attached to WEC-SNC letter GP-1 8006 and Table 7.8-1 of WCAP-1 6736-P Note: Per section B.3.4.16, page B3.4.16-2 of VEGP Tech Spec Bases, noble gas activity in the reactor coolant assumes 1 %failed fuel, which closely equals the LCO limit of pCi/gm for gross specific activity.

Reference:

Page 3 of LTR-CRA-06-1 79 RI attached to WEC-SNC letter GP-1 8006 and Table 7.8-2 of WCAP-1 6736-P Note: Iodine concentrations developed using thyroid DCFs from FGR #11.

Southern Nuclear Operating Company SOUTaN 4 Plant: VEGP X6N1 COMPANY Unit: 1&2 Til:NI9-1Rev 6 EAL Calculations SHEET 12 16. Deleted 17. Core Fission Product Inventory

-Fuel Rod Gap Isotope Fuel Rod Isotope Fuel Rod Isotope Fuel Rod Gap Gap Gap Inventory Inventory inventory_ _ _ _ _ (Ci) (Ci) (Ci)I-131 1.03E+07 Kr-85m 2.68 E+06 Xe-131 m 7.13E+04 1-132 1.50E+07 Kr-85 3.12E+05 Xe-133m 3.01E+06 1-133 2.10E+07 Kr-87 4.93E+06 Xe-133 2.12E+07 I-134 2.26E+07 Kr-88 7.02 E+06 Xe-135m 4.18E+06 1-135 1 .95E+07 Xe-1 35 4.65E+06 Xe-I138 1 .69E+07

Reference:

Table 15A-3, VEGP FSAR, Revision 19, February 2014 Update Dose Rate vs. SFP Water Depth 18. VEGP SFP Irradiated Fuel Dose Rate vs. Water Depth above fuel See Assumption

1. 8 for discussion of adjusting this data in Attachment E2 for use with irradiated fuel in Vogtle RPV Depth Dose Rate (ft) (mREM/hr)8 1 .2696E+04 10 6.3753E+02 11.1 1.2712E+02 12 3.1412E+01 14 1.8272E+00 16 1.1519E-01

Reference:

Appendix D, X6CDE.01 Bases:* 193 fuel assemblies in spent fuel storage racks* 100-hours after S/D* Equivalent cylinder diameter = 13.7 feet* Core source term multiplied by 0.72 to account for larger cross sectional area of effective cylindrical source in SFP Release Path Flow Rates 19. SJAE via Turbine Building Vent Flow rate = 900 CFM (4.25E+05 mL/s)Release Elevation

= Ground Level X/Q = 2.55 E-06 sec/in 3

Reference:

Table 3-4, VEGP-ODCM~

20. Plant Stack Vent Southern Nuclear Operating Company SUH , Plant: VEGP I X6CNA15 SnitTH1&2 TiteNE990Rv6EA Clua ion SHEET 13 I COMPANY Unt &Til:EI9-1Rv6ELCluaos U1 Flow rate = 187,000 OEM (8.83E+07 mL/s)U2 Flow rate = 112,500 CEM (5.31 E+07 mL/s)Release Elevation

= Mixed-Mode X/IQ = 4.62E-07 sec/rn 3

Reference:

Table 3-4, VEGP-ODCM Southern Nuclear Operating Company SOUHERM Plant: VEGP X6N1 SOTE~Unit:

t &2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET1 Conversion Factors 21. FGR 12 Effective Dose Equivalent (EDE) Dose Conversion Factors for external exposures Isotope EDE Air EDE Air Immersion Immersion DCF DCF (Svlsec)l (mREMIhr)/

______(Bqlm^3)

Kr-85 1.19E-16 1.59E+03 Kr-85m 7.48E-15 9.96E+04 Kr-87 4.12E-14 5.49E+05 Kr-88 1 .02E-1 3 1 .36E+06 Xe-131m 3.89E-16 5.18E+03 Xe-133 1.56 E-15 2.08 E+04 Xe-133m 1.37E-15 1.82 E+04 Xe-135 1.19E-14 1.59E+05 Xe-135m 2.04E-14 2.72 E+05 Xe-138 5.77E-14 7.69E+05 I-131 1.82E-14 2.42 E+05 I-132 1.12E-13 1.49E+06 1-133 2.94E-14 3.92E+05 I-134 1.30E-13 1.73E+06 1-135 7.98E-14 1.06E+06

Reference:

"Effective" column, Table 111.1, "Dose Coefficients for Air Submersion," in FGR 12, "External Exposure to Radionuclides in Air, Water, and Soil" Geometry:

semi-infinite (i.e., infinite hemispherical) cloud source, air density = 1 .2 kg/in 3.To derive coefficients for other densities, multiply DCFs by (1 .2/p), where p = air density in kg/rn 3 EDE DCF EDEODCE (REM/hr) = (Svlsec) x 100 REM x 3600 sec x 1 Bq x 1.0 Ci x 1(Bq/m^3) 1 Sv 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2.703E-11 Ci 1.0E+06 j, Ci 1 m^3 1 .0E+06 cc EDE DCF (mREM/hr)(1.iCilcc)EDE DCF= (REM/hr)x 103 mREM 1 REM Southern Nuclear Operating Company SOUMANY Uk lnit: VEGPI Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 SoUHer UntI& SHEET 15 22. FGR 11 CEDE Dose Conversion Factors Isotope CEDE Air CEDE Air Thyroid CDE Air Thyroid CDE Air Inhalation Inhalation Inhalation Inhalation DCF DCF DCF DCF (SvlBq) (Sv/Bq) (mREM/jItCi)

Kr-85 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-87 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-88 0.00E+00 0.00E+00 0.O0E+00 0.00E+00 Xe-131 m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-1 33 0.00E+00 0.00E+00 O.00E+00 0.00E+00 Xe-1 33m O.OOE+OO 0.00E+00 0.00E+00 0.00E+O0 Xe-1 35 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-I135m O.OOE+O0 0.00E+00 0.00E+00 0.00E+00 Xe-I138 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-131 8.89E-09 3.29E+01 2.92E-07 1.08E+03 I-132 1.03E-10 3.81E-01 1.74E-09 6.44 E+00 I-133 I.58 E-09 5.85E+00 4.86E-08 1.80E+02 1-134 3.55E-11 1.31E-01 2.88E-10 1.07E+00 1-135 3.32E-10 1.23E+00 8.46E-09 3.13E+01

Reference:

"Table 2.1, Federal Guidance Report 11 CEDE DCF: Column labeled "Effective" Thyroid CDE DCF: Column labeled "Thyroid" Per page 121, FGR-1 1: CEDE DCF (mREM/piCi)

= 3.7x109~ x CEDE DCF (Sv/Bq)

Southern Nuclear Operating Company SOUTHER£ Ulnit: VEGP X6N1 U lnit: VE&2 Title: NEI 99-01 Rev 6 EAL Calculations

'SET1 23. Unit Conversions Conversion Reference secant = 1/cosine Page 2-25, "Marks' Standard Handbook for Mechanical Engineers" 1 atmosphere

= 76 cm Hg = 14.7 psia Page F-303, "CRC Handbook of Chemistry

& Physics" 1 cubic foot = 0.02831 6847 cubic meters Page F-308, "CRC Handbook of Chemistry

& Physics" I day = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Page F-308, "CRC Handbook of Chemistry

& Physics" 1 foot = 12 inches Page F-310, "CRC Handbook of Chemistry

& Physics" 1 foot = 0.3048 meter Page F-31 0, "CRC Handbook of Chemistry

& Physics" 1 foot = 30.48 centimeters (cm) Page F-31 0, "CRC Handbook of~Chemistry

& Physics" 1 gallon = 3.7854118 liters Page F-31 1, "CRC Handbook of Chemistry

& Physics" 1 gram = 0.001 kilogram Page F-312, "CRC Handbook of Chemistry

& Physics" 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> = 3600 seconds Page F-31 3, "CRC Handbook of Chemistry

& Physics" 1 milliliter

= 1 cubic centimeter Page F-31 8, "CRC Handbook of Chemistry

& Physics" 1 pound = 453.59237 g Page F-320, "CRC Handbook of Chemistry

& Physics" 1 pound/cubic foot = 0.016018463 g/cc Page F-321, "CRC Handbook of Chemistry

& Physics" 1 year = 365.25 days Page F-325, "CRC Handbook of Chemistry

& Physics" I Becquerel (Bq) = 2.703x1 0-11 Curie (Ci) Page 22, Lamarsh, "Introduction to Nuclear Engineering" 1 Sievert (Sv) = 100 REM Page 404, Lamarsh, "Introduction to Nuclear Engineering" Temperature Scale Conversions Page 16, Holman, "Heat Transfer'0 F = 1.8 x °C + 32°R = F +460 _____________

Miscellaneous Design Inputs 24. Iodine boiling point = 184 °C = -363 °F

Reference:

Page B-I, "CRC Handbook of Chemistry

& Physics" 25. Density of Refueling Cavity and Spent Fuel Pool Water @ 130 °F = 61.55 Ibm/cu ft

Reference:

See Attachment C2.

