Text

to License Amendment Request 286, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors
	ML17233A283
Person / Time
Site:	Point Beach
Issue date:	08/21/2017
From:	Coffey R; Point Beach
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NEI 99-01, Rev 6, NRC 2017-0041
	Download: ML17233A283 (118)
	v • d • e

NEXTera ENERGY .

~*

BEACH August 21, 2017 NRC 2017-0041 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Supplement 1 to License Amendment Request 286, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors"

References:

(1) NextEra Energy Point Beach, LLC letter to NRC, dated June 23, 2017, License Amendment Request 286, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01 , Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" (ML17174A458)

(2) NRC electronic mail to NextEra Energy Point Beach , LLC , dated July 20, 2017, LIC- 109 Acceptance Review Supplement Request-Point Beach Units 1 and 2 LAR 286 -Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" (MF9859/MF9860)

In Reference 1, NextEra Energy Point Beach, LLC (NextEra) submitted a request for an amendment to revise the facility operating licenses for the Point Beach Nuclear Plant (PBNP)

Units 1 and 2. Specifically, the proposed change involves revising the Emergency Plan for PBNP to adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC .

In Reference 2, the NRC staff requested supplemental information to complete its review of the requested amendment. Enclosures 1 through 3 provide the NextEra response to the NRC staffs request for supplemental information .

Additionally, the NRC staff requested via teleconference , the basis for the changes of some Alert classifications. This license amendment will result in changing the EAL thresholds for each of the Alert classifications based on release point monitor readings in the absence of dose calculation capability . This is a direct result of the approved guidance in NEI 99-01 Revision 4 (i.e. 200 times the ODCM alarm set point) being changed in Revision 6 to 1/100 of the EAL threshold for declaring a General Emergency."

This letter contains no new regulatory commitments and no revisions to existing regulatory commitments.

This response to the request for supplemental information does not alter the conclusion in Reference 1 that the proposed change does not involve a significant hazards consideration.

NextEra Energy Point Beach, LLC 6610 Nuclear Road , Two Rivers, WI 54241

Document Control Desk Page 2 If you have any questions regarding this letter, please contact Mr. Eric Schultz, Licensing Manager at (920) 755-7854.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on August 21 , 2017.

Sincerely, o ey Site Vice President Point Beach Nuclear Plant Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENTAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" CALCULATION NEE-363-CALC-001, REVISION 0 DOSE RATE EVALUATION OF REACTOR VESSEL WATER LEVELS DURING REFUELING FOR EAL THRESHOLDS ,

49 pages follow

NEE-363-CALC-001 , Rev. 0 Attachment 2 Page 1 of 5 Design Information Transmittal From: Point Beach Design Engineering To: ENERCON DocumenU EC/ Tracking Date: 4/10/2017 DIT No: n/a EC 0288568 Number:

Document

Title:

Calculations NEE-363-CALC-001 (Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds) and NEE-363-CALC-001 (Fuel Handling Accident Monitor Res~onse for EAL Thresholds}

Quality Facility/ Unit: PBNP/0 QR Classification

SUBJECT:

This Design Information Transmittal (DIT} is to provide ENERCON with requested inputs for use in calculating EAL thresholds.

Check if applicable:

D This DIT confirms information previously transmitted on N/A under DIT-001 Draft Attachment 1 and 2.

Additional information has been added since DIT-001 Draft was sent.

D This information is preliminary. See explanation below.

SOURCE OF INFORMATION:

All sources documented in attached text.

DESCRIPTION OF INFORMATION:

Various source terms, geometries, dimensions, etc. of the fuel, reactor vessel, and core internals . Details provided in the attached text.

To ensure complete documentation, this DIT shall be included with any QA documents (e.g. calculations) citing it as a source of information.

DISTRIBUTION (Recipients should receive all attachments unless otherwise indicated . All attachments are uncontrolled unless otherwise indicated)

Jay Bhatt (ENERCON)

Ryan Skaggs (ENERCON)

PREPARED BY (The Preparer and Approver may be the same person .)

Tom Kendall Principal Engineer

//<<L/A/

'-1;1¢/t 1 Preparer Name Position Signature Date VERIFIED BY (Design verification is required if the information is not a verified design output. Verification is also required if the information is developed , interpreted , or extracted from an unverified source. Otherwise, N/A).

Steve Bach Senior Engineer

~>At;t.~ ~ //

'Y)~/17.

Verifier Name Position Signature Date APPROVED BY (The cognizant Engineering Supervisor has release authority. Consult the Design Interface Agreement or local procedures to determine who else has release authority.)

Civ/Mech Design Jane Marean

.Eng . Su~ervisor 1\r-U~~oe:::::

.. '-(*1 f

f (

Approver Name Position &-nature\ Date A copy of the DIT (along with any attachments not on file) should be included with the associated EC or document record .

EN-M-1 00-1 003-F02 Revision 0 Page 1 of 1

CALCULATION COVER SHEET (Page 1 of 1)

Document Information:

Calculation (Doc) No: NEE-363-CALC-001 I Controlled Documents Revision: 0

Title:

Dose Rate Evaluation of Reactor Vessel Water Levels During Refueling for EAL Thresholds Type: CALC Sub-Type: Discipline: MECH Facility: PB I Unit: 0 Safety Class: D SR D Quality Related [:g] Non-Nuclear Safety D Important to Safety D Not Important to Safety Special Codes: 0 Safeguards D Proprietary Vendor Doc No : NEE-363-CALC-001 I Vendor Name or Code: ENERCON Executive Summary (optional):

Review and Approval:

Associated EC Number: 288568 EC Revision: 0 ARJ Other Document Number:

Description of Calculation Revision: Initial issue EC Document Revision: 0 Prepared by: N/A (Vend or) Date: - -

(signature) (print name)

. /;,..0/ ///

Rev1ewed by: * . f:~

~ .

7.c, kft1/i?A Ll- Date:0;6q.

(signature) (print name)

Type ofReview: D Design Verification 0 Review ~ Owner Acceptance Review Method Used (For DV Only): 0 Design Review D Altemate Calculation

\

Approved by: c-'\, \:.:::~ ,, --.Jc. ~, ti:.,. < Date: .s: 6

17

\

(sigbature) I (print name)

EN-AA-1 00-1 004-F01, Revision 0

CALCULATION REVISION

SUMMARY

SHEET (Page 1 of 2)

Calculation Number: NEE-363-CALC-001 Rev. Affected Pages Reason for Revision 0 All (49 pages) Initial issue. Detailed contents:

Site cover sheet (I page)

Rev. summary sheet (I page)

Vendor calculation (37 pages)

Vendor Calculation Preparation Checklist (5 pages)

Design Information Transmittal (5 pages)

EN-AA-1 00-1 004-F02, Revision 1

CALC NO. NEE-363-CALC-001 0 ENERCON E II.Cf:o'fMr~ - f~*Ny (il 0)~,,... L EvtiY rl::})'

CALCULATION COVER SHEET REV. 0 PAGE NO. 1 of 37 Dose Rate Evaluation of Reactor Vessel Client: NextEra - Point Beach

Title:

Water Levels During Refueling for EAL Thresholds Project Identifier: NEE-363 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary D ~

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

2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design D ~

verified calculation.)

Design Verified Calculation No.

3 Does this calculation supersede an existing Calculation? (If YES, identify the design [8]

verified calculation.)

0 Superseded Calculation No.

Scope of Revision:

Initial Issue Revision Impact on Results:

Initial Issue Study Calculation 0 Final Calculation ~

Safety-Related 0 Non-Safety-Related ~

/J /2 tnnt Name and Sign)

[1/Jl/ 1t01 1I Originator: Jay Bhatt Design Verifier1 (Reviewer if N~): Caleb rarnor *e ~*(~/

/_ A- ~

Date:

~

4 /! 1!/; 7 Approver: Aaron Holloway~~ ---- ~_.;;;?---------- Date:

11/;)1/17 I I Note 1: For non-safety-related calculation, de~on can be substituted by review.

CALC NO. NEE-363-CALC-001

" ENERCON CALCULATION Excellence-Every project. £very doy. REVISION STATUS SHEET REV. 0 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 4/21/17 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 0 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO. OF REVISION ATTACHMENT NO. OF REVISION PAGES NO. NO. PAGES NO.

A 1 0 1 5 0 8 2 0 2 5 0 c 1 0 Page 2 of 37

CALC NO. NEE-363-CALC-001

-.~:d ENERCON Excellence- Every project. Every day.

TABLE OF CONTENTS REV. 0 Section

1. Purpose and Scope .................................. .. ..... ..... ..... .......... ....... ... ... ....... ..... ........ ... ... ...... . 6

2. Summary of Results and Conclusion ... .. ........ .......... ... ... ........ ... ..... .. ... ........ .... ....... .... .... ... 6

3. References .... ................ .......... ..... ... ... .. ..... ............ ........ ... .. ..... ........... .... ............... ... .... ..... 7

4. Assumptions ..... ..... ... ...... ... ... .. ... ... .......... ..... ............ .. ..... .. ..... ......... ........... ....... ........ .. ... .... 8

5. Design Inputs ......... .......... ... .. .. .. ........... .. ... ................. .. ............. ......... ... ...... ... ...... ... ..... ... 10 5.1 Fuel Assembly Parameters .. ... .. .. .. ............... ........ .... .. ...... ............................ .... ... 10 5.2 Containment Dimensions .. ....... ......... ..... .......... .... ...... .. ..................... ............. ..... 10 5.3 Core Isotopic Inventory ..... .. .. .. .... .... ....... ........ ....... ......... ............ ....... ........ ... ... .. .. 13 5.4 Material Compositions .... ... ..... ....... ............. .... .................... .. .. ... .............. ....... .... . 14 5.5 Upper Internals/Upper Fuel Hardware ............... ... .... .. .... .... .... .... ....................... .. 15 5.6 Containment High Range Detector Range ..... ........... ..... .. ... ... ...... ............ ........... 15 5.7 ANSI/ANS- 1977 Flux to Dose Factors ........ ..... ..... ........ .. ......... .. ................. ..... . 15

6. Methodology ..... ..... .. ........... ... .............. .. ............. .. .. .. .............. .. ...... ....... .. ...... .... ....... ....... 16

7. Calculations ...... ......... ... ............ ....... ...... ........ ... .......... .. .. ......... .... ....... .... .... ..................... 17 7.1 Source Terms .... ... .. ... .. ... .. ..... ............ ....... .... ................ ............ ............... ............ 17 7.2 MCNP Model Core Homogenization ............................ .. ........ ...... ...................... . 19 7.3 MCNP Model Upper Internals Homogenization ..................................... .... .......... 20 7.4 MCNP Model Geometry ...... .. ... ......... .. .. ........ .. ..... ... .... ..... .. .. ......... ... ....... .... .. ...... 21 7.5 MCNP Source Definition ............. ... ..................................................................... 28 7.6 MCNP Tally Specification .... .. .. ......... .. ....................................... ...... ............ ...... .. 29 7.7 MCNP Material Cards ... .................... .......... ............. .......... .... ..... ................ ......... 30 7.8 Results ...................... ... .. ... ...... .... ... .... .. ... .. ........ ..... ........... ... ...... .. .... ............ ... .... 31 7.8.1 Results without Head .. .... .............. .......................................................... .......... .. 32 7.8.2 Results with Head .................... ...... ... ........................ ..... ........... .......... ... ..... ... ..... 33

8. Computer Software ........ .... .. .. ...... .. ..... ... .......... ..... ........ .... .... ... ....... .. ...... ... .......... ....... .. .. 33 Appendix A -Electronic File Listing ......... .. .............. .. .......................................... .... .............. ..... . 34 Appendix B- PB.xlsx Sheets ................... ... ........ ...... ........... .. ..... ............ ... ...... ... ... .............. ...... 35 Appendix C- ORIGEN-S Input for Source Term Calculation ......................... .. ........................... 37 Page 3 of 37

CALC NO. NEE-363-CALC-001 F..:d ENERCON TABLE OF CONTENTS EJ:cellence- Every prOJect. Evety day.

REV. 0 List of Figures Figure Figure 7-1 Y-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No Head) .................. 22 Figure 7-2 Y-Z VISED Plot of Containment ........ .... ... ....... ............. ...................... ........... .... .... .... 22 Figure 7-3 Y-X VISED Plot of the Containment Geometry at 66' Elevation Level ... ........ ........ ... 23 Figure 7-4 MCNP Model Surface Cards ............. .... .. .. .... ....... ........ ..... ....... .. ........ .... .. ....... ....... ... 25 Figure 7-5 MCNP Model Cell Cards (No Head) .. .......... ......... ..... ................... .. .. ........ ..... ..... ... .... 26 Figure 7-6 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (With Head) .. .. .... .... .. . 27 Figure 7-7 MCNP Source Definition Cards .. .................. ... ...... .......... ........ ..... ............... ... .. ... .... .. 28 Figure 7-8 MCNP Tally Cards ....................... .................. .. ...... .... ... ... .. .... ................................. ... 29 Figure 7-9 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors .......... ...... ............ 29 Figure 7-10 MCNP Material Cards .. ........................ ... ....... .. ....... .. ..... ..... ....... ......... ... ...... ... ......... 30 Figure 7-11 Dose Rate versus Water Height Plot for no Head Configuration .. ............ ............ .. . 32 Page 4 of 37

CALC NO. NEE-363-CALC-001 f.:jl ENERCON TABLE OF CONTENTS Exceflence-Every project. Every day.