Southern Nuclear Operating Company SOUHERNA4t Plant: VEGP II X6CNA15 SOMPNY Uit 12 Title: NEI 99-01 Rev 6 EAL Calculations SET1 26. Deleted 27. Deleted Radiation Monitor System Parameters

28. VEGP Digital Radiation Monitoring System (ORMS) post-accident monitors required sensitivity

& accuracy shall be in accordance with Reg Guide 1 .97 R2.

Reference:

Sections 2.6 & 3.1.23, Specification X6AZ01A, VEGP 1&2 Digital Radiation Monitoring System (DRMS)29. Deleted Southern Nuclear Operating Company IsoutHERN.

Plant: VEGP TteNE990Rv6EACacliosX6CNA1 5 COMpa~nY Unit: 1&2 TteNI990Re6EACacliosSHEET 18 ASSUMPTIONS

1. TEDE and Thyroid dose calculations based on one hour of inhalation Justification:

Page 33, NEI 99-01 Revision 6 2. Breathing rate = 3.47x10-A m 3/sec Justification:

VEGP-FSAR Table 15.6.3-8 and Section 4.1.3 of Reg Guide 1.183 3. The following partition factors are assumed to determine release activities Radionuclide PF Justification Noble Gases 1.0 NUREG-0017 (PWR-GALE), Section 1.5.1.8 Iodines Steam Generator 0.01 VEGP FSAR Table 15.2.6-1 (sheet 2 of 2)NUREG-0017 (PWR-GALE), Section 1.5.1.8 Air Ejector 1 .0E-04 FNP FSAR Table 12.2-1 Reasonable to assume Iodines behavior in VEGP Main Condenser same as in FNP Main Condenser.

Liquid leakage to Auxiliary 0.01 VEGP FSAR Table 12.2.2-1 (sheet 1 of 2)Building (Round up from 0.0075)4. No noble gases are retained in the S/G: i.e., all noble gases leaked to the secondary system are continuously released with steam through the SJAE.Justification:

VEGP FSAR sections 15.2.6.3.1.3 (page 15.2-11), 15.3.3.3.1.3 (page 15.3-9), & 15.4.8.3.1.3 (page 15.4-31)5. Core inventory release fractions Noble gases: 1.00 Iodines: 0.40 Justification:

Table 2, "PWR Core Inventory Fraction Released into Containment," page 1.183-14, Regulatory Guide 1.183 6. Specific volume of steam release = 26.804 cu ft/Ibm (ASME Steam Tables)Justification:

Specific volume of saturated steam at atmospheric pressure (Attachment C2)7. Specific volume of main steam = 0.4461 cu ft/Ibm (ASME Steam Tables)

Southern Nuclear Operating CornpanyPlant: VEGP ITteNE990Re6EACacliosI X6CNA1 5 SOUTPANY Unit: 1&2 Til:NI9-1Rv6 LCluaion SHEET 19 Justification:

Full power main steam pressure = 1000 psia (VEGP FSAR Table 10.3.2-1, sheet I)8. The VEGP SEP dose rate at water surface vs. SFP water depth assessment in Appendix D of calculation X6CDE01 is acceptable for estimating the water surface dose rate vs. depth for fuel in the reactor vessel at Vogtle.Justification:

• The source is assumed to consist of an offloaded core (193 fuel assemblies) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.*The VEGP SFP analysis assumed each fuel assembly's source term was spread across a storage cell cross sectional area of 110.03 sq in. This reduced the source term (MeV/sec-cc) by a factor of 0.72.*Adjusting the source term back to a geometry matching the closer spacing in the reactor increases the source term by 1/0.72, or 1.39. Since dose rate is proportional to source strength, the VEGP SFP dose rates are multiplied by 1.39.* The effective diameter of the VEGP SFP cylindrical source is 13.7', compared to the Vogtle reactor core effective diameter of 11.06 ft (Design Input #12). This effect is evaluated in Attachment E2 of this calculation.

9. The reflected dose rate at the operating deck area radiation monitors will be calculated using the methods of Davisson's "Gamma Ray Dose Albedos" (copy in Attachment C1).The calculation will be based on an iron reflector at the top of containment, with a diameter equal to the reactor pressure vessel inside diameter (RPV ID), and a distance (r feet) from the reflector to the radiation monitor equal to the hypotenuse of the triangle formed by the difference in elevations of the reflector (177.75 feet = 397'-9" -220"-0";

Design Input #5)and the monitor and one-half of the Containment ID (70 ft = 1/2Ax 140 feet; Design Input #5).Justification:

The iron reflector is selected because the Vogtle containment has a carbon steel liner. The reflected dose rate is proportional to the area of the reflector.

Assuming the reactor vessel functions as a collimator with reduced ROS inventory will reduce the reflected area. This in turn reduces the dose rate at the radiation monitor and, therefore, the EAL threshold for reduced RCS inventory.

Simplified diagram, based on 1X2D48E008 (2X2D48E008 dimensions same)10. The RCS concentration of 1-131 (lo) at a known fuel clad defect level (Do) may be used to determine the defect level (Di) at another RCS 1-131 concentration (1i) with a simple ratio: Dlh= Do/Io Di = (li/lo) x Do Justification:

Sheet 4-2 of WCAP-8253 (excerpt in Attachment G1 of this calc)11. The temperature of the air in containment is -235 0 F Justification:

VEGP FSAR Figures 6.2.1-4 & 6.2.1-5 12. The pressure of the air in containment is -,30 psig = -45 psia Justification:

VEGP FSAR Figures 6.2.1-1 & 6.2.1-2

13. The density of the air in containment (pair) at 45 psia and 235 0 F is 2.811 kg/rn 3 Justification:

Air is assumed to behave as an ideal gas. The ideal gas law (page 3-44,"Marks' Standard Handbook for Mechanical Engineers")

is (P1 X V1)IT1 = (P2 X V2)/T2 '4 v2 = Vl x (PI/P2) x (T2/'rl)where P = pressure (psia)T = temperature

(°R)v = specific volume (ft 3/lbm)subscript 1 = reference condition subscript 2 = containment conditions, 45 psia & 235 °F (695 °R)The reference conditions for the air density of FGR-12 Table II1.1, 1.2 kg/rn 3 , correspond to 20 °C (68 0 F =528 R) and 1 atmosphere (14.7 psia)per the dry air density table on page F-11 of "CRC Handbook of Chemistry

&Physics" v2 = (1/1.2 kg/in 3) x (14.7 psia/45 psia)x (695 0 R/528 0 R)v2 = (0.8333 m 3/kg) x (0.3267) x (1.3068)v2 = 0.3583 m 3/kg pair2 = 1/v2 = 1/(0.3583 m 3/kg)pair2 = 2.791 kglm 3 DENSiTY 01' DRY AIR A2 -= TmulaarwaI