REV. 0 List of Tables Table 2-1 Dose Rate at Top of Active Fuel. ........... ... .. .... .... ................ ..... ........... .. ..... ... ...... ... .... ... 6 Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Fuel .. .. ...... .................... 10 Table 5-2 Design Input Containment Elevations ................................................... .............. .. ...... 10 Table 5-3 Design Input Containment Dimensions .................. ........................................ .... .... .... 11 Table 5-4 Core Shutdown Source Term .. .. .. ...................... .. .... ....................................... ...... .. .. .. 13 Table 5-5 SCALE Standard Compositions used in MCNP Model. .. .. ....... .. ...... ........................... 14 Table 5-6 ANSI/ANS-6.1.1-1977 Flux to Dose Factors .. .. .......... ...... ...... ................ ...... .......... .. .. 15 Table 7-1 Binned Total Core Source Term .... .. ........ .. .... .. .. .............. .................................... ....... 18 Table 7-2 Homogenization of Active Fuel Region ...... .................. .. ...... .... ................................... 20 Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (mrem/h)

..... ... .... ....... ..... ............... ... .. .. ....... .. ...... ... .. ........ ..... .... ....... ...... ..... .. .. ................ .. ................ .. .. .. ... 32 Table 7-4 Dose Rate Response for Head in Place Configuration (mrem/h) ............................... 33 Page 5 of 37

Dose Rate Evaluation of CALC NO. NEE-363 -CALC-001 F.;~ ENERCON Reactor Vessel Water Levels Exc~llence- Every p rojecr. Every day. During Refueling for EAL REV. 0 Thresholds

1. PURPOSE AND SCOP E The purpose of this calculation is to evaluate dose rates as a function of water height in the reactor vessel during cold shutdown or refueling operations in order to set Emergency Action Level (EAL) thresholds for core uncovery (RA2 , CS1, CG1 ). The dose rates are calculated at the locations of the containment monitors RE-126 , RE-127 and RE-128 so that dose rate measurements by these devices can be correlated to the water level in the core, upon failure of other water level detection systems. This calculation will determine the dose rate at full core uncovery, as well as maximum water levels with a detectable dose rate response applicable to both Unit 1 and Unit 2. This calculation is not Nuclear Safety Related as the results of the calculation do not affect the design basis or Safety Related systems structures or components . These results are best estimates based on as-built conditions and provide information to operators with respect to classifying an emergency , therefore no acceptance criteria is required.

2.

SUMMARY

OF RESULTS AND CONCLUSION The dose rate results for the configuration without the reactor vessel head and with the reactor vessel head are provided in Section 7.8.1 and Section 7.8.2, respectively. The minimum dose rates with the core uncovered (i .e. water at the top of the active fuel) are shown in the table below.

The dose rates reported below do not include the ambient readings associated with the monitor calibration (generally 1 to 2 R/h).

Table 2-1 Dose Rate at Top of Active Fuel Model Description Dose Rate (R/h)

Head Off 1.09E+02 Head On 2.94E-02 1 Detailed results of the dose rate as a function of water height are provided in Table 7-3 for cases with the head removed.

1 For the case with the head in place , the dose rate is below the detectable range of the radiation monitors of 1 R/h .

Page 6 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001

-:d ENERCON Reactor Vessel Water Levels Excellence- Every project. Every day. During Refueling for EAL REV. 0 Thresholds

3. REFERENCES

1. "Standard Composition Library ," ORNLINUREG/CSD-2N1/R6 , Volume 3, Section M8, March 2000

2. Design Information Transmittal (Attachment 2)

3. CGDG-SCALE-6 .1.2, Revision 00 , Commercial Grade Dedication SCALE Version 6.1 .2

4. C-2128 , Rev. 8 Cant Interior Plans El10ft., 21 , 24ft. Sin. & 38

5. CGDG-MCNP6-V1 .0, Revision 00 , Commercial Grade Ded ication MCNP6 Version 1.0

6. ANSI/ANS 6.1.1-1977, Neutron and Gamma Flux-To-Dose Conversion Factors

7. C-128, Rev. 10, Cant Interior Plans El10ft., 21ft., 24ft.8in . & 3

8. C-129 , Rev. 10, Cant Interior Plans El46ft. 66ft. 76ft. & 100

9. M-9, Rev. 13, U 1 Equipment Location Plans Sections A-A & B-B

10. LF-01 DP1 01 , Rev. 3, Closure Head General Assembly 11 . M-500 , Rev. 8, Cant Operating Floor Mise Upper Floor South

12. C-134, Rev. 13, Cant Interior Reinforcing Sections

13. FSAR Table 3.2-5, Core Mechanical Design Parameter

14. FSAR Table 14.3.5-1, Core Activities

15. FSAR Section 5.1.2.1, Containment System Structure Design -General Description

16. FSAR Table 11.5-1B, Radiation Monitoring System Area Monitors

17. M-2500, Rev. 6, Cant Operating Floor Mise Uppers Floors North

18. C-2129, Rev. 8, Cant Interior Plans El 46ft. 66ft. 76ft. & 100 Page 7 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001 F."~

I ' ENERCON Reactor Vessel Water Levels Excellence - Every project. Every day. During Refueling for EAL REV. 0 Thresholds

4. ASSUMPTIONS The following assumptions are used in the core uncovery dose rate calculation :

1. The core is homogenized based on the typical Westinghouse 14x14 fuel assembly dimensions, taking into account the fuel rods and space between . Any small variations in fuel parameters will have a negligible effect on containment dose rates. The cladding is modeled as Zircaloy 4 in lieu of ZIRLO; this is acceptable due to the similarity of the materials.

2. Any non-fuel hardware is ignored in the active fuel region, since the primary self-shielding occurs in the fuel itself, and there may be some unknown streaming effects through the non-fuel hardware. This homogenization takes into account the presence of water when calculating the isotopic weight fraction and homogenized density. For the case with the reactor vessel head in place, the region between the head and the active fuel region is homogenized based on the actual mass of the upper internals over the entire region. Homogenization of source regions and shields is acceptable for determining the best estimate response at the detector locations.

3. The compositions of the containment structure and components are based on the values in the SCALE standard composition library [1]. These material properties are acceptable for modelling the structures and components used to determine the best estimate response at the detector locations.

4. The containment outer concrete thickness is modeled with a thickness of 3 feet.

This value is chosen upon inspection of the drawings [7, 8] and contributes to the dose rate through backscattering only. Variations in concrete thickness would not have a large effect on the calculation results.

5. Based on a review of recent refueling outages, the minimum time to start fuel movement is 121 hours0.0014 days 0.0336 hours 2.000661e-4 weeks 4.60405e-5 months [2]. This calculation assumes a decay time of 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months to allow EAL thresholds to be determined for reactor vessel conditions that exist prior to the commencement of fuel movement which is representative of the applicable operating modes (cold shutdown, refueling) . This decay time is appropriate to produce best estimate results for both the head on and head off configurations.

6. It is assumed that the Unit 1 configuration approximates the Unit 2 configuration sufficiently such that only a single evaluation is performed to obtain best estimate results . This is acceptable based on review of the Unit 2 drawings [4, 18]. The detectors are modeled at the Unit 1 locations. The Unit 2 detectors are located at the same elevation (66'-0"), and are similarly located along the containment wall

[17]. The difference between the Unit 1 and Unit 2 detector locations will not significantly change the results for the calculation, as the distances and amount of shielding between the source and detectors are relatively similar.

7. The hardware in the upper internals region between the active fuel region and reactor vessel head is assumed to be stainless steel type 304. While the actual Page 8 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.:~ ENE RCON Reactor Vessel Water Levels Excellence - E~wy project. Evety doy. During Refueling for EAL REV. 0 Thresholds composition of the hardware may vary slightly, small variations in the material will have a negligible effect on the dose rate response at the detectors.

8. It is assumed that the water above the active fuel region is liquid at a constant temperature . Using the density in Table 5-5 is common in shielding applications.

9. The source term is generated shortly after shutdown , therefore, the fuel gamma source term will predominate and the N-gamma and hardware activation can be neglected.

10. The high range detectors read out in R/h, which is a measure of the ionization caused by radiation. The MCNP output is provided in mrem/h which is a measurement of the equivalent dose that represents the biological effects of ionizing radiation. The relationship between roentgens and milirem is not straightforward and depends on different absorption of particles in a medium . It is assumed that 1 R is approximately 1000 mrem. This is acceptable as only the gamma source term is considered .

11 . The volume of stainless steel for the upper internals is based on calculation N 015 which performed a detailed evaluation of the water displacement of the upper internals to one foot below the reactor vessel flange. This calculation has been superseded by another calculation that does not provide detailed accounting of the volumes. However, there is no indication that the volumes have changed .

Therefore , these volumes are acceptable for approximating the volume of stainless steel in the area between the active fuel and one foot below the reactor vessel flange, and can be used to determine best estimate dose rates at the detector locations.

Page 9 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001 F..:d ENE RCON Reactor Vessel Water Levels Excellence-Every project. Every day. During Refueling for EAL REV. 0 Thresholds

5. DESIGN INPUTS 5.1 Fuel Assembly Parameters The following fuel assembly parameters are used to homogenize the core in the MCNP model.

They are based on typical fuel assembly values for Westinghouse 14x14 fuel.

Table 5-1 Design Input Fuel Assembly Parameters for Westinghouse Fuel Parameters Value Unit Reference

Fuel Rods per Assy 179 [13]

Assembly Array 14x14 [13]

Assembly Width 7.761 [in] [13]

Density (% of theoretical) 96 [13]

Fuel Pellet OD 0.3659 [in] [13]

Fuel Rod OD 0.422 [in] [13]

Clad Thickness 0.0243 [in] [13]

Active Length 143.25 [in] [13]

5.2 Containment Dimensions The following elevations and dimensions are based on drawings of the Unit 1 containment building and equipment. Some parameters are estimated using drawing scales when exact dimensions are not provided.

Note: All elevations listed in centimeters are relative to the bottom of the active fuel elevation of 18.34' (559 em) [2 and 12]. The x plane is in the direction of the Refueling Canal.

Table 5-2 Design Input Containment Elevations Dimension: ft. in em reference Elevation at Top of Concrete Around 40 Reactor Pressure Vessel (RPV) 2 1/8 665 .59 [8]

66' Elevation 66 0 1452.68 [8]

76' Elevation 76 0 1757.48 [8]

Containment Upper Elevation 156 9 4218.74 [9]

Detector Elevations 70 6 1589.20 [11]

Page 10 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.::d ENERCON Reactor Vessel Water Levels Excellence- Ewry project. Every day. During Refueling for EAL REV. 0 Thresholds Table 5-3 Design Input Containment Dimensions Dimension: ft. in em reference RPV Closure Head Thickness 0 5.5 13.97 [1 0]

RPV Outer Diameter 9 9 297 .18 [7]

RPV Thickness 0 6.656 16.91 [2]

Concrete around RPV (x plane) 17 3 525 .78 [7]

Concrete around RPV (y plane) (Note 4 ft concrete modeled as "D ring) 11 9 358 .14 [7]

Steam Generator (SG)/ Reactor Coolant Pump (RCP) D Rings East D Ring Outer Length (x plane) 41 0 1249.68 [7]

East D Ring Width Outer (y plane)

(Includes 4 ft thickness of concrete around RPV) 26 11 820.42 [7]

East D Ring Inner Length (x plane) 35 0 1066.8 [7]

East DRing Inner Width (y plane) 19 11 607.06 [7]

West D Ring Outer Length (x plane) 43 11 1338.58 [7]

West D Ring Width Outer (y plane) 26 11 820.42 [7]

West D Ring Inner Length (x plane) 37 11 1155.70 [7]

West D Ring Inner Width (y plane) 20 0 609.6 [7]

Overall Modeled Height 66 0 1452.68 [8]

East SG compartment Outer Length (x

plane) 24 10 756 .92 [8]

East SG compartment Outer Width (y plane) 25 5 774.70 [8]

East SG compartment Inner Length (x plane) 19 10 604 .52 [8]

East SG compartment Outer Width (y plane) 20 5 622 .30 [8]

West SG compartment Outer Length (x plane) 24 10 756.92 [8]

Page 11 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.':el E N E R C 0 N Reactor Vessel Water Levels Excelfence- Evt!ry wojecl. Eve1y day. During Refueling for EAL REV. 0 Thresholds Dimension: ft. in em reference West SG compartment Outer Width (y plane) 25 5 774 .70 [8]

West SG compartment Inner Length (x plane) 19 10 604.52 [8]

West SG compartment Outer Width (y plane) 20 5 622 .30 [8]

Active Fuel Active Fuel Height 143.25 363.86 [13]

Diameter of Active Fuel 96.5 245 .11 [13]

South Concrete Wall Thickness (x direction) 3 0 91.44 [8]

Length (y direction) 23 6 716 .28 [8]

Containment Midpoint X 178 [8] Scaled Midpoint Y 68 [8] Scaled Inner Radius 52 3 1592.58 [7]

Liner Thickness 0 0.25 0.64 [15]

Assumption Concrete Thickness 3 0 91.44 4 Detector Locations Detector RE-126 Distance from RPV (x-plane) 1595.32 [11] Scaled Detector RE-126 Distance from RPV (y- [11] Scaled plane) -633 .04 Detector RE-127 Distance from RPV (x- [11] Scaled plane) -1102.16 Detector RE-127 Distance from RPV (y- [11] Scaled plane) 982.4 Detector RE-128 Distance from RPV (x- [11] Scaled plane) 482 .8 Detector RE-128 Distance from RPV (y- [11] Scaled plane) -1486.48 Page 12 of 37

Dose Rate Evaluation of CALC NO. NEE-363 -CALC -001 F::d ENERCON Reactor Vessel Water Levels Excellence- Every project. Every day. During Refueling for EAL REV. 0 Thresholds 5.3 Core Isotopic Inventory Core isotopic activities are taken from Reference [14]. A table of the input values is shown in Table 5-4, below.