~ii uz m Pmssm. H xow Maou smr thism b1, w, pi m, St (=~i 76l.? c m b H tel PhaqOn o.pulbi Pains Him Cern a 72.0 78.0 [~j 78.0 EOIj ~I1 12 14 16l 12 18 13 0.01~1 o.oo1141 171 o.oolnel 181 loi 144 141 0.001101 I11 101 2.001181i 1uN 161 172 ill 0.00118 184 1.t6 p,00t121 210 S20 1M 186 181 17"7 168 0.001113 143 146 142 186 0.601331 221 201 211 214 0.601.214 201 201 101 141 341 141 381 m 181 127 175 011*166 161 157 158* .061324 1321 1311'3.001W 101 141 141 0.001141 Ii'171 141 0.001264'U 281 281 244 0.001241 231-I.221'U I3.00I22~r 214 211'U 004 0.001204 134 161 141 184 17 0.1 0.8 0.4 0.8 0.8 0.7 0.8 0.3 16 0.1 0.4 0.5 0.6 0.7 0.8 0.3 15 em 0.1 0.3 0.4 0.6 0.6 0.7 0.8 0.3 2 8 8 10 11 18 14t 2 8 8t 7 8 10 1,2 14 15t 1 I 6 7 9 12t 11 0.0 110410.6011110.001184iS 110118010.00116810.001180 I-IT368 Fz=20C-0.0012 g/mil- 1.2 kglm 3 Southern Nuclear Operating Company souTiERNA U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 REFERENCES

1. VEGP 1 &2 FSAR, Revision 20, December 2015 Update 2. VEGP-1 Technical Specifications, Amendment 177, December 2015 Update 3. VEGP-2 Technical Specifications, Amendment 158, December 2015 Update 4. X6CNA14, V7.0, "[VEGP] NEI 99-01 EAL Calculations" 5. X6CNA16, V1 .0, "[VEGP] NEI 99-01 Revision 4 Fission Product Barrier EAL Setpoints" 6. SNC024-CALC-005, Rev. 0, "Vogtle EALs RAl Threshold to Address NEI 99-01 Revision 6" 7. SNC024-CALC-004, Rev. 0, "VEGP Determination of Emergency Action Level for Initiating Condition E-HUt" Methods 8. NEI 99-01 Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012, (http://pbadupws.nrc.ciov/docs/MLi232/M LI2326A805.pdf)

9. Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (http ://pbadupws.nrc..qov/docslMLOO371MLOO3716792.pdf)

10. Regulatory Guide 1.195, Revision 0, "Methods and AssumPtions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003 (http://pbadupws.nrc..qov/docs/M L0314/ML031490640.pdf)

11. NUREG-001 7, Revision 1, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1985 Reg Guide 1.197 12. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 (http://pbadupws.nrc.qov/docs/MLO6071MLO60750525.pdf)

System Specifications Containment

& Containment Penetration Specifications

13. DC-1818, Revision 6, "Electrical Penetration System" 14. DC-2101, Revision 5, "Containment Building" 15. X3AR01-E3, Revision 6, "[Specification for] Reactor Containment Electrical Penetrations for Vogtle Electric Generating Plant Units 1&2" 16. X4AQ10, Revision 7, "Specification for Pipe Penetrations for [Vogtle Electric Generating Plant Units 1&2]" Radiation Monitoringq System Specifications

17. X6AZO1A, VI.0, "Specification for Digital Radiation Monitoring System (DRMS) for Alvin W. Vogtle Electric Generating Plant -Units 1, 2," 31 August 2007 Southern Nuclear Operating Company SOUTHERNA Plant: VEGP Til:HEE9-TRv 2A3CluatosX6N1 CPNY Unit: 1&2 TileHEEE-0TRv6 A2Clcltin Emergency Plan & Procedures

18. VEGP-EPLAN, Revision 65, "Vogtle 1&2 Emergency Plan" (available at SNC Emergency Planning website, http:l/n uclear.southernco.com/support-services/emerqency-planning/default.html)

19. NMP-EP-1 10-GL03, V5.2, "VEGP EALs -l~s, Threshold Values, and Basis", November 2014.Procedures Dose Calculations

20. VEGP-ODCM, V30.0, "Vogtle Offsite Dose Calculation Manual" (available at SNC Regulatory Affairs Vogtle Licensing Documents website, http://n uclear.southernco.com/regulatory-affairs/Voqtle-Licensingq-Documents.

html)Radiation Monitoringq System Setpoints 21. 34306-C, VI19.1, "Operation of DRMS Steam Generator Blowdown Liquid Process Monitor RE-002 1" 22. 34310-C3, V21.0, "Operation of DRMS Turbine Building Drain Liquid Effluent Monitor RE-0848" 23. 34311-C, V28.0, "Operation of DRMS Liquid Release Monitors 1(2)RE-0018" 24. 34333-C, V12.0, "Gaseous Effluent Monitor Setup For Releases" Refueling Containment Integrity 25. 14210-1, V18.4, "Containment Building Penetrations Verification

-Refueling" Drawings Shared Drawingqs 26. AX4DR023, V4.0, "Volumes & Water Elevations in the Primary System" Unit I Drawings 27. 1X2D01A001, Revision 5, "Containment Concrete Forming General Arrangement" 28. 1X2D48E007, Revision 4, "Containment Internals General Arrangement Section Looking North" 29. 1X2D48E005, Revision 5, "Containment Internal General Arrangement Plan at EL 220-0" 30. 1X2D48E008, Revision 4, "Containment Internals General Arrangement Section Looking West" 31. 1X2D48E009, Revision 2, "Containment Internals Slab Thickness Plan at EL 220-0" 32. 1X4DL4A013, Revision 7, "Containment Building Unit I Containment Wall Pipe Penetration Design List" 33. 1X4DL4A014, Revision 9, "Containment Building Unit I Containment Wall Pipe Penetration Design List"

34. 1X4DL4A017-1, V25.0, "Containment Bldg. Piping Area 4A, B, C & 0 Level B -Plan &Sections Reactor Coolant Loops" 35. 1X5DS4B002, Revision 9, "Instrument Location Drawing Area 4B -Levels 1 & 2 Containment Building Plan -EL 220-0 to EL 238-0" 36. 1X5DS4D002, Revision 8, "Instrument Location Drawing Area 4D -Level 1 Containment Building Plan -EL 220-0 to EL 238-0" 37. 1X6AN09-1 0000, Revision 2, "Top and Bottom 17 X 17 High Burnup Fuel Assembly Outline& Reprocessing" Unit 2 Drawingqs 38. 2X2D01A001, Revision 5, "Containment Concrete Forming General Arrangement" 39. 2X2D48E005, Revision 6, "Containment Internal General Arrangement Plan at EL 220-0" 40. 2X2D48E007, Revision 0, "Containment Internals General Arrangement Section Looking North" 41. 2X2D48E008, Revision 1, "Containment Internals General Arrangement Section Looking West" 42. 2X2D48E009, Revision 0, "Containment Internals Slab Thickness Plan at EL 220-0" 43. 2X4DL4A01 3, Revision 5, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" 44. 2X4DL4A014, Revision 4, "Containment Building Unit 2 Containment Wall Pipe Penetration Design List" 45. 2X4DL4A017-1, V18.0, "Containment Bldg. Piping Area 4A, B, C & D Level B -Plan &Sections Reactor Coolant Loops" 46. 2X5DS4B002, Revision 3, "Instrument Location Drawing -Containment Building -Area 4B-Level I Plan -EL 220-0 to EL 238-0" 47. 2X5DS4C002, Revision 3, "Containment Building -Instrument Location Drawing Area 4C-Level 1 Plan -EL 220-0 to EL 238-0" 48. Deleted 49. 2X6AN09-1 0000, Revision 2, "Top and Bottom 17 X 17 High Burnup Fuel Assembly Outline& Reprocessing" Calculations

50. X6CDE.01, V5, "[VEGP] Spent Fuel Pool Shielding" 51. X6CNA11, V10.0, "Severe Accident Management Guideline (SAMG) Calculations" Source Term 52. WCAP-1 6736-P, Revision 1, "Vogtle Electric Generating Plant Measurement Uncertainty Recapture Power Uprate Program Engineering Report," May 2007 53. GP-1 8006, "Vogtle Electric Generating Plant Units 1 and 2 -Revised Source Terms for Measurement Uncertainty Recapture Power Uprate Program," 21 September 2006 Southern Nuclear Operating Company souTH=l,,'Pat V IG HEETNE 201Rv6EA acuainsXCA5 Um lnit: VEGP SHCNA25 I COMPANY' Ui:12 Tte E 901Rv6ELCluain

54. VEGP 1&2 Technical Specifications Bases, Revision 36, 29 December 2015 Dose Conversion Factors 55. Federal Guidance Report #11 (EPA 520-1-88-020), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," September 1988 (http://nepis.epa..qov/Simple.

html)56. Federal Guidance Report #12 (EPA-402-R-93-081), "External Exposure to Radionuclides in Air, Water, and Soil," September 1993 (http://nepis.epa.,qov/Simple.html)

GRODEC Computer Code 57. F-86-03, Revision 3, "Verification of the GRODEC Computer Program" Engineering References

58. ANSI/ANS-6.4.3-1991, "American National Standard for Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials," 26 August 1991 59. Davisson, "Gamma Ray Dose Albedos," pages 5-27 thru 5-38 in ANS/SD-76/14, A Handbook of Radiation Shielding Data, edited by J. C. Courtney, July 1976, (copy in Attachment C1)60. Holman, "Heat Transfer," fourth edition, 1976 61. Lamarsh, "Introduction to Nuclear Engineering," second edition, 1983 62. Singer, "Strength of Materials," second edition, 1962 63. CRTD-VOL 58, "ASME International Steam Tables for Industrial Use," second edition, September 2008 64. "CRC Handbook of Chemistry

& Physics," 5 7 th edition 65. "Marks' Standard Handbook for Mechanical Engineers," 8th edition 66. Geick & Geick, "Engineering Formulas," seventh edition 67. Murphy & Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," pages 401 -430, CONF-740807, "Proceedings of the Thirteenth AEC Air Cleaning Conference," Ma~rch 1975 (http:llwww.osti.qov/scitech/servletslpurl141 79572)68. Jaeger (editor-in-chief), "Engineering Compendium on Radiation Shielding:

Volume I: Shielding Fundamentals and Methods," 1968 69. Shultis & Faw, "Fundamentals of Nuclear Science and Engineering," 2002 Southern Nuclear Operating Cornpany SOUTH NY' Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SHEET 26 METHOD OF SOLUTIONS NEI 99-01 Revision 6 Methods conform to the guidance of NEI 99-01 Revision 6. Detailed descriptions of the methods are included in the individual EAL threshold calculations in the Analysis section of this calculation.

Use of Regulatory Guide 1.183, Alternate Source Term Method The NEI 99-01 Revision 6 Recognition Category A (Abnormal Rad Levels/Radiological Effluent)Initiating Conditions (l~s) for declaring an Alert, a Site Area Emergency and a General Emergency (Emergency Action Levels RS1 and RG1, respectively) are expressed in terms of Total Effective Dose Equivalent (TEDE) and Thyroid Committed Dose Equivalent (ODE).Regulatory Guide 1 .195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," is the current license basis for performing dose calculations for Vogtle. However, it expresses doses in terms of Whole Body and Thyroid.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," does express doses in terms of TEDE and CDE, but it is not the current licensing basis for performing dose calculations for Vogtle. However, per section 1.1.4 on page 1.183-6, "This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with emergency dose projections, protective measures, and severe accident management guides." Per section 4.1.1 of RG 1.183, TEDE is defined as the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure.Per section 4.1.2 of RG 1.183, Table 2.1 of Federal Guidance Report 11 provides tables of conversion factors acceptable to the NRC staff. The dose conversion factors (DCFs) factors in the column headed "effective" yield doses corresponding to the CEDE.Per sections 4.1.4 and 4.1.5 of Reg Guide 1.183, the DDE should be calculated assuming submergence in a semi-infinite cloud for the most limiting person at the EAB. The effective dose equivalent (EDE) from external exposure is nominally equivalent to the DDE, thus EDE may be used in lieu of DDE in determining the external dose contribution to the TEDE. Table II].1 of Federal Guidance Report 12 provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.TEDE & Thyroid CDE Calculations The basic process for calculating an offsite dose consists of first determining the concentrations of radionuclides in the release stream, be it air, steam, or water. The release stream concentration is determined by dividing the release rates of the radionuclides of interest, expressed as microcuries per unit time, by the release fluid's volumetric flow rate, expressed as cubic centimeters (cc) per unit time:

Southern Nuclear Operating Cornpany SOUTHaEm 41 Plant: VEGP Title: N M EI 99-01 Rev 6 EAL Calculations X6CNAI 5 COMPANY Unit: 1&2 SHEET 27 liCi/cc =

time]/[cc/unit time]As we are back-calculating release concentrations based on pre-established dose limits (100 mREM TEDE and 500 mREM Thyroid CDE), the upstream modeling of the specific release paths is not necessary.

The gaseous effluent noble gas radiation monitors care not a whit how those radionuclides arrive at them.Step 1: Identify the radionuclides of interest.

Select the same radionuclides used to calculate doses for the design basis accidents in FSAR chapter 15: the fission product noble gases and iodines. The other fission products and activated corrosion products are particulates and will not contribute significantly to the offsite dose.Step 2: Determine the ROS coolant radionuclide activity for each radionuclide (Xrcs-i pCi/g). This is assumed to be the sum of core fission product inventory released during a LOCA divided by RCS coolant mass (Mrcs g) and the equilibrium RCS coolant activity (Xeq-i) for that rad ion uclide.Xrcs-i = Xeq-i + (1 .0E+06 Ci) x [Core Inventory (Ci)] x [Release Fraction]/(Mrcs g)For no fuel damage, the release fraction is 0 and the RCS activity is the equilibrium RCS coolant activity.

If fuel damage is assumed (release fraction > 0), the quotient of core inventory and RCS coolant mass will be orders of magnitude greater than the contribution from the coolant equilibrium activity.Step 3: Convert coolant activity (Xrcs-i to release stream activity (Xris-i jiCi/cc).

This conversion is accomplished by multiplying the ROS coolant activity by a dimensionless partition factor (PFi) and an arbitrarily selected density, prls g/cc: Xrls-i

= (Xrcs-i x PFi x (pris g/cc)The partition factor will depend on the radionuclide and the release path. The partition factors used in this calculation are discussed in Assumption

1. 3 of this calculation.

Arbitrarily set pris = 1.0 g/cc to make the math easy. The justification for this will be provided in Step 9.Step 4: Determine radionuclide concentration at Exclusion Area Boundary (XEAB-I liCi/cc).

This is done using standard dose assessment methods. The release concentration is multiplied by the release volumetric flow rate (Qris m 3/sec) and the diffusion coefficient

[(X/Q) m 3/sec]: XEAB-i

= Xris-i x [Qris (m 3/sec)] x [(X/Q) (m 3/sec)]Step 5: Calculate the TEDE for each radionuclide for one hour exposure time at EAB. This is done using the appropriate FGR-11 and FGR-12 dose conversion factors (DCFs), as discussed in the previous subsection.

TEDE 1 (mREM) = External Exposure + Internal Exposure TEDEi (mREM) = XEAB-I (pCi/cc) x texp (hours) x DDEDcF-i

+XEAB-i (j, Ci/cc) X texp (hours) x BR (cc/hr) x CEDEDCF-I TEDEi (mREM) = XEAB x texp (hours) x TEDEDcF-i

[(mREM/hr)/(pCi/cc)]

where TEDEDcF-i

[(mREM/hr)/(p.Ci/cc)]

= DDEDcF-i [(mREM/hr)l/QiCi/cc)]

+BR (cc/hr) x CEDEDcF-I (mREM/pCi)

BR (cc/hr) = breathing rate Step 6: Add the individual TEDEs to obtain the TEDE for the release (TEDEris):

TEDEris = [TEDEI]TEDEris = [Xris-i x (X/Q) x Qris X texp X TEDEDcF-i]

TEDEris = [(X/Q) x Qris X texp ] X [Xrls-i X TEDEDcF-i]

Step 7: Calculate Thyroid ODE for each Iodine isotope for one hour exposure time at EAB. This is done using the appropriate FGR-11I dose conversion factors (DCFs), as discussed in the previous subsection.