Table 5-4 Core Shutdown Source Term Isotope Ci Isotope Ci Kr-85 6.15E+05 Sr-92 6.73E+07 Kr-85m 1.36E+07 Ba-139 9.42E+07 Kr-87 2.68E+07 Ba-140 9.05E+07 Kr-88 3.60E+07 Ru-1 03 7.79E+07 Xe-131 m 5.55E+05 Ru-105 5.42E+07 Xe-133 1.02E+08 Ru-1 06 2.54E+07 Xe-133m 3.21 E+06 Rh-1 05 5.08E+07 Xe-135 2.17E+07 Te-99m 8.47E+07 Xe-135m 2.20E+07 Mo-99 9.62E+07 Xe-138 9.05E+07 Ce-141 8.52E+07 1-130 1.05E+06 Ce-143 8.03E+07 1-131 5.1 OE+07 Ce-144 6.72E+07 1-132 7.47E+07 Pu-238 1.33E+05 1-133 1.06E+08 Pu-239 1.45E+04 1-134 1.19E+08 Pu-240 2.25E+04 1-135 1.01 E+08 Pu-241 5.73E+06 Rb-86 9.95E+04 Np-239 9.65E+08 Cs-134 9.52E+06 Y-90 5.01 E+06 Cs-136 2.14E+06 Y-91 6.56E+07 Cs-137 6.27E+06 Y-92 6.82E+07 Cs-138 9.89E+07 Y-93 7.67E+07 Te-127 4.54E+06 Nb-95 8.87E+07 Te-127m 7.48E+05 Zr-95 8.76E+07 Te-129 1.33E+07 Zr-97 8.80E+07 Te-129m 2.52E+06 La-140 9.69E+07 Te-131 m 9.95E+06 La-142 8.25E+07 Te-132 7.30E+07 Pr-143 7.75E+07 Sb-127 4.63E+06 Nd-147 3.33E+07 Sb-129 1.42E+07 Am-241 6.16E+03 Sr-89 5.03E+07 Cm-242 1.70E+06 Sr-90 4 .80E+06 Cm-244 1.58E+05 Sr-91 6.30E+07 Page 13 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001 F.:JJ ENERCON Reactor Vessel Water Levels Excellence - Every project. Every day. During Refueling for EAL REV. 0 Thresholds 5.4 Material Compositions The following compositions used in the MCNP model are taken or developed from the SCALE standard composition library [1] and are shown in Table 5-5.

Table 5-5 SCALE Standard Compositions used in MCNP Model Material Isotope Weight Fraction Reference Zry- 4 Zr 0.9823 [1]

(6.56 g/cm 3 ) Sn 0.0145 Cr 0.0010 Fe 0.0021 Hf 0.0001 UOl U-235 0.0348 (10.5216 g/cm 3 ) U-238 0.8466 0 0.1186 Air c 0.0001 [1]

(1.21 E-03 g/cm 3 ) N 0.7651 0 0.2348 Water H 0.1111 [1]

(0.9982 g/cm 3 ) 0 0.8889 SS-304 Fe 0.6838 [1]

(7 .94 g/cm 3 ) Cr 0.1900 Ni 0.0950 Mn 0.0200 Si 0.0100 c 0.0008 p 0.0004 Concrete 0 0.5320 [1]

(2.30 g/cm 3 ) Si 0.3370 Ca 0.0440 AI 0.0340 Na 0.0290 Fe 0.0140 H 0.0100 Carbon Steel c 0.0100 [1]

(7 .82 g/cm 3 ) Fe 0.9900 2

Based on 96% of theoretical density Page 14 of 37

Dose Rate Evaluation of CALC NO. NEE-363 -CALC-001 F.r:d ENERCON Reactor Vessel Water Levels Excellence- Ewry project. Every doy. During Refueling for EAL REV. 0 Thresholds 5.5 Upper Internals/Upper Fuel Hardware The following are used in the MCNP model for the Upper Internals/Upper Fuel Hardware Region

[2] :

The volume of stainless steel for the Upper Core Plate is 4.797 ft 3 .

The volume of stainless steel for the Upper Internals region is 62 .856 ft 3 .

The volume of stainless steel for the Upper Support Plate is 5.2 ft 3 .

5.6 Containment High Range Detector Range The range of RE-126, RE-127 and RE-128 is 10°-108 R/h [16] .

5.7 ANSI/ANS - 1977 Flux to Dose Factors Flux to dose conversion factors are taken from ANSI/ANS-6.1.1-1977 [6] and are shown in Table 5-6.

Table 5-6 ANSI/ANS-6.1.1-1977 Flux to Dose Factors MeV mrem/h r/(y/cm 2/s) MeV mrem/hr/(y/cm 2/s) 0.01 3.96E-03 0.8 1.68E-03 0.03 5.82E-04 1 1.98E-03 0.05 2.90E-04 2.2 3.42E-03 0.07 2.58E-04 2.6 3.82E-03 0.1 2.83E-04 2.8 4.01 E-03 0.15 3.79E-04 3.25 4.41 E-03 0.2 5.01 E-04 3.75 4.83E-03 0.25 6.31 E-04 4.25 5.23E-03 0.3 7.59E-04 4.75 5.60E-03 0.35 8.78E-04 5 5.80E-03 0.4 9.85E-04 5.25 6.01 E-03 0.45 1.08E-03 5.75 6.37E-03 0.5 1.17E-03 6.25 6.74E-03 0.55 1.27E-03 6.75 7.11E-03 0.6 1.36E-03 7.5 7.66E-03 0.65 1.44E-03 9 8.77E-03 0.7 1.52E-03 11 1.03E-02 Page 15 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.~ ENERCON Reactor Vessel Water Levels Excellence-Ev~ry ptoject. Every doy. During Refueling for EAL REV. 0 Thresholds

6. METHODOLOGY The reactor source terms are computed with OR IGEN-S of the SCALE 6.1 code package [3] . The ORIGEN-S decay sequence is used to bin design input isotope specific activities into energy dependent photon bins. These energy specific photon emission bins are used as input for the energy distribution described by the MCNP source definitions.

The MCNP6 [5] Monte Carlo transport code is used to determine the dose rates via the flux to dose conversion factors in Table 5-6 .

The detailed engineering drawings are converted into MCNP surface and cell cards in the dimensions shown in Table 5-2 and Table 5-3 . The radiation monitors of interest are modeled as point detectors to determine the expected dose rate for those detectors. The dose rates are calculated as a function of water height for two reactor refueling conditions:

1. With Head- the reactor is modeled with a 5.5 inch carbon steel plate as indicated in Table 5-3, which is additional attenuation between the source and detector. The mass of the Upper Internals and Upper Fuel Hardware including Upper Core Plate , Upper Support Assembly, and Upper Support Plate is homogenized between the active fuel region and the vessel head .

2. Without head - the reactor is modeled with air between the active fuel zone and containment.

Variance reduction is accomplished with a geometric importance map that is imposed on the homogenized core . In addition, cell based importance weighting and source biasing (see Section 7.5) are utilized to improve the variance reduction of the simple geometric scheme. A superimposed weight window mesh is utilized where necessary to improve variance . The weight windows are iteratively generated using the MCNP weight windows generator card . All final dose rates presented in this calculation include weight windows variance reduction.

Page 16 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.:il ENERCON Reactor Vessel Water Levels E.xcellence-[\'UY project. Every day. During Refueling for EAL REV. 0 Thresholds

7. CALCULATIONS 7.1 Source Terms The ORIGEN-S input deck, PBEALa .inp , is provided in Appendix C. This input produces a simple case where the isotopic composition from Table 5-4 is decayed . The isotope is specified in the 73$$ card using the special identifier described in Section F7 .6.2 of the ORIGEN-S manual , and the activity in curies is specified in the 74** card . The time steps for the decay are given on the 60** card in hours. Although multiple time steps are calculated, the source term with 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months decay time is used in this calculation to model the core shortly after shutdown. The output of the decay is given in terms of photons/s/Energy-Group , which is automatically normalized in the MCNP input.

Page 17 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F:':d ENERCON Reactor Vessel Water Levels Excellence-Every project. Every doy. During Refueling for EAL REV. 0 Thresholds The results of this calculation are summarized below in Table 7-1 . These values are used in the MCNP input source definition .

Table 7-1 Binned Total Core Source Term Energy Energy Boundaries Photons/sec Group (MeV) 1 0.01-0.05 1.891 E+19 2 0.05-0 .1 6.096E+18 3 0.1-0.2 1.419E+19 4 0.2-0 .3 8.760E+18 5 0.3-0.4 3.436E+18 6 0.4-0 .6 7.251E+18 7 0.6-0 .8 1.334E+19 8 0.8-1 2.118E+18 9 1-1.33 4.634E+17 10 1.33-1 .66 3.617E+18 11 1.66-2 6.647E+16 12 2-2 .5 7.526E+16 13 2.5-3 1.122E+17 14 3-4 8.752E+14 15 4-5 1.781E+10 16 5-6.5 5.314E+08 17 6.5-8 9.046E+07 18 8-10 1.920E+07 19 10-11 1.040 E+06 totals 7.844 E+19 Page 18 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F..~ E N E RC 0 N Reactor Vessel Water Levels Excelfence- E\'ery project. Every day. During Refueling for EAL REV. 0 Thresholds 7.2 MCNP Model Core Homogenization Because the source term is given for the entire core , the self-shielding from the assemblies is an important part of the dose rate response. Particles born in the lower section of the core are very unlikely to penetrate through the core itself, and make it to the radiation monitors. For simplicity, the core is modeled as a three dimensional cylinder with a uniformly distributed spatial particle distribution. The calculations for determining the mass of fuel, cladding and water for the core and the resulting density is shown below. The inputs are based on the dimensions in Table 5-1 .

Rod Volume= n(Pellet Radius) 2 x Active Length= rr(0.18295 in) 2 (143.25 in)

= 15.06 in 3 (15.06 in 3 ) (2.54 ~m) = 2596.6 g 3

Rod Massuo2 = p xV = (10.5216i!..) cc rn Number of Fuel Rods Assembly Massuo 2 =Rod Mass x A bl ssem y

= (2596.6 g)(179)

= 464.79 kg Clad Volume = rr ( 4aDz - 4ID 2) x Active Length (0.422 in) 2 (0.3734 in) 2 ]

= (rr) [ 4 4

(143.25 in) = 4.35 in 3 g em 3 Rod MaSSzry- 4 = p XV= ( 6.56 cJ ( 4.35 in 3 ) ( 2.54-;;:-) = 467.6 g Number of Fuel Rods Assembly MaSSzry- 4 =Rod Mass x A ssem y bl = (467.6g)(179) = 83.70 kg Assembly H 2 0 Volume

= [(Assembly Width) 2 -n(Rod Radius) 2 x Number of Fuel Rods]

x Active Length= [(7.761 in) 2 - (rr)(0.211 in) 2 (179)](143.25 in)

= 5042 in 3 g em 3 Assembly MassH2 0 = p x V = ( 0.9982 -) cc (5042 in 3 ) (2.54-.-) = 82.47 kg rn Page 19 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001

..:d ENERCON Reactor Vessel Water Levels bceller.ce-Evay project. Evety day. During Refueling for EAL REV. 0 Thresholds Assembly Volume= Activ e Length x (Ass embly Width) 2 = (143.25 in)(7.761 in) 2

= 8628.4 in 3

. Total Mass 1000gjkg(464.79 + 83.7 + 82.47) kg Density = = 3

= 4.46 g / cc Volume 8628.4 in 3 ( 2.54 ~7:)

The corresponding isotopic composition for the homogenized active fuel region is calculated based on the compositions in Table 5-5. An example calculation for the mass fraction of U-235 is included below.

Assembly Mas suo 2 Mass Fraction U235 = l Tota Mass x weight fraction U2 35 464.79 kg 256 (464.79 + 83.7 + 82.47) kg X 0.0 348 = 0.0 The remaining calculations for the homogenization are done in the worksheet Compositions of the EXCEL workbook PB.xlsx and are shown in Appendix B. The isotopic compositions are calculated with the water level above the top of the fuel. Note that the EXCEL workbook uses add itional significant figures .

Table 7-2 Homogenization of Active Fuel Region ZAID Atom Mass Fraction Number Active Fuel Region Homogenized 92235 U-235 0.0256 92238 U-238 0.6236 8016 0 0.2035 40000 Zr 0.1303 50000 Sn 0.0019 24000 Cr 0.0001 26000 Fe 0.0003 72000 Hf 0.0000 1001 H 0.0145 7.3 MCNP Model Upper Internals Homogenization For the case with the RPV head in place, the Upper Internals and Upper Fuel Hardware region are modeled as a discrete cylinder with a uniformly distributed homogenized material to account for the mass of stainless steel between the active fuel height and RPV head . The homogenization accounts for the mass of metal from Section 5.5 (assumed stainless steel type 304 per Assumption 7) distributed evenly across the volume between the active fuel height ant the head.

Page 20 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.,:d ENERCON Reactor Vessel Water Levels Excellenc~-[\'r!ry pro}t!ct. Every day. During Refueling for EAL REV. 0 Thresholds The total mass of the stainless steel is calculated by multiplying the volume of stainless steel (Design Input 5.5) by the regular density of stainless steel (Table 5-5).

Mass Upper Internals =p xV 3 3 g ( in) em

= ( 7.94 cJ ( 4.797 + 62.856 + 5.2 ft 3 ) 12ft ( 2.54in")

= 1.638 X 10 7 g The mass is divided by the volume of the region between the active fuel height and the RPV head to determine the density.

Density Upper Internals= Mass Upper Internals-;- V

= 1.638 x 10 7 g-;- (301.73cm x (n(280.27cm) 2 ) = 0.22 ~

cc 7.4 MCNP Model Geometry The following MCNP model geometry is based on the containment dimensions summarized in Table 5-2 and Table 5-3. The model only focuses on the primary systems and components that provide shielding or reflection from the core to the radiation monitors. These components include the reactor vessel, concrete in reactor pit, containment walls and floor slabs. VISED plots of the model geometry are provided in Figure 7-1, Figure 7-2, and Figure 7-3 . The MCNP surface cards with the model dimensions (em) are shown in Figure 7.4 and the cell cards are shown in Figure 7-5 for the cases with no reactor head . A VISED plot of the model with the reactor head is shown in Figure 7-6 . Areas that are not of interest are given an importance of zero (white areas) so MCNP will not track particles in locations that will not contribute to the detector response.

Page 21 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001 F.:d ENE R C 0 N Reactor Vessel Water Levels Excelfence-Evf!ry project. Every day. During Refueling for EAL REV. 0 Thresholds Figure 7-1 Y-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (No Head)

Air Concrete Reactor Pit

~~-1------ Reactor Vessel 665.59 em omogenized Core 363 .86 em Page 22 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.JJ ENERCON Reactor Vessel Water Levels Excelfence-Ewry p toject. Every day. During Refueling for EAL REV. 0 Thresholds Figure 7-2 Y-Z VIS ED Plot of Containment3 1592.58 em Radiation Radiation Monitor Monitor

+

1704.34 em (66' Level) 0/ ~0 121.92 e

..._ ~ 91.44 em 1856.74 em (46' Level) 81.08 em (21 ' Level) I I 0 em (12' 1" Level) 3 Radiation monitors are not on the same Y-Z plane as the center of the core shown above. They are included for visualization purposes only.

Page 23 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 1\::d ENERCON Reactor Vessel Water Levels Excellence-E\'ery project. Evety day. During Refueling for EAL REV. 0 Thresholds Figure 7-3 Y-X VISED Plot of the Containment Geometry at 66' Elevation Level 4 RE-127 60706 omI 716 .28 em + "D" Rings 1155.7cm RE-126 RE-128 4 Note that radiation monitors are actually located 4'6" above the 66' elevation, and are included for visualization purposes only.

Page 24 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F..~ E N ERC 0 N Reactor Vessel Water Levels Ex(e/lence - E~'ery ptoject. Every day. During Refueling for EAL REV. 0 Thresholds Figure 7-4 MCNP Model Surface Cards 5 c surfaces 1 rcc 0 0 0 0 0 363.86 122.555 $ Active Fuel Region 2 rcc 0 0 0 0 0 665.59 280.27 $ Reactor Pressure Vessel Inner Surface 3 rcc 0 0 0 0 0 665 . 59 297 . 18 $ Reactor Pressure Vessel Outer Surface 4 rpp -525.78 525.78 - 358 . 14 358 .1 4 0 665.59 $ Concrete Surrounding RPV 5 pz 0 6 pz 665.59 $ Elevation at Head Level 7 rpp - 436.88 812.8 358.14 1178 . 56 0 1452.68 $ East D Ring Outer Surface 8 rpp -345 . 4 4 721 . 36 480.06 1087.12 0 1452 .68 $ East D Ring Inner Surface 9 rpp - 812 . 8 525.78 -1178.56 -358 . 14 0 1452 . 68 $ West D Ring Outer Surface 91 rpp -721.36 434 . 34 -1087.12 - 480 . 06 0 1452 . 68 $ West D Ring Inner Surface 10 pz 1452.68 $66 Elevation Top of Slab 11 pz 1437.44 $66 Elevation Bottom of Slab 12 rpp 55.08 812.8 403.86 1178.56 1452 . 68 1757.48 $ East SG Compartment Outer Surface 13 rpp 131.28 736 . 6 480 . 06 1102.36 1452 .68 1757.48 $ East SG Compartment Inner Surface 14 rpp - 812.8 - 55 . 08 -11 78 . 56 -403.86 1452 . 68 1757.48 $ West SG Compartment Outer Surface 15 rpp -736.6 -131.28 -1102.36 - 480.06 1452.68 1757 . 48 $ West SG Compartment Inner Surface 16 pz 1757 .48 $ Top of SG Compartments 17 rpp 434.34 525 . 78 -358.14 358.14 665.59 1452.68 $ South Concrete Wall 18 pz 81.08 $ 21 Elevation Top of Slab 19 pz 65 . 84 $21 Elevation Bottom of Slab 20 pz 363.86 $ Water Elevation Surface 21 rcc 178 68 0 0 0 4126.66 1592.58 $ Containment Inner Liner Surface 22 rcc 178 68 0 0 0 4127 . 3 1593 . 22 $ Conta inment Inner Concrete Surface 2 3 rcc 178 68 0 0 0 4218.74 1684.66 $ Containment Outer Concrete Surface 24 pz 843.08 $ 46 Elevation Top of Slab 25 pz 827.84 $46 Elevation Bottom of Slab 26 py 358.14 $East Sl ab Wa ll 27 py -358 . 14 $West Slab Wall 28 rcc 0 0 665.59 0 0 16.91 297.18 $ Reactor Head 101 pz 36.386 1 02 pz 72.772 103 pz 109 . 158 104 pz 14 5.544 105 pz 181. 93 106 pz 218 . 316 1 07 pz 254.702 108 pz 291.088 109 pz 327.474 110 pz 363.86 5 The surface cards for the MCNP models without the reactor vessel head do not have surfaces 28. Surface 20 is variable for the cases without the reactor vessel head only.

Page 25 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.J~ ENERCON Reactor Vessel Water Levels Excellence-E\'?ry project Every day. During Refueling for EAL REV. 0 Thresholds Figure 7-5 MCNP Model Cell Cards (No Head) c cells 101 1 -4.46 101 5 imp : p=1 $ Active Fuel Regi on 102 1 - 4.46 - 1 101 - 102 imp : p=2 $ Active Fuel Region 103 1 -4.46 - 1 102 -1 03 imp:p=3 $ Act i ve Fuel Region 1 04 1 -4.46 -1 103 - 104 imp : p = 4 $ Active Fuel Region 105 1 -4.46 -1 104 - 105 imp:p=8 $ Active Fuel Region 106 1 - 4 . 46 - 1 105 -106 imp:p=16 $ Active Fuel Region 107 1 -4.46 - 1 1 06 -107 imp:p= 32 $ Active Fuel Region 108 1 - 4.46 - 1 107 - 108 imp:p=64 $ Act i ve Fuel Region 109 1 - 4 . 46 -1 108 - 109 imp:p=128 $ Active Fuel Region 110 1 - 4.46 -1 109 -110 imp:p=256 $ Active Fuel Region 2 2 - 0 . 9982 1 - 2 5 -20 imp:p= 256 $ Water Region 3 4 - 0 . 22 20 -2 imp:p=256 $ Homogenized Steel 4 4 -7.94 2 -3 5 - 6 imp : p=256 $ RPV Shell 7 5 - 2.3 3 -4 imp : p=256 $ Concrete Surrounding RPV 8 5 - 2.3 - 7 8 imp:p=256 $ East D Ring Concrete 9 5 -2.3 - 9 91 i mp:p =256 $ West D Ring Concrete 10 5 -2.3 - 12 13 imp : p=256 $ East SG Compartment Concrete 11 3 -1. 21E-03 -13 i mp:p=256 $ East SG Compartment Air 12 5 - 2.3 -1 4 15 imp:p=256 $ West SG Compartme n t Concrete 13 3 - 1. 21E 15 imp:p=256 $ West SG Compartment Air 14 5 - 2.3 - 17 #9 #7 imp:p=256 $ South Concrete 15 5 -2 . 3 - 1 8 19 -21 26 7 imp:p=256 $ 21 Elevation East Slab 1 6 5 -2.3 -18 19 27 9 imp:p=256 $ 21 Elevation West Slab 17 5 -2.3 -24 25 -21 26 7 imp:p=256 $ 46 El evation East Slab 18 5 -2 . 3 -24 25 27 9 imp : p =256 $ 46 Elevation West Slab 19 5 -2.3 -10 11 - 21 26 7 imp : p =256 $ 66 Elevation East Slab 20 5 - 2.3 -10 11 27 9 imp:p=2 56 $ 66 Elevation West Slab 21 4 - 7 . 94 21 - 22 imp:p=256 $ Containment Liner 22 5 - 2 . 3 22 - 23 imp:p=256 $ Containment Wall 23 6 -7.82 - 28 30 3 -1 . 21E 21 4 #7 #8 #9 #10 #11 #12 #13 #14 #15

16 #17 #18 #19 #20 #23 imp:p=256 999 0 1 23 imp:p=O $ Problem Boundary Page 26 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC -001 F.::d E N E R C 0 N Reactor Vessel Water Levels Ext;ellenc~-E\'ery projea. Every day. During Refueling for EAL REV. 0 Thresholds Figure 7-6 X-Z VISED Plot of Reactor Vessel and Concrete Reactor Pit (With Head) 13.97 em (5.5")

Reactor Cell for the Head homogeniza tion of the Upper Internals stainless steel 0.22 g/cm 3 Homogenized Core Page 27 of 37

Dose Rate Evaluation of CALC NO. NEE-363 -CALC-001 ft:d ENERCON Reactor Vessel Water Levels Excellence-Every project. Every day. During Refueling for EAL REV. 0 Thresholds 7.5 MCNP Source Definition The core source term is modeled as uniformly distributed throughout the homogenized core, and has an energy spectra based on the core inventory [14]. Only the gamma source term is taken into account for this evaluation . The source term is generated shortly after shutdown, therefore, the fuel gamma source term will predominate , and the N-gamma and hardware activation source terms can be neglected (Assumption 5). The source is defined on the MCNP sdef card using distributions to define the particle location and energy. The radius of the core is defined with the rad parameter, which automatically creates a uniform distribution based on a cylindrical geometry.

The ext and axs parameters define the direction and distance of the cylinder axis . These parameters combined define the core where the particles can be born. The erg parameter defines the energy spectrum of source particles and is based on the results of the ORIGEN-S calculation discussed previously . This distribution is a histogram of energies represented by activities. These are automatically normalized by MCNP to create a probability distribution. The total activity is preserved in the tally multiplier. The MCNP source definition cards are shown below in Figure 7-7.

The sb card is a source biasing card, which in this case biases the particle generation to the upper end of the core . This is a variance reduction technique to improve the statistical certainty in the results .

Figure 7-7 MCNP Source Definition Cards sdef rad=dl ext=d2 axs =O 0 1 e rg=d8 ~ Source Defi n ition Card

-Radius = d l

-Extent = d2

- Axis = +Z

-Energy = d8 sil 122.555 ~ Core Radius Distribution s i 2 h 0 36 . 386 72.772 1 09 .1 58 14 5 . 544 18 1. 93 218 . 316 2 54 . 702 ~ Core Axi a l Distr i but i on 291.088 327 . 474 363.86 sp2 0 1 1 1 1 1 1 1 1 1 1 ~ Actual Uniform Distribution sb2 0 0 . 001 0 . 00 1 0.0 1 0 . 01 0.01 0.1 0.1 0. 1 1 1 ~Bi ased to Top Di st ribu ti o n c Fuel Gamma Spectra siS h l.OOOe - 002 5 . 000e - 002 l.OOOe - 001 2 . 000e - 001 3 . 000e - 00 1 4 . 000e - 00 1 ~ Sou r ce Ene rgy Group s 6.000e - 001 S.OOOe - 001 l.OOOe+OOO 1 . 330e+OOO 1.660e+OOO 2.000e+OOO 2 . 500e+OOO 3 . 000e+OOO 4. 000e+OOO 5 . 000e+OOO 6 . 500e+OOO 8 . 000e +OOO l.OOOe +OOl l . lOOe +OO l sp8 O.OOE +OO 1 .89 1E+ l 9 6.096 E+l8 1 . 419E+ l 9 8 . 760E+l8 3 .4 36E+l8 7 . 251E+l8 ~ Sou r ce Emiss i on on Energy Basis 1.334E+l9 2.118E+l8 4.634E+ l7 3 . 61 7E+ l 8 6.647E+ l 6 7 .5 26E+ l 6 1 . 122E+l7 8 . 752E+l4 1 . 78 1 E+l0 5.314E+08 9.046E+07 1.920E+07 1 . 040E +06 Page 28 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 p.:*r ENERCON Reactor Vessel Water Levels Excellence- Every p10jecr. Every day. During Refueling for EAL REV. 0 Thresholds 7.6 MCNP Tally Specification The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-126 , RE-127 and RE-128 for Unit 1. Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions . The inputs to this card are the coordinates of the dose points followed by an exclusion zone (reduce variance) , as well as a multiplier card, which represents the total core activity in photons/sec. The tally cards are shown in Figure 7-8 .

Figure 7-8 MCNP Tally Cards f5c RE - 126, 127 and 128 ~ Tally Comment Card f5:p 1595.32 -633.04 1589.84 20 ~ Tally 5 (point detector)

-1102 . 16 982.4 1589.84 2 0 x y z exclusion 482 . 8 -1486 . 48 1589.84 20 fm5 7 . 844E+l9 ~ Ta l ly Mul t ip l ier (Total Activity)

In addition , the flux is multiplied by ANSI/ANS flux-dose conversion factors [6]. This is specified in MCNP using the de/df cards. These are shown in Figure 7-9.