CDETHY-i (mREM) = XEAB-i (pCi/cc) x texp (hours) x BR (cc/hr) x CDETHY-DCF-i (mREM/4LCi)

Step 8: Add the individual Thyroid ODEs to obtain the Thyroid ODE for the release (CDEris): CDEris = [CDETHY-i]

CDEris = [Xrls-i X (X/Q) x Qris X BR X CDETHY-DCF-i]

CDEris = [(X/Q) x Qris x texp X texp] X [Xrls-i X TEDEDcF-i]

Step 9: Determine the 100 mREM TEDE threshold release concentrations for each noble gas (Xi oo-i This is done by multiplying each noble gas' release concentration (Xris-i #C~i/cc)determined in Step 3 by the quotient of 100 mREM and the sum of the TEDEs for all of the radionuclides considered (TEDErls mREM). Only noble gas concentrations are adjusted because the gaseous effluent monitors are noble gas detectors.(Xia0-i p.Oi/cc)/(Xris-i pCi/cc) = (100 mREM)/(TEDEfls mREM)Xioo-i (liCilcc)

= (Xrls-i jiC i/cc) x (100 mREM)I(TEDEris mREM)The following demonstrates that the arbitrarily assumed release stream density has no effect on the final result.Xris-i x (100 mREM)XlOO-1 =[(X/Q) x Qris x texp ] X Z [Xrls-i X TEDEDcF-i]

= Xrcs-i X (1.0) X pris X (100)Xrcs-i X prls X (100)Xpr001x (X/Q Qris x texp X T- [Xrcs-i X PFi X TEDEDcF-i]

Southern Nuclear Operating Company SOUTHERZ U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CN15~COMPANY Pln:VGI Xi 0-iXrcs-i X pr-le X (100)paIs X (XIQ) x Qris x [Xrcs-i X PFi X texp X TED EDcF-i]= Xrcs-i X (1 00)(X/Q) x Qris x texp X [Xrcs-i X PFi x TEDEDcF-I]

The assumed release stream density has no effect on the final result: it cancels out. Thanks to the power of Excel, it is easier to calculate a postulated dose rate and adjust release concentrations than to set up the above equations.

Now to perform a dimensional check:= Xris-i X (100 mREM)1, [Xris-i X (X/Q) x Qris x PFi x texp x TEDEDcF-i]

"7 (jiCi/cc) xmREM-(pCi/cc) x (sec/in 3) x (m 3/sec) x (hour) x [(mREM/hour)/(pCi/cc)](= ./,,G,,G}.

v x (fna/seG) x x [(mREM/h~eu-)I(iCi/cc)]= 1/(pC i/cc)]J4Ci/cc = Step 10: Determine the 500 mREM Thyroid CDE threshold release concentrations for each noble gas (X500T-i ptCi/cc).

This is done using the same method as in Step 9. Again, the arbitrarily assumed release stream density cancels out and has no effect on the final result.Several general trends can be inferred from the equation derived in Step 9 above. Holding other factors constant:* Increasing the diffusion coefficient (XIQ m 3/sec) will reduce the 100 mnREM release concentration.

• Increasing the release flow rate (Q CFM) will reduce the 100 mREM release concentration.

• Increasing the exposure time (t hours) will reduce the 100 mnREM release concentration.

• Increasing the total release [Xrcs-i x PFi x pris X TEDEDcF-I])

will reduce the 100 mREM release concentration.

When assessing TEDE and Thyroid CDE doses against NEI 99-01 Rev. 6, initiating conditions RA1, RS1 and RGI, there are two release paths that will be evaluated:

One release path via the Plant Stack Vent and one release path via the condenser steam jet air ejector (SJAE). The pathways and their major assumptions are summarized below.

Radiation Monitor Release Path Core Damage Partition Factors RE-I12442C Plant Vent Stack Yes Noble Gases: 1.0 RE-I12444C/D/E Iodines: 0.01 RE-12839 SJAE Two Cases: With and Noble Gases: 1.0 Without Core Damage Iodines: I1.0E-06[S/G PF (0.01) x SJAE PF (1 .0E-04)]The release paths via the S/G SRV or PORV and the Turbine Driven AFWP Turbine Exhaust are not evaluated because the Main Steam Line radiation monitors (RE-i13119 thru RE-I13122) were deleted per the 30 September 2014 license amendment (ADAMS # ML14170A91 1).Effects of Plant Configuration on Dose Conversion Factors The FGR 12 Table II1.1 air immersion DCFs are based on immersion in an infinite hemispherical cloud in which the radioisotopes are dispersed.

Per page 415 of the 1 3 th AEC Air Cleaning Conference (CONF-740807 VOL 1), the dose rate at the center of a finite hemispherical cloud of volume V cu ft is the dose rate from an infinite hemispherical cloud divided by a geometry factor (GF), where GF = 1173 /V 0" 3 3 8 Depending on area radiation monitor location, a portion of the dispersed radionuclides may be shielded by structures such as the steam generator enclosures.

This reduces the finite cloud volume.

Southern Nuclear Operating Company VEGP X6N1 SOMPERAN Unit: 1,&2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET3 BODY OF CALCULATION RU1 Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:

Emergency Action Levels: All (1 OR 2OR 3)1. Reading on ANY effluent radiation monitor greater than two times the ODCM limits for 60 minutes or longer.Liquid & Gaseous Effluent Monitors' Default Alarm Setpoints Per the guidance of NEI199-01 Rev 6 for Initiating Condition RU1 and its associated Emergency Action Level 1 (EAL 1), which are stated as follows: From page 25 of NEI 99-01 Rev 6: EAL #1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.As a release permit alarm setpoint is dependent upon the mix of radionuclides, their concentrations, the effluent release rate, the dilution stream flow rate, and the presence of concurrent contaminated releases, determining a default alarm threshold is infeasible.

The following liquid and gaseous release paths are continuous (per Tables 2-4 and 3-4 of VEGP ODCM), with release permits and alarm setpoints generated weekly (per 27AUG14 e-mail from Reggie Collins, VEGP 1&2 Chemistry Manager; copy in Attachment C3 of this calculation).

Liquid Releases Release Path Monitor SIG Blowdown RE-0021 Effluent Line Turbine Building RE-0848 Drain Effluent Line Gaseous Releases Release Path Monitor Plant Vent Stack RE-i12442C Plant Vent Stack RE-I124440 Turbine Building RE-12839 (SJAE)Thus, default setpoints are deleted from this version of the calculation.

Southern Nuclear Operating Company SOTENr Plant: VEGP Tite NI9-1Rv6ALClua ion6N1 CPNY Unit: 1&2 , SHEET 32 2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer: Liquid Effluent Release Path Monitors:

None Gaseous Release Path Monitors:

None Main Steam Line radiation monitors (RE-131 19 thru RE-13 122) were deleted per the 30 September 2014 license amendment (ADAMS # ML14170A911).

3. Sample analysis for gaseous or liquid release indicates a concentration or release rate greater than two times the ODCM limits for 60 minutes or longer.

Southern Nuclear Operating Company soutPEl Y. Unit: 1&2 Title: NEI 99-01 Rev 6 EAL Calculations SET3 RAI: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20R3OR4 1. Reading on ANY of the following radiation monitors greater than the readings shown for 15 minutes or longer.RE-12444E 0.50 pCi/cc RE-I12839E 2.1 x 101 pCi/cc This calculation is performed in Attachment K 2. Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the site boundary.3. Analyses of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the site boundary for one hour of exposure.4. Field survey results indicate EITHER of the following at or beyond the site boundary:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer* Analysis of field survey samples indicate thyroid ODE greater than 50 mrem for one hour of inhalation.

Southern Nuclear Operating Company SUHERN, 4.Plant: VEGP X6N1 SOUT en Unit: 1&2 ITitle: NEI 99-01 Rev 6 EAL Calculations SET3 RSI: Release of gaseous radioactivity resulting in offsite dose greater than 100 mREM TEDE or 500 mREM thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20OR3 1. VALID reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer: Radiation Vent Path Reading Monitor RE-I12444E Plant Vent -High Range 5 jiCi/cc RE-I12839E Turbine Building Vent (SJAE) -210 PCi/cc High Range Plant Vent (RE-12444E)

The threshold calculations are performed using the Excel spreadsheet in Attachment D1 of this calculation.

The thresholds calculated in the spreadsheets have been rounded off to one significant below to reflect the radiation monitoring system accuracy.100 mREM TEDE threshold

= 500 mREM Thyroid CDE threshold

=

Limiting threshold:

100 mREM TEDE = 5/,uCi/cc Steam Jet Air Ejector (RE-12 839)The threshold calculations are performed in the Excel spreadsheet in Attachment D2A (no core damage) and D2B (core damage) of this calculation.