Figure 7-9 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors c - -------- ------------------- ---------- ------------------ ----------

c ANSI/ANS-6.1.1-1977 c Gamma Flux to Dose Conversion Factors c (mrem/hr)/(photons/cm2 - s) c -- ---------- --------- - -------- - - ------ ----------------------------

deO .01 .03 .05 . 07 .10 .15 .20 .25 .30 .35 .40 ~Energy Bins for Fl ux

. 45 .50 .55 .60 . 65 .70 .80 1 . 1.4 1.8 2.2 to Dose Conversion 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5 . 75 6.25 6.75 7.5 9. 11.

dfO 3 . 96E - 03 5.82E-04 2 .90E - 04 2.58E-04 2 .83E-04 3.79E- 04 ~Energy Dependent 5.01E- 04 6.31E- 04 7.59E - 04 8.78E-04 9.85E-04 1.08E - 03 Flux Multipliers l.l7E-03 l. 27E- 03 1 . 36E-03 1 . 44E-03 1.52E- 03 1 . 68E-03

l. 98E- 03 2.51E-03 2.99E - 03 3.42E-03 3.82E- 03 4.01E - 03 4 . 41E-03 4 . 83E - 03 5 . 23E-03 5.60E-03 5 . 80E-03 6.0 1E-03 6 . 37E- 03 6. 7 4E-03 7 . llE-03 7.66E-03 8.77E-03 1.03E- 02 Page 29 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 lt d ENERCON Reactor Vessel Water Levels Exce lfence- Ewry p roject. Every day. During Refueling for EAL REV. 0 Thresholds 7.7 MCNP Material Cards The MCNP material cards are provided in Figure 7-10. These are based on the compositions described in Table 5-5 or calculated in Section 7.2.

Figure 7-10 MCNP Material Cards rn l 92235 - 0 . 0 2 56 $ Homogeniz e d Ac tiv e Fu e l Region 92238 - 0.623 6 8016 -0. 2 035 40000 -0.1303 50000 -0.0019 24000 -0.0001 26000 - 0.0003 1001 -0.0145 rn2 1001 2 8016 1 $ Water rn3 6012 - 0.000 12 6 $ Air 70 1 4 -0. 7 6508 8016 -0 . 23479 3 rn4 6000 -0.0008 $ ss 304 14000 -0.01 15031 - 0.00045 24000 -0 . 19 25055 -0.02 26000 -0 . 68375 28000 - 0.09 5 rn5 26000 -0 . 014 $ Reg-Concrete 1001 -0 . 01 13027 - 0. 034 20000 -0.044 8016 - 0.53 2 14000 -0 . 337 11023 -0.0 2 9 rn6 6012 -0.01 $ Carbon Ste e l 26056 -0 . 99 Page 30 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.::d ENERCON Reactor Vessel Water Levels b:cellence-fvery project. Every day. During Refueling for EAL REV. 0 Thresholds 7.8 Results File Naming Scheme:

The MCNP input files are named with the following convention:

P-height-condition-iteration where :

P = Project (PB)

Condition =h - with head n- no head Height= water height from top of active fuel region (ft)

Iteration (used for weight window optimization) =a-z Page 31 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001

,j J ENERCON Reactor Vessel Water Levels Excelfence -E~*~ry project. Every day. During Refueling for EAL REV. 0 Thresholds 7.8.1 Results without Head The dose rate as a function of water level is provided in Table 7-3 and the smallest of the three dose rates is plotted in Figure 7-11 , below. All of the water levels described in the following sections refer to the level at the top of the fuel (i .e. 0 foot water level is at the top of the fuel assemblies) . The water level is less than four feet above the top of active fuel before a detectable response of 1 R/h (1 E+03 mrem/h) is seen at the detectors. Note that the uncertainty fo r some of the dose rates for 6 feet above active fuel is higher than 10%, which is indicative of the difficulty in converging a thick-shielded problem such as this . The slope of the curve is nearly linear which is the expected response for attenuation through water, therefore, the results are judged to be accurate.

Table 7-3 Dose Rate Response as a Function of Water Level for no Head Configuration (mrem/h)

Water Dose Rate 1 Dose Rate 2 Dose Rate 3 Level (ft) RE-126 fsd 6 RE -127 fsd RE-128 fsd Tally File 0 1.09E+05 3.99% 1.79E+05 0.79% 1.26E+05 5.57% pbnOem 2 4.75E+03 4.12% 7.74E+03 0.86% 5.55E+03 5.44% pbn2em 4 1.55E+02 6.73% 2.62E+02 3.36% 1.63E+02 6.76% pbn4hm 6 4.94E+OO 10.94% 9.15E+OO 11.78% 5.66E+OO 11 .81% pbn6im Figure 7-11 Dose Rate versus Water Height Plot for no Head Configuration l .OOE+06 - -- - ----

Ll 5

E

~

E l .OO E+OS

l. OO E+04

-; l.OOE+03 1**

. ~ .~

- - - - ... I...

-l- 1-co cr::

QJ **** ~

3 l .OOE+02 - ~

a ....

1. .

l. OOE+Ol - - - *************-+ --

l. OOE+OO I I 0 1 2 3 4 5 6 7 Water Level from Top of Fu el (ft) 6 Fraction standard deviation.

Page 32 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.:d ENERCON Reactor Vessel Water Levels Excellence-Ewry project. Every day. During Refueling for EAL REV. 0 Thresholds 7.8.2 Results with Head The dose rate results for the case with the head in place and water level at the active fuel height are lower due to the increased shielding. The dose rates are listed in Table 7-4 . Note that dose rates are below the detectable range of the radiation monitors of 1 R/h (1 E+03 mrem/h).

Table 7-4 Dose Rate Response for Head in Place Configuration (mrem/h)

Water Dose Rate 1 Dose Rate 2 Dose Rate 3 Tally Level (ft) RE-126 fsd 7 RE-127 fsd RE-128 fsd File 0 2.94E+01 9.38% 4.87E+01 3.06% 3.19E+01 6.68% pbhOdm

8. COMPUTER SOFTWARE This calculation uses ORIGEN-S of the SCALE Version 6.1.2 code package [3] and MCNP Version 6.1.0 [5] in accordance with CSP 3.09.

7 Fraction standard deviation.

Page 33 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F.:cl ENERCON Reactor Vessel Water Levels Ex,ellence - Ewry project. Every day. During Refueling for EAL REV. 0 Thresholds APPENDIX A -ELECTRON IC FILE LISTING Origen output:

02/14/2017 09:47AM 84 ,321 PBEALa.out MCNP output :

Directory of \No head\0 feet 04/04/20 17 05:44 PM 1,918,056 pbnOeo Directory of \No head\2 feet 04/04/20 17 05:58AM 1,955,698 pbn2eo Directory of \No head\4 feet 04/04/20 17 05:07 AM 1 , 802 ,4 17 pbn4ho Directory of \No head\6 feet 04/04/2017 03:19 AM 1,5 69,576 pbn6io Directory of \With Head\

0 4/0 4/2017 06:08 AM 2 ,0 23 , 879 pbhOdo Page 34 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001

'll ENERCON c

Reactor Vessel Water Levels fxcellence -E\'~ry p10ject. Ev~ ry day. During Refueling for EAL REV. 0 Thresholds APPENDIX 8- PB.XLSX SHEETS A 8 c D E G H K i\Iass ZAID Mass Fraction Active Fuel i\Iaterial Isotope Veigbt Fractio Reference l\Iaterial Atom (kg) Number Region Homogenized 2

Zry- 4 Zr 0.9823 [1] uo2 464.79 92235 U-235 0.0256 4 ( 6.56 glcm3) Sn 0.0145 Zry-4 83.7 92238 U-238 0.6236 5 Cr 0.001 Water 82.47 8016 0 0.2035 6 Fe 0.0021 40000 Zr 0.1303 7 Hf 0.0001 50000 Sn 0.0019 8 uo2 U-235 0.0348 [1] 24000 Cr 0.0001 9 10.52161'lcm3) U-238 0.8466 26000 Fe 0.0003 10 0 0.1186 72000 Hf 0.0000 11 Air c 0.0001 [1] 1001 H 0.0145 12 (1.21E-03 g/cnl) N 0.7651 1.0000 13 0 0.2348 14 Water H 0.1111 [1]

r I

15 (0.9982 g/cm3) 0 0.8889 16 SS-304 Fe 0.6838 [1]

17 'r7.941'lcm3) Cr 0.19 18 Ni 0.095 19 Mn 0.02 20 Si O.ot 21 c 0.0008 22 p 0.0004 23 Concrete 0 0.532 [1]

24 (2.30 g/cm3) Si 0.337 25 Ca 0.044 26 A1 0.034 27 Na 0.029 28 Fe 0.01 4 29 H 0.01 30 Carbon Steel c 0.01 [1]

31 (7.82 g/cm3) Fe 0.99 Page 35 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 F,"~

I r ENERCON Reactor Vessel Water Levels Excellence - E~*uy project. Every day. During Refueling for EAL REV. 0 Thresholds A D G H

?t l aterhd Isotope \\'eight Fraction Rererence l\laterial Mass (kg) lAID Number Atom Mass Fraction Active Fuel Region Homogenized Zl)*- 4 Zr 0.9823 (!) uo, 464.79 92235 U-235 =( H3/SUM(H3:H5))"0 8 (6.56 glan3) Sn 0.0145 Zl)*- 4 83.7 92238 U-238 =(H3/SUM(H3 :H5) ) "09 Cr 0.001 'Ya ter 82.47 8016 0 =( ( H3/( SUM( H3:H5))) " 0 10)+{ ( HS/( SUM(H3:H5)))) "DIS Fe 0.0021 40000 Zr =($H$4/SUM($H$3:$H$5))" 0 3 Hf 0.0001 50000 Sn =($H$4/ SUM($H$3:$H$5)) "04 uo, U-235 0.0348 (!) 24000 Cr =($H$4/SUM($H$3:$H$5))" 0 5 10.5216 Rlcm') U-238 0.8466 26000 Fe =($H$4/SUM($H$3:$H$5))"06 10 0 0. 11 86 nooo Hf =($H$4/SUM($H$3 :$H$5))" 07 11 Air c 0.0001 [I) 1001 H =(H5/SUM ( H3:H5)) "014 12 (1.21E-03 glcm3) N 0.7651 I .SUM(l3:U1) 1l 0 0.2348 14 'Vat e r H 0.1 111 [1 )

15 (0.9982 glcm3) 0 0.8889 16 SS-30 4 Fe 0.6838 (!)

17 7.94 elcm3l Cr 0.19 18 Ni 0.095 19 Mn o.oz 2() Si 0.01 21 c 0.0008 22 p 0.0004 23 C oncrete 0 0.532 [1) 24 (2.30 glcm') Si 0.337 25 Ca 0.044 26 A! 0.034 27 Na 0.029 28 Fe 0.014 29 H 0.01 30 Carbon Steel c 0.01 (!)

31 (7.82 glcm3) Fe 0.99 Page 36 of 37

Dose Rate Evaluation of CALC NO. NEE-363-CALC-001

..:cl E N E R C 0 N Reactor Vessel Water Levels

&cellen ce - Ev~ry project. Every day. During Refueling for EAL REV. 0 Thresholds APPENDIX C - ORIGEN-5 INPUT FOR SOURC E TERM CALC ULATION

=origens ~ Call Origen - S Sequence 0$$ all 71 e t ~L ogical Unit Assignments

-Binary Photon Library (71)

PWR Source Term PB EAL Analysis ~Case Title 3$$ 21 1 1 a4 27 al6 4 a33 19 e t ~L ibrary Integer Constants

-Units 83** Card Ci (4)

-Gamma Energy Groups (19) 35$$ 0 t ~ Not Used 54$$ a8 0 all 2 e ~ Special Calculation Optio ns

-Cutoff Value (Default )

-(a , n) Composition Dependent 56$$ 0 6 a6 1 alO 0 al3 66 3 3 0 2 0 e ~ Subcase Control Constants

-Decay Only Subcase (0)

-Number of Time Intervals (6)

-Number of Nuclides (66)

-Unit of Time in Hours (3) 57** 0 a3 1-16 e ~ No t Used 95$$ 0 t ~ Not Used PBEAL ~ Subcase Title Ci Source Terms ~Subc ase Basis 60** 0 24 40 50 60 70 ~ Time (hours) 61** 5rl-8 1+6 1+4 ~Cuto ff Values 65$$ ~Decay Period Pr i nt Triggers

' GRAM-ATOMS GRAMS CURIES WATTS-ALL WATTS - GAMMA 3Z 0 1 0 1 0 0 1 0 0 3Z 6Z 3Z 1 1 1 1 0 1 1 1 1 3Z 6Z 3Z 1 1 1 1 1 1 1 1 1 3Z 6Z 81$$ 2 0 26 1 e ~Gamma Source Constants 82$$ f2 ~P roduces Gamma Source Spectrum 83** 1.10E+07 l.OOE+07 8.00E+06 6.50E+06 5.00E+06 4.00E+ 06 3.00E+06 ~ Gamma Energy Groups 2.50E+06 2.00E+06 1. 66E+06 1. 33E+06 1. OOE+06 8.00E+05 6.00E+05 4.00E+05 3.00E+05 2 . 00E+05 1. OOE+05 5.00E+04 l.OOE+04 e 84** 2.00E+07 6.43E+06 3.00E+06 1.85E+06 1.40E+06 9. OOE+05 4 . OOE+05 ~N eutron Energy Groups l.OOE+05 1.70E+04 3.00E+03 5.50E+02 l . OOE+02 3.00E+Ol l.OOE+Ol (Not Used) 3.05E+OO 1 . 77E+OO 1 . 30E+OO l.l3E+OO 1. OOE+OO 8.00E-01 4.00E-Ol 3.25E-01 2.25E-Ol 1. OOE-0 1 5.00E-02 3.00E-02 l.OOE-02 l.OOE-05 e 73$$ 360850 360851 360870 360880 541311 541330 541331 54 1 350 ~ Nuclide Identifiers 541351 541380 531300 531310 53 1320 531330 531340 531350 370860 55 1 340 551360 551370 551380 521270 521271 521290 521291 521311 521320 511270 511290 380890 380900 380910 380920 561390 561400 441030 441050 441060 451050 520991 420990 581410 581430 581440 942380 942390 942400 942410 932390 390900 390910 390920 390930 410950 400950 400970 571400 571420 591430 601470 9524 10 962420 962440 74** 6. 15 E+05 1.36E+07 2.68E+07 3 . 60E+07 5.55E+05 1.02E+08 3.21E+0 6 ~ Nuclide Concentrations (Ci) 2.17E+07 2 . 20E+07 9.05E+07 1. 05E+06 5.10E+07 7.47E+07 1 . 06E+08

1. 1 9E+08 1. 01E+08 9.95E+04 9 . 52E+06 2.14E+06 6.27E+06 9.89E+07 4 . 54E+06 7.48E+05 1 . 33E+07 2 . 52E+06 9 . 95E+06 7.30E+07 4 . 63E+06

1. 42 E+ 07 5 . 03E+07 4.80E+06 6.30E+07 6.73E+07 9.42E+07 9 .0 5E +0 7 7 . 79E+07 5.42E+07 2.54E+07 5.08E+07 8.47E+07 9.62E+07 8.52E+07 8.03E+07 6 . 72E+0 7 1.33E+05 1 .45 E+0 4 2 . 25E+04 5 . 73E+06 9 . 65E+08 5.01E+06 6.56E+07 6.82E+07 7.6 7E+07 8.87E +07 8.76E+07 8.80 E+07 9.69E +0 7 8.25E+07 7.75E+07 3 . 33E+07 6.16E+03 1. 70E+06 1 . 58E +05 75$$ 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3~Library Kind 3 3 3 3 3 3 3 3 3 3 3 3 2 2 2 2 2 3 3 3 3 3 3 3 3 3 3 3 2 2 2 2-Actinide 3-Fission Product t

56$$ fO t Page 37 of 37

CALC NEE-363-CALC-001 Attachment 1 NO.

CALCU LATION PREPARATION

.::~ ENERCON Excellence-Every project Every doy CHECKLIST REV. 0 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? D D ~

The Calculation is performed in accordance with ENERCON procedures.

2. Are the proper forms being used and are they the latest revision? ~ D D The Calculation is performed in accordance with ENERCON procedures.

3. Have the appropriate client review forms/checklists been completed? D D ~

OAR will be performed after calculation submittal

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure? ~ D D

5. Is all information legible and reproducible? ~ D D

6. Is the calculation presented in a logical and orderly manner? ~ D D

7. Is there an existing calculation that should be revised or voided? D ~ D There is no existing calculation that should be revised or voided.

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation?

D ~ D No existing calculation would be applicable.

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 D D ~

apply to the calculation revision being performed .

No existing calculation is used for design inputs

10. Is the format of the calculation consistent with applicable procedures and expectations? ~ D D 11 . Were design input/output documents properly updated to reference this calculation? D D ~

There are no design output documents.

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

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation?

~ D D

14. Does the calculation provide a clear statement of quality classification? ~ D D

15. Is the reason for performing and the end use of the calculation understood? ~ D D

16. Does the calculatio'n provide the basis for information found in the plant's license basis? D ~ D This does not provide basis for license basis Page 1 of 5

CALC NEE-363-CALC-001 Attachment 1 NO.

CALCULATION PREPARATION F.~ ENERCON CHECKLIST Excellence-Every projecr. Every doy.

REV. 0 CHECKLIST ITEMS 1 YES NO N/A

17. If so, is this documented in the calculation? D D ~

See above

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D D ~

This does not provide basis for design basis

19. If so, is this documented in the calculation? D D ~

See above

20. Does the calculation otherwise support information found in the plant's design basis documentation? D ~ D This does not provide support for information found in design basis documentation 21 . If so , is this documented in the calculation? D D ~

See above

22. Has the appropriate design or license basis documentation been revised , or has the change notice or change request documents being prepared for submittal?

D I D ~

See above DESIGN INPUTS I I I

23. Are design inputs clearly identified? I ~ I D I D

24. Are design inputs retrievable or have they been added as attachments? I ~ I D I D

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

I D I D

26. Are design inputs clearly distinguished from assumptions? I ~ I D I D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation? I ~

I D I D The Design Information Transmittal is included as an Attachment is properly referenced in the calculation 28 . Are input sources (including industry codes and standards) appropriately selected I and are they consistent with the quality classification and objective of the calculation? ~

I D I D

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

~

I D I D

30. If applicable, do design inputs adequately address actual plant conditions? I ~ I D I D

31. Are input values reasonable and correctly applied? I ~ I D I D

32. Are design input sources approved? I ~ I D I D The Design Information Transmittal contains information from a superseded calculation.

33. Does the calculation reference the latest revision of the design input source? I ~ I D I D The calculation uses information from a superseded calculation. This information is provided in a Design Information Transmittal.

34. Were all applicable plant operating modes considered? I ~ I D I D Page 2 of 5

CALC NEE-363-CALC-001 Attachment 1 NO.

CALCULATION PREPARATION fti.l ENERCON CHECKLIST Excelfence-Every projecr. Every doy.

REV. 0 CHECKLIST ITEMS 1 YES NO N/A ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? ~ D D

36. Is adequate justification/basis for all assumptions provided? ~ D D

37. Are any engineering judgments used? D ~ D Engineering judgement not used as design input.

38 . Are engineering judgments clearly identified as such? D D ~

Engineering Judgement is not used as a design input.

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, D D ~

engineering principles, physical laws or other appropriate criteria?

Engineering Judgement is not used as a design input.

METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing basis?

D D ~

The scope of calculation is outside of plant licensing basis 41 . If the methodology used differs from that described in the plant's licensing basis , has the appropriate license document change notice been initiated? D D ~

see above.

42. Is the methodology used consistent with the stated objective? ~ D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? ~ D D BODY OF CALCULATION 44 . Are equations used in the calculation consistent with recogn ized engineering practice and the plant's design and license basis? ~ D D

45. Is there reasonable justification provided for the use of equations not in common use? D D ~

There are no uncommon equations used in the calculation.

46. Are the mathematical operations performed properly and documented in a logical fashion? ~ D D

47. Is the math performed correctly? ~ D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied?

~ D D

49. Has proper consideration been given to results that may be overly sensitive to very small changes in input? ~ D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation? ~ D D Page 3 of 5

CALC NEE-363-CALC-001 Attachment 1 NO.

CALCULATION PREPARATION F.r:d ENERCON CHECKLIST

£xcelfence-Every project. Every day.

REV. 0 CHECKLIST ITEMS 1 I YES I NO I N/A MCNP and Scale are used 51 . Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met? J 0 j D I D

52. Are the codes properly identified along with source vendor, organization, and revision I level? 0 I D I D

53. Is the computer code applicable for the analysis being performed? I 0 I D I D

54. If applicable, does the computer model adequately consider actual plant conditions? I 0 I D I D

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? I 0 I D I D 56 . Is the computer output clearly identified? I 0 I D I D

57. Does the computer output clearly identify the appropriate units? I 0 I D I D

58. Are the computer outputs reasonable when compared to the inputs and what was expected? I 0 I D I D

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? J 0 _) D -r D RESULTS AND CONCLUSIONS I I I

60. Is adequate acceptance criteria specified? I D I D I 0 There is no acceptance criteria as discussed in calc.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and D

intended use? I I D I 0 See above

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable D

licensing commitments and industry codes, and standards? I I D I 0 See above

63. Do the calculation results and conclusions meet the stated acceptance criteria? J D I D I 0 See above.

64. Are the results represented in the proper units with an appropriate tolerance, if applicable? I 0 I D I D

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives? I 0 I D I D

66. Is sufficient conservatism applied to the outputs and conclusions? I 0 I D I D Page 4 of 5

CALC NEE-363-CALC-001 Attachment 1 NO.

If)

,J,__. EN ERCON CALCULATION PREPARATION CHECKLIST fr..-cllcH::t' * - h,*r~* fJ/ujeo . EvtiY dat REV. 0 CHECKLIST ITEMS 1 I YES I NO I N/A

67. Do the calculation results and conclusions affect any other calculations? I D I ~ I D No other calculations are affected by this calculation .

68. If so, have the affected calculations been revised? I D I D I ~

No other calculations are affected by this calculation.

69. Does the calculation contain any conceptual, unconfirmed or open assumptions D ~ D requiring later confirmation? I I I There are no open assumptions requiring confirmation later.

70. If so, are they properly identified? l D l D l ~

There are no open assumptions requiring confirmation later.

DESIGN REVIEW I J I

71. Have alternate calculation methods been used to verify calculation results? I D I ~ l D No a Design Review was performed .

Note:

1. Where required, provide clarification/justificatio r answers to the questions in the space provided below each question. An explanation is required for any e lions a wered as "No' or "N/A".

Originator: Jay Bhatt

{) Lf / fie/ /7 Date Page 5 of 5

NEE-363-CALC-001, Rev. 0 Attachment 2 Pag e 2 of 5 8

Containment High Range gamma detectors read out in R/hr, and have a range of 1 R/hr to 10 R/hr (source: Vendor Technical Manual 978, General Atomics)

Parameters Value Unit Reference*

Fuel Type Westinghouse 14x14 Table 2-1

Fuel Rods per Assy 179 Table 2-1 Assembly Array 14x14 Table 2-1 Assembly Width 7.761 [in] Table 2-1 0.96 Eval 2016-0005 Density (%of theoretical STD (rounded to 2 Fuel) significant digits)

Fuel Pellet OD 0.3659 [in] Table 2-1 Fuel Rod OD 0.422 [in] Table 2-1 Clad Thickness 0.0243 [in] Table 2-1 Active Length 143.25 [in] Table 2-1 Diameter of Active Fuel 96.3 [in] Derived**

RPV Thickness (including clad) 6.656 [in] Dwg. 117802E Liner Thickness 1/4 [in] Drawing C-123

Unless otherwise noted above, all data in this table is from the Reload Transition Safety Report for the Point Beach Units 1 and 2 Fuel Upgrade to 422 Vantage Plus Fuel ("RTSR"L Revision 2, filed as Report NPC 1999-05997.

- There are 121 fuel assemblies in the core (Eval 2016-0005) . Each assembly is 7.761 inches square per the above table:

A= 121 x (7.761 in) 2 = 7,288 in 2 The equivalent diameter is therefore:

7288in 2

- - - = 96.3in 1[

The vertical distance from the top of the active fuel to the centerline of the reactor outlet nozzles is approximately 50" [1189E025 Rev 0 and 6669E93 Rev 1].

NEE-363-CALC-001, Rev. 0 Attachment2 Page 3 of 5 Table 5-1 Design Basis Core Shutdown Source Term (Calculation CN-CRA-08-21 Revision 1)

Isotope Ci Isotope Ci Kr-85 6.15E+05 Te-127 4.54E+06 Ce-143 8.03E+07 Kr-85m 1.36E+07 Ba-139 9.42E+07 Ce-144 6.72E+07 Kr-87 2.68E+07 Ba-140 9.05E+07 Pu-238 1.33E+05 Kr-88 3.60E+07 Te-127m 7.48E+05 Pu-239 1.45E+04 Xe-131m 5.55E+05 Te-129 1.33E+07 Pu-240 2.25E+04 Xe-133 1.02E+08 Te-129m 2.52E+06 Pu-241 5.73E+06 Xe-133m 3.21E+06 Te-131m 9.95E+06 Np-239 9.65E+08 Xe-135 2.17E+07 Te-132 7.30E+07 Y-90 5.01E+06 Xe-135m 2.20E+07 Sb-127 4.63E+06 Y-91 6.56E+07 Xe-138 9.05E+07 Sb-129 1.42E+07 Y-92 6.82E+07 1-130 1.05E+06 Sr-89 5.03E+07 Y-93 7.67E+07 1-131 5.10E+07 Sr-90 4.80E+06 Nb-95 8.87E+07 1-132 7.47E+07 Sr-91 6.30E+07 Zr-95 8.76E+07 1-133 1.06E+08 Sr-92 6.73E+07 Zr-97 8.80E+07 1-134 1.19E+08 Ru-103 7.79E+07 La-140 9.69E+07 1-135 1.01E+08 Ru-105 5.42E+07 La-142 8.25E+07 Rb-86 9.95E+04 Ru-106 2.54E+07 Pr-143 7.75E+07 Cs-134 9.52E+06 Rh-105 5.08E+07 Nd-147 3.33E+07 Cs-136 2.14E+06 Te-99m 8.47E+07 Am-241 6.16E+03 Cs-137 6.27E+06 Mo-99 9.62E+07 Cm-242 1.70E+06 Cs-138 9.89E+07 Ce-141 8.52E+07 Cm-244 1.58E+05

NEE-363-CALC-001 , Rev. 0 Attachment 2 Page 4 of 5 Upper Internals/Upper Fuel Hardware To approximate volume of metallic reactor internals is needed to estimate the shielding such components provide. A review of available documents found that calculation N-93-015 performed a detailed evaluation of the water displacement volume of such components. Calculation N-93-015 has now been superseded by another calculation that does not provide a detailed accounting of the reactor internal metal volumes. However/ based on a review of the contents of N-93-015 1 it is apparent that the estimates of metallic component volumes remains valid and appropriate for use in establishing EAL initiating conditions.