The thresholds calculated in the spreadsheets have been rounded off to one significant below to reflect the radiation monitoring system accuracy.Dose Threshold No Core Damage Core Damage 100 mREM TEDE 2E+03 210 500 mREM Thyroid CDE 3E+07 4E+06 jiCi/cc Limiting threshold:

100 mREM TEDE, with core damage = 210 ,iCi/cc Southern Nuclear Operating Company SAT=m 1 Plant: VEGP ITteNE991Re6EACacliosj X6CNA15 S ;OUTLNY Unit: 1&2 TiteNE9-0Rv6EA Clua ion SHEET 35 2. Dose assessment using actual meteorology indicates doses greater than 100 mREM TEDE or 500 mREM thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the site boundary.* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Southern Nuclear Operating Company SUHm-- Plant: VEGP I X6CNA15 UnSO12 Til:UET9-1He 6ELacuaion SHEET 36 I COMPANY Unt & Til:NI9-1Rv6ELCluaos RGI: Release of gaseous radioactivity resulting in offsite dose greater than 1000 mREM TEDE or 5000 mREM thyroid CDE.Operating Mode Applicability:

Emergency Action Levels: All 1 OR20OR3 1. Reading on ANY of the following radiation monitors greater than the readings shown below, for 15 minutes or longer: Because the RG1 EALI dose limits are ten times the RS1 EALI dose limits, these threshold values are ten times the RS1 EAL1I threshold values.Radiation Vent Path RS1 EAL1 RG1 EAL1 Monitor Threshold Threshold RE-I12444E Plant Vent 5 50 j.+/-Ci/cc RE-I12839E Turbine Building 210 !.iCi/cc 2100 (Steam Jet Air Ejector)2. Dose assessment using actual meteorology indicates doses greater than 1000 mREM TEDE or 5000 mREM thyroid CDE at or beyond the site boundary.3. Field survey results indicate EITHER of the following at or beyond the site boundary.* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

Southern Nuclear Operating Company SOUTHE=B U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6EETNA15 CA1: Loss of RPV inventory.

Operating Mode Applicability:

Emergency Action Levels: Cold Shutdown, Refueling 1 OR2 1. Loss of RPV inventory as indicated be level less than elevation 1 85'-1 0" (73% on Full Range RVLIS).The RPV water level elevations corresponding to the RCS loop piping bottom IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2 x ID 13.75" Bottom ID = Centerline

-(1/2Ax ID) 185'-1 0.25" Hot Leg Inside Diameter 29.0" Sx ID 14.5" Bottom ID = Centerline

-(A x ID) 185'-9.5" The dimensions and elevations are taken from Design Input #9. The RPV water level elevation corresponding to the Bottom ID of the RCS piping is ~185'10".Because the core barrel is a right circular cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearly interpolating between the TOAF (EL 181"-10" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):

Southern Nuclear Operating Company s Plant VEGP ITitle: NEI 99-01 Rev 6 EAL Calculations j XCA5 S Unit: 1&2 SHEET 38 VEGP RVLIS Indication vs. RPV Water Level Elevation 80 75 IL. 70 80 183 184 185 186 RPV Water Level Elevation (feet)188 The RPV water level elevation corresponding to the Bottom~73% on Full Range RVLIS.ID is 185'-10" or 2. a. RPV level cannot be monitored for 15 minutes or longer AND b. UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT)or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.

Southern Nuclear Operating CompanyPlant: VEGP Titl:NI9-1Rv6ELCluain6N1 5 0~PIN Unit: 1&2 Til:NI9~ e A acltosI SHEET 39 0S1: Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

Emergency Action Levels: Cold Shutdown, Refueling 1 OR20OR3 1. a. CONTAINMENT CLOSURE not established AND b. RPV water level less than 1 85'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).The RPV water level elevations corresponding to 6" below the cold leg (CL) and hot leg (HL) bottom/IDs are found as follows: Dimension Elevation Loop Centerline Elevation 1 87'-00" Cold Leg Inside Diameter 27.5"1/2%x ID 13.75" Bottom ID = Centerline

-(1/2 x ID) 185'-1 0.25" 6" Below CL Bottom ID I185'-4.25" Hot Leg Inside Diameter 29.0" Sx ID 14.5" Bottom ID = Centerline

-(1/2 x ID) 185'-9.5" 6" Below HL Bottom ID 1 85'-3.5" The dimensions and elevations are taken from Design Input #9. The elevation corresponding to 6" below the Bottom ID of the RCS piping is ~185'4".Because the core barrel is a right circular cylinder, the RVLIS indication corresponding to the above RPV water level can be determined by linearly interpolating between the TOAF (EL 181"-1O" or 63% RVLIS) and the CL and HL centerline elevation (EL 187"0O" or 76% RVLIS):

Southern Nuclear Operating Company SO1J, Unit: 1&2 ITitle: NEI 99-01 Rev 6 EAL Calculations I SHEET 40 VEGP RVLIS Indication vs. RPV Water Level Elevation... ..... ..... .... Centerline

-7 __ -__ 6" Below RCS __--1........ .. .. ..........

.. .Piping Bottom~ID C 70 70~I 181 1 82 183 184 185 188 18 RPV Water Level Elevation (feet)The RPV water level elevation corresponding to 6" below the Bottom ID is 185'-4" or -72% on Full Range RVLIS.2. a. CONTAINMENT CLOSURE established AND b. RPV level less than 181'-1 0" [TOAF] (63% on Full Range RVLIS).3. a. RPV level cannot be monitored for 30 minutes or longer AND b. Core uncovery is indicated by ANY of the following:

RE-005 O...R 006 >40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Southern Nuclear Operating CompanyPant: VEGP Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 Unit: 1&2 S HEET 41 CG1 : Loss of RPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

Cold Shutdown, Refueling Emergency Action Levels: 1 OR 2 1. RPV water Level: a. Less than TOAF: EL 181'-1 0" or 63% on Full Range RVLIS.AND b. ANY indication from the Containment Challenge Table C1.Containment Challenge Table C1 CONTAI NMENT CLOSU RE NOT established*

Explosive mixture inside containment greater than OR equal to 6% H2 greater than OR equal to 13 psig WITH CONTAINMENT Containment PressureCLSResalhd greater than O..E equal to 52 psig WITH Tech Spec containment integrity intact* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.2. a. RPV water level cannot be monitored for 30 minutes or longer: AND b. Core uncovery is indicated by ANY of the following:

Monitor Location EAL Threshold RE-0005 or RE-0006 Containment Operating

> 40 REM/hr Deck High Range Source Range N/A Erratic Indication Southern Nuclear Operating Company smHm 1 Plant: VEGP Til:NI9-1Rv6ELCluaion6N1

• ui~ Unit: 1&2 TileHEE9-0ERv6TA Clcltin UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) level of sufficient magnitude to indicate core uncovery.AND c. ANY indication from the Containment Challenge Table C1 (above).Containment Operating Deck High Range (RE-O005 or RE-O006): This radiation monitor reading corresponds to the reflected dose rate from the irradiated fuel in the RPV with an RPV water level of less than TOAF (181 '-10" or 63% on Full Range RVLIS). It is calculated in Attachment E3 of this calculation.

Erratic Source Range Mon itor Indication Basis: NEI 99-01 R6, page 74.Explosive mixture inside containment

> 6% by volume hydrogen: Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.Pressure>_

14 psig WITH CONTAINMENT CLOSURE established:

NMP-EP-1 10-GL03 (pages 88, 90, 922, & 94) defines CONTAINMENT CLOSURE per uOperating Procedure 142 10-1/2, Containment Building Penetrations Verification

-Refueling." Section 6.0 of 14210-1/2 lists the acceptance criteria for CONTAINMENT CLOSURE, among them the requirement that >23' of water (EL 21 7"0") is maintained above the RPV flange. This corresponds to the SEP water low level (Design Input #4).During Refueling Operations, fuel is moved between the Containment and the Fuel Handling Building via the Fuel Transfer Tube. The Fuel Transfer Tube will maintain an air-to-air barrier during refueling operations if either the gate valve is closed or the water level in the refueling cavity is high enough to provide an air-to-air barrier.