Calculation N-93-015 Revision 0 lists the following volumes:

Volume of steel from the top of the upper core plate to % pipe is 2I.O I9 fe (pg 3I)

Volume of steel from% pipe to I ft below the reactor vessel flange is 41.837 ft 3 (pg 3I)

Volume of the Upper Core Plate is 4.797 ft 3 (pg 26).

Volume of the Upper Supp011 Plate is I.30083 ft 3/inch x 4"=5.2 ft 3 (pg 30)

Decay time for Fuel Handling Accident

1) The latest time after reactor shutdown that we typically are handling fuel (i.e. end of the core re-load evolution). The basis for this value is a review of recently scheduled and/or completed refueling outages.

a. Unit 2 35th refueling:

i. Rx shutdown at 0034 on 3/18/2017 (source: Station Logs) ii. Core reloaded by I93I on 4/4/20I7 (source: Station Logs) iii. Duration: 403 hours0.00466 days 0.112 hours 6.66336e-4 weeks 1.533415e-4 months

b. Unit I 36th refueling:

i. Rx shutdown at OI07 on 3/12/20I6 (source: Station Logs) ii. Core reloaded by 0404 on 3/26/20I6 (source: Station Logs) iii. Duration: 339 hours0.00392 days 0.0942 hours 5.605159e-4 weeks 1.289895e-4 months

c. Unit 2 34th refueling

i. Rx shutdown at 0048 on 10/3/20I5 (source: Station Logs) ii. Core reloaded by 04I4 on I0/18/20I5 (source: Station Logs) iii. Duration: 364 hours0.00421 days 0.101 hours 6.018518e-4 weeks 1.38502e-4 months The maximum duration from these three outages was therefore 403 hours0.00466 days 0.112 hours 6.66336e-4 weeks 1.533415e-4 months . Rounding up to nearest full day (17 days) gives 408 hours0.00472 days 0.113 hours 6.746032e-4 weeks 1.55244e-4 months . This is judged to be an adequate approximation for the purposes of establishing the EAL IC.

2) The earliest time to start fuel movement. As with the latest time to complete fuel handling, this is based on a review of recently scheduled or completed refueling outages.

a. Unit 2 35 1h refueling (scheduled to start 3/18/2017):

i. Rx shutdown at 0034 on 3/18/2017 (source: Station Logs) ii. Core off-load started at 2031 on 3/25/17 (source: Station Logs) iii. Duration: 188 hours0.00218 days 0.0522 hours 3.108466e-4 weeks 7.1534e-5 months

b. Unit 1361h refueling:

i. Rx shutdown at 0107 on 3/12/2016 (source: Station Logs)

NEE-363-CALC-001, Rev . 0 Attachment 2 Page 5 of 5 ii. Permission to offload the core logged at 1214 on 3/17/2016 (source: Station Logs) iii. Duration: 131 hours0.00152 days 0.0364 hours 2.166005e-4 weeks 4.98455e-5 months

c. Unit 2 341h refueling

i. Rx shutdown at 0048 on 10/3/2015 (source : Station Logs) ii. Permission to offload the core logged at 1010 on 10/12/2015 (source: Station Logs) iii. Duration: 225 hours0.0026 days 0.0625 hours 3.720238e-4 weeks 8.56125e-5 months The minimum duration was 131 hours0.00152 days 0.0364 hours 2.166005e-4 weeks 4.98455e-5 months . Rounding down to the nearest full day (5 days) gives 121 hours0.0014 days 0.0336 hours 2.000661e-4 weeks 4.60405e-5 months . This is judged to be an adequate approximation for the purposes of establishing the EAL IC, though shorter durations (e.g. for head lift) may also be acceptable.

Fuel Assembly Activity from Calculation CN-CRA-08-14 Revision 0 (Table 4.6.1.4.1-2):

Nuclide Core Activity Activity of 1 (121 assys) average fuel assy

@, 65h (Ci) @, 65h (Ci) 1-130 2.77E+04 2.29E+02 1-131 4.15E+07 3.34E+05 1-132 4.23E+07 3.50E+05 1-133 1 .25E+07 1 .03E+05 1-135 1.06E+05 8.76E+02 Kr-85m 5.92E+02 4.89E+OO Kr-85 6.14E+05 5.07E+03 Kr-87 1.13E-08 9.34E-11 Kr-88 4.65E+OO 3 .84E-02 Xe-131m 5.47E+05 4.52E+03 Xe-133m 2.00E+06 1.65E+04 Xe-133 8 .41E+07 6 .95E+05 Xe-135m 1.73E+04 1.43E+02 Xe-135 1.75E+06 1.45E+04 Worst Case Radial Peaking Factor from Calculation CN-CRA-08-14 Revision 0 (Table 4.6.1.4.1-2):

RPF = 1.7 Rods per Assembly and Total Rods from Reload Transition Safety Report for the Point Beach Units 1 and 2 Fuel Upgrade to 422 Vantage Plus Fuel ("RTSR"), Revision 2, filed as Report NPC 1999-05997:

Rods per Assembly 179 Total Fuel Rods 21,659

ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENTAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" CALCULATION NEE-363-CALC-002, REVISION 0 FUEL HANDLING ACCIDENT MONITOR RESPONSE FOR EAL THRESHOLDS 39 pages follow

CALCULATION COVER SHEET (Page 1 of 1)

Document Information:

Calculation (Doc) No: NEE-363-CALC-002 I Controlled Documents Revision: 0

Title:

Fuel Handling Accident Monitor Response for EAL Tlu*esholds Type: CALC Sub-Type: Discipline: MECH Facility: PB I Unit: 0 Safety Class: 0 SR 0 Quality Related [g] Non-Nuclear Safety 0 Impmiant to Safety 0 Not Impmiant to Safety Special Codes: 0 Safeguards 0 Proprietary Vendor Doc No: NEE-363-CALC-001 I Vendor Name or Code: ENERCON Executive Summary (optional):

Review and Approval:

Associated EC Number: 288568 EC Revision: 0 ARI Other Document Number:

Description of Calculation Revision: Initial issue EC Document Revision: 0 Prepared by: N/A (V eridor) Date: - -

(signature) (print name)

Reviewed by: --/:~ 7.2*. l<'f;Jf/IJU Date: ~*

(signature) (print name)

Type of Review: 0 Design Verification 0 Review [g] Owner Acceptance Review Method Used (ForDVOnly): 0 Design Review 0 Altemate Calculation Approved by: *~0-,.J\11!.~,= ----JcLAC- I - ~~e ""b Date: :r-.s. 1 7

\,

( s1gnature

. \ ) I (print name)

EN-AA-1 00-1 004-F01 , Revision 0

CALCULATION REVISION

SUMMARY

SHEET (Page 1 of 2)

Calculation Number: NEE-363-CALC-002 Rev. Affected Pages Reason for Revision 0 All (39 pages) Initial issue. Detailed contents:

Site cover sheet (I page)

Rev. summary sheet (I page)

Vendor calculation (23 pages)

Vendor Appendix A (I page)

Vendor Attachment 1 (2 pages)

Vendor Attachment 2 (1 page)

Vendor Calculation Preparation Checklist (5 pages)

Design Information Transmittal (5 pages)

EN-AA-1 00-1 004-F02, Revision 1

CALC NO. NEE-363-CALC-002 0 ENERCON t.w lici'Kt--1.-tt> f*Cl (Y't'lf ckJy CALCULATION COVER SHEET REV. 0 PAGE NO. 1 of 23

Fuel Handling Accident Monitor Response for Client; Nextera - Point Beach

Title:

EAL Thresholds Project Identifier: NEE-363 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary 0 ~

Information, that require confrrmation? (If YES, Identify the assumptions.)

2 Does this calculation serve as an "Alternate _Calculation"? (If YES, identify the 0 [81 design verlfled calculation.)

Design Verified Calculation No. _ _

3 Does thls calculation supersede an existing Calculation? (If YES, identify the 0 [8]

design verified calculation.)

Superseded Calculation No.

Scope of Revision:

Initial Issue Revision Impact on Results:

Initial Issue Study Calculation 0 Final Calculation fZl Safety-Related D Non-Safety-Related ~

(Print Name and Sign)

Originator: Ryan Skaggs

/~-./S-,7~ Date:'/ /U)//7 Oe51gn Verlfler 1 (Reviewer If NSR): Caleb Traino/A-~-

Date: i--J/.,)-6/j 7'-

Approver: Aaron Holloway ~ ~*~...,.. ~ Date:

Lf/21 /17 Note 1; For ncm-safe lated caloulation, des lg n verification arfbe substHuted by review. I I

CALC NO. NEE-363-CALC-002

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A 1 0 1 2 0 2 1 3 5 4 5

CALC. NO. NEE-363-CALC-002

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PAGE NO. 3 of 23 Section Page

1. Purpose and Scope ........... ..... ... ................ ............................ ..... .. ... ... ... ........ .. ........ ........... 6

2. Summary of Results and Conclusion ............................................................ ............. .. ....... 6

3. References ..... ...... .. .... ..... .. ..... .. .. ......... ... .... ..... ... ......... ....... .. ..... .. ....... .... ......... .... ............. .. 7

4. Assumptions ... .... ... .. ........ ..... ........ .. .................. .. .. .. ................... .. ............................ ........... 8

5. Design lnputs ............ .. ..... ...... ...... .. ............. .. ................................ .. ..... .. .... .... ....... .. ........... 10 5.1 Source Term ............ .. .......... ............. ........... ... .. .. .......... .. ............ .............. .. ................ 10 5.2 Model Dimensions .... .. ...... .... ........................ ............. .... ...... ....................................... 10 5.3 Nuclide Gap Fraction ................ .. .. ... ........ .. .................... .... ...... .. ................................. 11 5.4 Detector Range ....................... .. ... .... .................... .... ................................................... 11 5.5 Decontamination Factors ............................................................................................ 11 5.6 ANSI/ANS Flux-Dose Conversion Factors ..................................................... ............. 12

5. 7 Steel Liner .......... .. ................... .... ..... ........................... ........... .. ............. .. ......... ... ... .. .. 12

6. Methodology ........................................................................ ........... ...... ...... ............. .. .. .. .... 13

7. Calculations .. .......... .. ...... ... ................................................................................................ 14 7.1 Release Activity from Pool ...... .. ........ ... .. .. .............. .. ....... ......... ..... ...... .. ...... ....... ........ . 14 7.2 Source Terms .. ............ .... .... ......................................... ..... .... ...... ......... .. ... ... .............. 14 7.3 MCNP Model Geometry ...... .... ....... ............................................................................. 16 7.4 MCNP Source Definition ........ .. ............ ................................ ......... ................. .. ........... 20 7.5 MCNP Tally Specification ... ...... ............................. .. ... ... ..... ........ ............. ...... ... .. ...... ... 22 7.6 MCNP Material Cards .. .. ... ...... .. .... .... .... .... .. .............. ..... .... ......................................... 23 7.7 Results .... .................. .................. .................. .. ..... .................. ................... ...... .... ... .. ... 23

8. Computer Software ............................................................................................................23

CALC. NO. NEE-363-CALC-002

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PAGE NO. 4 of23 List of Figures Figure Page Figure 7-1 ORIGEN-S Input Deck for MCNP Source Term Calculation .... .. ........ ...... ... .... ....... ... 15 Figure 7-2 Y-Z VISED Plot of RCB ........................... ..................... ................ .. .............. .. .......... 17 Figure 7-3 X-Y VISED Plot of RCB .. .. ... ........ ............ ...... ............. ........... ... .... .. .. .......... .. ............ 17 Figure 7-4 Y-Z VISED Plot of SFP/Auxiliary Building (looking North) ............ .. .. .. ...................... 18 Figure 7-5 X-Y VISED Plot of SFP/Auxilary Building at 66' El. .. .... .............. ............... .. ............. 18 Figure 7-6 MCNP RCB Model Surface Cards ......................... ......... ......... ........ .... .................... 19 Figure 7-7 RCB MCNP Model Cell Cards .................................................... ... ... ...... ....... .. ......... 19 Figure 7-8 SFP/Auxiliary Building MCNP Model Surface Cards .......... ... ......... ............ .. ............ 19 Figure 7-9 SFP/Auxiliary Building MCNP Model Cell Cards ........ ...... .................................... .... 19 Figure 7-10 MCNP Source Definition Cards ...................... ... ...... ..... .... ................................ .. .... 21 Figure 7-11: MCNP Tally Cards .. .. ................... .. .................................... .......... .. .. ..... .. ....... .. .... 22 Figure 7-12 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors .... .... ....... ........ .. .22 Figure 7-13 MCNP Material Cards ... ... ..... .................................................................................23

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TABLE OF CONTENTS REV.