If Containment pressure (PCTMT) exceeds the static head (Al-) due to the difference between the Transfer Tube centerline elevation (EL 186"-93/4";Design Inputs #4 & #5) andPTT the SFP low operating water level (EL 217"-0"; Design Input #4), the Transfer Tube air-to-air.

barrier is not maintained.--...-

A1H (ft) = 217"-0" -186'-9. 75" = 30'-2. 25" = -30 ft Pctmt (psig) > AH (ft) x p (Ibmlft) x g (ft/sec 2) x 1 ft 2 ge (Ibm-ft)/(Ibf-sec

2) 144 in 2 Pctmt (psig) > 30 ft x 61.55 Ibm x 32.2 (ft/sec 2 J x 1 ft 2 ift 32.2 (Ibm-ft)/(Ibf-sec

2) 144 in 2 (Design Input #25)Pc~r > -13 psig Pressure > 52 psig WITH Tech Spec containment integrity intact NMP-EP- 11 0-GLO3 (page 8) defines CONTAINMENT INTEGRITY as The Primary Containment is OPERABLE per Technical Specification

3. 6.1.1." Tech Spec surveillance requirement

3. 6.1.1 states "Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program." Tech Spec section 5. 5.17 describes the Containment Leakage Rate Testing Program. Per Tech Spec Bases B3. 6.1, the Containment is designed to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA).Per section 3.1.3.1 of VEGP Design Criteria DC-2101, the mechanical (piping) and electrical penetrations, in conjunction with the carbon steel liner, form a leak-tight barrier. Thus, these penetrations must meet the design accident pressure requirement of section 3. 4.5 of DC-2101, 52 psig.The absence of air-to-air containment penetrations during Modes 5 and 6 in VEGP procedure 142 10-1 was confirmed by e-mails from John Stanley (VEGP Operations Outage Manager; see Attachment C5) and Ron Cowen (Westinghouse Site Services Manager," see Attachment C6).

Southern Nuclear Operating CompanyPlant: VEGP X6N1 CPAY Unit: 1& il: E 90 Rev 6 EAL Calculations SHEET 44 SOU MPATil:NEI990 Piping Penetrations The piping penetrations are listed in drawings 1X4DL4A0 13, 1X4DL4A0 14, 2X4DL4A0 13, and 2X4DL4A0 14. Cross-sectional views are shown in drawings 1X4DL4A014 and 2X4DL4A0 14.Per section 4.1.2 of specification X4AQIO, these penetrations provide part of the containment boundary.

Section 4.1.3.3.2 of this specification directs the user to Attachment 2 for the design temperature and pressure for these penetrations.

Per Attachment 2 of specification X4AQIO, the emergency operation design containment pressure is 50 psig.The evaluation in Attachment F of this caic demonstrates that the pipe penetrations should not fail due to a containment pressure of 52 psig.VEGP Condition Report 876376 has been submitted to review and resolve this difference between the Containment and the piping penetration accident design pressure criteria.

There are no operability or functionality issues because the peak containment DBA pressure is ~37 psig (VEGP FSAR Tables 6.2.1-1 & 6.2.1-66).

Electrical Penetrations Per section 3.1.3 of DC-1818, the electrical penetration assemblies shall withstand the pressure, temperature, and environmental conditions resulting from a DBA without exceeding the electrical penetration design leakage rate.Per section E3. 6.2 of specification X3ARO1-E3, the electrical penetration design leakage rate is 0.01 cc/sec at DBA conditions.

CONTAINMENT CLOSURE no.t established.

Basis: NEI 99-01 Rev 6, page 81.

Southern Nuclear Operating Company SOUTHERN U1 lnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNA15 I E-HUI Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability:

ALL Emergency Action Level: 1 1. Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table El. Readings are combined gamma + neutron.Location A H I-TRAC 125 Side -Mid -height 950 Top J 200 HI-STORM 100 Side -60 inches below mid- 170 Side -Mid -height 180 Side -60 inches above mid- 110 Top -Center of lid 50 Top -Radially centered 60 Inlet duct 360 Outlet duct 130 This calculation is performed in Attachment L

Fission Product Barrier Emergency Action Levels Fuel Clad Barrier Fuel Clad Barrier Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 > 2.6E+5 mR/hr The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300pCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the fuel clad barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss Threshold 3.A since it indicates a loss of both the fuel clad barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

Evaluation of Containment High Range Rad Monitor Locations Readings at Power A review of the instrument location drawings (Unit 1:1IX5DS4BO02

& 1X5DS4DO02; Unit 2: 2X5DS4BO02

& 2X5DS4CO02) indicates the rad monitors 1/2RE-0005

& 1/2RE-0006 are located above the Containment operating deck of EL 220'-0".The following full power area rad monitor data demonstrates that the monitors do not measure radiation from nearby piping and components containing elevated reactor coolant activity.

The data are documented in Attachment G2; Attachment G3 documents that the plant was at 100%power.

Unit 1 1 RE-005 1 RE-006 Tie Reading Tie Reading Tie (mREMI~hr)

Tie (mREM/hr)12:00 505 12:00 250 12:10 484 12:10 232 12:20 509 12:20 193 12:30 511 12:30 189 12:40 502 12:40 240 12:50 499 12:50 238 13:00 493 13:00 250 Average 500 Average 227 Unit 2 2RE-005 2RE-006 Tie Reading Tie Reading Tie (mREMIhr)

Tie (mREM/hr)12:00 42.2 12:00 19.2 12:10 29.7 12:10 21.3 12:20 53.3 12:20 23.1 12:30 54.9 12:30 23.9 12:40 47.4 12:40 15.4 12:50 61.3 12:50 18.8 13:00 52.8 13:00 16.6 Average 48.8 Average 19.8 The differences between the Unit 1 and Unit 2 readings are due to their installation locations as shown on the next sheet. The Unit 1 rad monitors are mounted on the S/G enclosures on opposite sides of the reactor vessel. The Unit 2 rad monitors are mounted near the outer containment wall, further away from the reactor vessel as compared to the Unit 1 monitors.Containment Geometry While the released radioisotopes are divided by the total containment volume (Vctmt), the operating deck (EL 220'-0") area radiation monitors locations limit them to "seeing" the sprayed portion of the containment volume. This view may be further reduced by containment structures.

As a result, the containment volume used to determine the geometry factor GF is expressed as: Vmon = fview x fspray x Vctmt where Vmon = containment volume "seen" by area radiation monitor (cu ft)fvwew = fraction of containment volume above operating deck "seen" by area radiation monitor fsry= fraction of containment volume above operating deck = 0. 771 Vctmt = Containment volume = 2. 95E+06 cu ft The fraction of the sprayed containment volume seen by the area radiation monitors is estimated below.The containment operating deck area radiation monitor locations identified in drawings 1X5DS4BO02, 1X5DS4DO02, 2X5DS4BO02, and 2X5DS4CO02 are overlaid on the containment structural drawings below:

~Southern Nuclear Operating Company=UI Al 4 Plant: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNAI5 I s@J~ham, Unit: 1&2 SHEET 48 Unit I Unit 2 While the Unit 2 area radiation monitors have a fairly unrestricted "view" of the containment volume above the operating deck, the Unit I area radiation monitors are each screened from a significant portion. The difference between the Unit 1 and Unit 2 radiation monitor locations explain the differences between their full power readings.The fraction of the sprayed containment volume seen by the Unit I area radiation monitors is fvjew = I -A segmnen/A op deck where A segment = area of segment outboard of dashed gray line Aop, deck = area of operating deck = (Hix lDctmnt 2)/4 IDctrnt = containment inside diameter = 140 ft The area of the segment is calculated as shown below.

A circular segment (in green) is enclosed between a chord (dashed line) of length c and the arc of length s shown above the green area.R = radius of circle = % x lDctmt R = 70Oft e = central angle (degrees)c = chord length s = arc length h = height of segment d = height of triangular portion S

References:

http://en.wikipedia.orq/wiki/Circular seqment and Geick&Geick, "Engineering Formulas" The central angle is The radius is 6 = 2 arccos = 2 arcsin2-R R =h+d The height of the triangle is d =R -h Area of segment is 80o Scaling from drawing 1X2D48EO05, h = 50 ft and d = 20 ft. The chord length c is determined using the Pythagorean Theorem: (c/2)2 + d 2 = R (c/2)2 = R 2_ cl c/2 = [R 2 -d 2c = 2 x [R 2 -l1 c = 2 x [(7 0)2- (20)21 c =134 ft e = 2 x arcsin[c/2R]

e = 2 x arcsin[(134)/(140)]

e = 2 x arcsin[O.

958]eO= 2x 73.40 Southern Nuclear Operating CompanyPlant: VEGP I X6CNA15 sOuy pnay Unt1& Title: NEI 99-01 Rev 6 EAL Calculations SET5 6 = 146.80 A segment = [R 2/2] X [(917/180)

-sinS7]A segment = [(70)2/2] x [(146. 8 x 17/180) -sin(146. 8)]A segmnent = [2450] x [(2. 562) -(0. 548)]A segment = [2450] x [(2.014]A segment = 4935 sq ft A op deck = Hl1Dctmt 2/4 A op deck = H7(140)2/4 Aop deck = 15,394 sq ft fview, = I -A segmnent/A op deck fview = 1 -- (4935/15,394) fview = 1 --0.321 fview = 0.6 79 For Unit 2, fview, = 1.0O.Thus Vmon = fview X fapray x Vctmt GE = 11 73AVmon 0" 3 3 8 The calculation of the geometry factor will be performed in the spreadsheets in Attachments H and I. Performing it here using a calculator will produce different results than the spreadsheet due to round off errors.RCS Barrier RCS Barrier Loss Threshold 3.A Containment radiation monitor RE-005 OR 006 >= 8.7E+2 mR/hr The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold 3.A since~it indicates a loss of the RCS Barrier only.Evaluation of the Fuel Clad Failure Southern Nuclear Operating Company ISOUTHERN Uk lnit: VEGPI Title: NEI 99-01 Rev 6 EAL Calculations I SHETA15 This threshold is determined assuming fuel clad failure and resultant instantaneous release and dispersal into containment the reactor coolant noble gas and iodine inventory, with RCS activity at the Tech Spec allowable limits.The iodine concentrations are set equal to the corresponding to the 1.0 ,uCi/g DE/1-131 values.The corresponding noble gas reactor coolant concentrations are expected to be proportional to the DE I-131 concentrations.

The ratio of each Iodine radionuc/ide's reactor coolant DE/1-131 concentration above is divided by its equilibrium reactor coolant concentration.

The minimum ratio is then used to multiply the noble gas equilibrium reactor coolant concentrations.

RCS Concentrations (pCi/g)Iodies quilriu 1 Ci/g (1 i.LCi/g)/lodnes Equibrum DE 1-131 Equilibrium 1-131 2.91 0.74 0.25 I-132 2.96 0.75 0.25 I-133 5.56 1.41 0.25 1-134 0.69 0.18 0.26 1-135 2.72 0.69 0.25 Minimum Ratio =0.25 RCS Concentrations (pCi/g)RCS Concentrations (pCiIg)NolI 1 Nobles Equilbrium DE 1-131 Noble Gases Equilbrium DE 1-131 GssEquivalent Equivalent Kr-85m 2.04 5.17E=-01 Xe-131m 2.02 5.121E-01 Kr-85 8.37 2.12E=+00 Xe-133m 17.6 4.46E+00 Kr-87 1.28 3.24E=-01 Xe-133 256 6.491E+01 Kr-88 3.68 9.321E-01 Xe-135m 0.56 1.421E-01 Xe-135 8.3 2.10E+00 Xe-138 ] 0.74 [1.88E-01 These concentrations

(/uCi/g) are then multiplied by the RCS coolant mass (grams) then divided by the containment volume (in 3) to determine the containment source strength (Ci/m 3).The resulting Unit I and 2 containment high range rad monitor dose rates for immersion in a finite hemispherical cloud of air are calculated in Excel spreadsheets in Attachments H2 and 12, respectively, of this calculation.

They uses the applicable FGR 12 Table III. 1 DCFs for immersion in a semi-infinite cloud of air and the geometry factor (GE) from sheet 30 of this Southern Nuclear Operating Company SOUTHEN ln:VG Title: NEI 99-01 Rev 6 EAL CalculationsX6N1 I COMPANY Unit: 1&2 I SHEET 52 calculation to convert the semi-infinite (infinite hemispherical) cloud dose rate to a finite hemispherical cloud dose rate.The results of the Loss of RCS FP Barrier setpoint calculations in Attachments H2 and 12 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (REM/h r) (m REM/hr)VEGP 1 0. 870 8. 7E+02 VEGP 2 0. 991 9.9E+02 Containment Barrier Containment Barrier Potential Loss Threshold 3.A Containment radiation monitor RE-005 OR -006 >= 1 .3E+7 mR/hr.The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous fuel clad barrier loss and RCS barrier loss thresholds.

NUREG-1 228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20 percent in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS barrier and the fuel clad barrier. It is therefore prudent to treat this condition as a potential loss of containment that would then escalate the emergency classification level to a General Emergency.

Evaluation of the Potential Loss Threshold This setpoint is based on the release of all iodine isotopes and noble gases corresponding to 20% fuel clad failure. The core wide fuel rod gap noble gas inventories of FSAR Table 15A-3 are multiplied by this failed fuel clad fraction to determine the released activity.

This released activity is then divided by the containment volume to determine the source strength (Ci/m 3).The resulting Unit I and 2 containment high range rad monitor dose rates for immersion in a finite hemispherical cloud of air are calculated in Excel spreadsheets in Attachments H2 and 12, respectively, of this calculation.

They uses the applicable FGR 12 Table 1lL.1 DCFs for immersion in a semi-infinite cloud of air and the geometry factor (GE) from sheet 30 of this calculation to convert the semi-infinite (infinite hemispherical) cloud dose rate to a finite hemispherical cloud dose rate.

Southern Nuclear Operating Company SOUTHERN~

UPlnit: VEGP Title: NEI 99-01 Rev 6 EAL Calculations X6CNA153 The results of the Loss of Clad FP Barrier setpoint calculations in Attachments H3 and 13 are summarized below. Given the system accuracy -a factor of two over the operating range -the threshold is rounded off to two significant figures.Unit Calculated Threshold Rounded-Off Threshold (REMIhr) (m REM/h r)VEGP I I.31E+04 1.3E+07 VEGP 2 1.49E+i04 1.5E+07 Containment Barrier Potential Loss Threshold 4.B Containment Hydrogen concentration greater than 6%.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.

It therefore represents a potential loss of the containment barrier.Sheet 23 of VEGP SAMG calculation X6CNA 11 established the 6% by volume hydrogen limit.

Southern Nuclear Design Calculation SPlant: Vogtle Unit: 1&2 ICalculation Number: X6CNAI5 Sheet: A-I Attachment A -SNC Emergency Planning Concurrence Calculation Number: X6CNAI5 Calculation Version: 1 Calculation Title: NEI 99-01 Rev 6 EAL Calculations I the undersigned have reviewed the subject calculation and concur that:* Its Methods of Analysis conform to the guidance of NEI 99-01 Revision 6* Its Assumptions are consistent with the guidance of NEI 99-01 Revision 6* Its conclusions are consistent with the Methods of Analysis, Assumptions, and Design Inputs.g~4. c.II~:~-M~@d~rr 9111t'Name 1 / j(,. I SNC Emergency Planning/ Signature

/ Date / Organization Southern Nuclear Design Calculation Plant: Votl Unit: 1&2 Calculation Number: X6CNA15 Sheet: C-I ATTACHMENT C -REFERENCES DescrptionNumber Descrptionof Pages C1 -Davisson, "Gamma Ray Dose Albedos," from "A Handbook of Radiation 13 Shielding Data" 02 -Validation of Spirax Sarco On-Line Steam Tables 10 03 -Collins-Bornt e-mail, "Re: Vogtle EAL Setpoints," 27AUG14 2 04 -Collins-Bornt e-mail, "Re: Turbine Building Vent Release Permit -1/2RE- I 128390," 28AUG14 C5 -Stan ley-Bornt e-mail, "RE: Vogtle Containment Penetrations

-Cold Shutdown 2& Refueling Modes," 04SEP14 C6 -Cowman-Bornt e-mail, "RE: Vogtle Containment Penetrations

-Cold Shutdown 2& Refueling Modes," 05SEP14 1-4-1-4-1-4-Total Number of Pages Including Cover Sheet 3 31