CALC. NO. NEE-363-CALC-002 0

PAGE NO. 5 of 23 List of Tables Table Page Table 2-1 Detector Response .. .. .... ........ ..... .................. ............. .... .... ... ................. ..... .. .. ........... 6 Table 4-1 Detector Locations .... .. ...... .. ................. ............... .. ....... .. ...... .. .. ... .......... ... .......... .. .. .... 9 Table 5-1 Activity in an Average Fuel Assembly .. .... ........ ........................... ..... ....... ... ... .. .......... .10 Table 5-2 RCB Model Dimensions ............. ............... ........ ...... .... .... ... ... ......... ........... ..... ...... ..... 10 Table 5-3 SFP Model Dimensions ....... ......... .. ... .... ... ....... ..... .. ........ ..... ..... ... ........ .. .. ....... ........... 11 Table 5-4 Detector Range ......... .. ..... .. ... .. ...... ..... ..... ........... .. ....... ... .. ..... ....... ........ .............. .. ... .. 11 Table 5-5 Dose Flux Conversion Factors ................... .............. .. ........ ... ... ... ........ ......... ......... .... 12 Table 7-1 Calculations ............. ... ...... ..... ........ ..... .. ........ .... .. ... .. .................... .. ...... .. .... ... ......... ...14 Table 7-2 Binned Total Core Source Term .. .. .......... ........ ... .... ........ ............ .. ................... .......... 16 Table 7-3 Dose Rate Response at RCB/SFP Detectors (mrem/h) ...... .. ... ... .. ...... ........... ... .... ... .. 23

CALC. NO. NEE-363-CALC-002

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1. Purpose and Scope The purpose of this calculation is to determine the expected dose rates on radiation monitors RE-126, RE-127, RE-128, and RE-135 during a fuel handling accident (FHA) at Point Beach Nuclear Plant (PBNP). Monitors RE-126, RE-127, and RE-128 are located in the reactor containment building (RCB) . Monitor RE-135 is inside the Auxiliary (AUX) Building near the Spent Fuel Pool (SFP). The accident occurs either in the SFP or RCB 408 hours0.00472 days 0.113 hours 6.746032e-4 weeks 1.55244e-4 months after shutdown. The results are used as threshold values for Emergency Action Level (EAL) RA2.2 in the PBNP EAL Technical Basis document, which implements NEI 99-01 , Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors". The containment building , the auxiliary building, and components within the buildings are modeled simplistically because only order of magnitude results are needed. As such, the dose rate results should be considered as reasonably representative of the magnitude of the actual dose rate only. This calculation is not Nuclear Safety Related as the results of the calculation does not affect the design basis or Safety Related systems structures or components. This calculation represents an as built analysis of plant conditions, therefore no acceptance criteria is required .

2. Summary of Results and Conclusion The results of this calculation are listed below.

Table 2-1 Detector Response Location Monitor Dose Rate (Rih)

RE-126 5.73 RCB RE-127 5.57 RE-128 5.57 SFP RE-135 3.84 Reading levels at or above the values listed in Table 2-1 will be indicative of a fuel handling accident. The dose rates reported do not include the background (ambient) radiation readings associated with the monitor calibration.

CALC. NO. NEE-363-CALC-002

. Fuel Handling Accident

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EAL Thresholds PAGE NO. 7 of 23

3. References

1. Design Input Transmittal (Attachment 4)

2. CGDG-SCALE-6.1.2, Revision 00, Commercial Grade Dedication SCALE Version 6.1 .2 .

3. CGDG-MCNP6-V1.0, Revision 00, Commercial Grade Dedication MCNP6 Version 1.0.

4. ANSI/ANS 6.1.1-1977, Neutron and Gamma Flux-To-Dose Conversion Factors.

5. C-129, Rev. 10, Concrete Containment Structure Interior Plans El 46'-0". El 66'-0". El 76'-0". & El 100'-5 1/2"

6. C-141, Rev. 10, Concrete Auxiliary BLDG. Central Part Plan at Elevations 46'-0" & 66'-0"

7. C-160, Rev. 10, Concrete Auxiliary BLDG. Spent Fuel Pool Plan and Sections

8. C-162, Rev. 07, Concrete Auxiliary BLDG.- Truck Access and Drumming Station Plans

9. C-304, Rev. 08, Auxiliary Building Elevations at Col. Line G, U & 9.9

10. M-9, Rev. 13, U1 Equipment Location Sections A-A & B-B

11. M-500, Rev. 08, Instrument Location Plan Containment Operating Floor & Mise Upper Floors South

12. Regulatory Guide 1.183, alternate radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors, July 2000

13. FSAR, Section 14.2.1, Table 14.2.1-1, Section 11.5

14. CN-CRA-08-14, Rev. 0, Point Beach- Fuel Handling Accident Doses for the EPU

15. NEI 99-01, Rev. 06, "Development of Emergency Action Levels for Non-Passive Reactors"

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

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4. Ass umptions The following assumptions are used in this calculations:

1. The fuel handling accident results in a gas release which is assumed to disperse instantaneously throughout the mixing volume. While the monitors may see higher concentrations initially, the lowest concentration is after full dispersal.

2. The RCB building volume is assumed to be the entire structure above the 66' 0" operating floor consisting of a cylinder and a curved dome top and dimensions taken from Reference 10. Using only the free volume above the fuel handling deck is appropriate because the congestion in the lower containment compartments would delay mixing and the gas released will be hot therefore rising, this scenario is not conservative but realistic. The free volume above the operating floor does not consider any of the components or structures occupying the space to simplify the model. This model simplification will not significantly impact the results as there is a low volume of structures above the operating floor elevation.

3. The SFP is assumed to a majority of the Auxiliary Building between the two containment units. The auxiliary building dimensions were taken from References 6, 7, and 8. The room was modeled to include sections of the Auxiliary building that are on the 46' elevation and the 26' elevation in addition to the 66' elevation.

Similar to the RCB case, this free volume is over estimated because it does not consider any of the components or structures occupying the space to simplify the model since the results will still be within the magnitude needed.

4. This calculation assumes one damaged fuel bundle, or 179 rods per Reference

15. This represents a source term which will be large enough to be indicative of a fuel handling accident and exclude water lowering events or other causes of elevated radiation levels. Larger fuel handling accidents will also be recognized using this threshold.

5. The MCNP models locate the monitors for the RCB to be just on the inside of the source volume and immersed in the source volume. For the SFP the monitor is located approximately 26' east of the SFP on the 66' elevation and immersed in the source volume. The SFP monitor location is provided by a site walkdown.

The results of the walkdown were communicated through an email, see Attachment 1. The approximate locations of each detector are found in Table 4-1.

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

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EAL Thresholds PAGE NO. 9 of23 Table 4-1 Detector Locations 1 23 Detector X (Ft) Y (Ft) EL. (Ft) Reference RE-126 5' -51'-3" 70'-6" [11](Scaled based on RE-127 15' 50' 70-6" Column lines)

RE-128 -42' -31' 70-6" RE-135 25'-10" 48'-2" 69' [Attachment 1], [6] , [7]

6. For the MCNP model, the RCB is modeled as a cylinder rather than a cylinder with a small dome. Due to the large volume, and the detectors being on opposite sides of containment and separated by over 100 feet, these differences have a negligible effect.

7. A peaking factor of 1.0 will be used since it is the most probable scenario and will not overestimate the activity of the fuel.

8. A decay time of 408 hours0.00472 days 0.113 hours 6.746032e-4 weeks 1.55244e-4 months is assumed. This models a fuel handling accident at the end of the refueling period which would release the smallest amount of activity. The time of 408 hours0.00472 days 0.113 hours 6.746032e-4 weeks 1.55244e-4 months is based on the most recent outage core reloading time, see correspondence with PBNP in Reference 1.

9. Buildup of secondary particle radiation from photon scattering was ignored due to the small effect on the results with the detector inside the radiation source.

10. Per Reference 13, only the noble gas and iodine nuclides are used in the source since these are the nuclides that are expected to be released and escape the water when the fuel assembly is damaged.

11. The detectors read out in Rlh, which is a measure of the ionization caused by radiation. The MCNP output is provided in mrem/h which is a measurement of the equivalent dose that represents the biological effects of ionizing radiation .

The relationship between roentgens and milirem is not straightforward and depends on different absorption of particles in a medium. It is assumed that 1 R is approximately 1000 mrem. This is acceptable as only the gamma source term is considered .

1 The location of the detectors are estimated.

2 The +X and +Y directions for RE-126, RE-127, and RE-128 are based on the called north per Reference 11 . Origin is center of Containment (+X is east and +Y is north) 3 RE-135's origin is Column 13 and the line shown in Attachment 2 (+X is south and +Y is east)

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5. Design Inputs 5.1 Source Term The source term is provided in Reference 13. A table of the input values is shown in Table 5-1 below. These values shown for activity are after 65 hours7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months of decay.

T a bl e 51- A cf 1vny "t .man A verage F ue I Asse mbly Nuclide Activity @ 65h (Ci) 1-130 2.29E+02 1-131 3.43E+05 1-132 3.50E+05 1-133 1.03E+05 1-135 8.76E+02 Kr-85m 4.89E+OO Kr-85 5.07E+03 Kr-87 9.34E-11 Kr-88 3.84E-02 Xe-131m 4.52E+03 Xe-133m 1.65E+04 Xe-133 6.95E+05 Xe-135m 1.43E+02 Xe-135 1.45E+04 5.2 Model Dimensions The dimensions used for the models as well as the reference are listed in the tables below:

Table 5-2 RCB Model Dimensions Dimension I 1 Description I Reference Ft-in em Radius 52'-3" I 1592.58 I Containment Radius I [51 Height 90'-9" I 2766.06 I 66' EL. to the 156'-9" EL. 1 [1 01

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Monitor Response for REV. a EAL Thresholds PAGE NO. 11 of 23 Table 5-3 SFP Model Dimensions Dimension Description Reference Ft-in em Width 78' 2377.44 AUX building (all elevations) [6]

AUX Building 46' EL.

Length 87'-2" 2656 .84 Column Ja through P [6]

Height 2a' 6a9.6 46' EL. to the 66' EL.

AUX Building 66' El.

Length 2a4'-2" 6223 Column F through Line shown in [6]' [7], [8]

Attachment 2.

Height 42'- 1a" 13a5.56 66' EL. to the 1a8'-1 a" EL. [9]

AUX Building 26' EL.

Length 44' 1341.12 Between the line shown in Attachment [8]

2 through U Height 82'-1a" 2524.76 26' EL. to the 1a8'-1 a" EL.

5.3 Nuclide Gap Fraction The fission product gap inventories used are 12% for 1-131 , 3a% for Kr-85, and 1a% for all other noble gas and iodine nuclides and are from Reference 14.

5.4 Detector Range The ranges of the detectors are shown in Reference 13.

Table 5-4 Detector RanQe Detector Range RE-126 1ao -1 as Rlh r RE-127 1ao -1 as Rlh r RE-128 1ao -1 as Rlh r RE-135 1ao -1 a4 R/h r 5.5 Decontamination Factors From Reference 13 the effective decontamination factors used for the iodine is 2aa which accounts for scrubbing of the iodine as it evolves through the pool water with a minimum water level of 23 ft above the top of the reactor vessel flange and over the top of the assemblies in the

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

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5.6 ANSI/ANS Flux-Dose Conversion Factors Table 5-5 contains the flux conversion factors for each energy bin to convert the energy of the particles into a dose in mrem/hr from Reference 4.

Table 5-5 Dose Flux Conversion Factors Energy Flux Energy Flux Bin Multiplier Bin Multiplier 0.01 3.96E-03 1 1.98E-03 0.03 5.82E-04 1.4 2.51 E-03 0.05 2.90E-04 1.8 2.99E-03 0.07 2.58E-04 2.2 3.42E-03 0.1 2.83E-04 2.6 3.82E-03 0.15 3.79E-04 2.8 1.04E-03 0.2 5.01 E-04 3.25 4.41 E-03 0.25 6.31 E-04 3.75 4.83E-03 0.3 7.59E-04 4.25 5.23E-03 0.35 8.78E-04 4.75 5.60E-03 0.4 9.85E-04 5 5.80E-03 0.45 1.08E-03 5.25 6.01 E-03 0.5 1.17E-03 5.75 6.37E-03 0.55 1.27E-03 6.25 6.74E-03 0.6 1.36E-03 6.75 7.11 E-03 0.65 1.44E-03 7.5 7.66E-03 0.7 1.52E-03 9 8.77E-03 0.8 1.68E-03 11 1.03E-02 5.7 Steel Liner The interior of the RCB is completely lined with a welded steel plate as a barrier to vapor and gas leakage. A 1/4" thick steel liner will be used as the edge of containment model per Reference 1.

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6. Methodology The activities listed in Table 5-1 must be modified to find concentration of each nuclide that is released from the pool. To do this each nuclide activity is multiplied by a radial peaking factor, the nuclide's specific gap fraction, and one over the nuclide's decontamination factor. Equation 6.1 was used to calculate the release concentrations from the pool of each nuclide.

Release Concentr ation= RPF

Ci

GF

1/DF Equation 6.1 Where:

RPF is the radial peaking factor, which is 1.0 for all nuclides based on Assumption 4.5, Ci is the gap inventory concentration of each nuclide from Reference 13 and listed in Table 5-1 ,

GF is the gap fraction for each nuclide from Reference 14 and listed in Table 7-1.

DF is the decontamination factor from Reference 12 and listed in Table 7-1.

After release activity for the pool was calculated, the nuclide inventories were entered in ORIGEN-S of the SCALE 6.1 code package. The ORIGEN-S decay sequence is used to decay the design input isotope specific activities and then bin each activity into energy dependent photon bins. These energy specific photon emission bins are used as input for the energy distribution described by the MCNP source definitions.

MCNP6 Monte Carlo transport code is used to determine the dose rates.

The detailed engineering drawings are converted into MCNP surface and cell cards in the proper dimensions. The radiation monitors of interest are modeled as point detectors to determine the expected dose rate for those detectors.

The SFP/Auxiliary Building was modeled as three rectangular prisms to account for the different sections of the Auxiliary Building's mixing volume. The source is evenly distributed throughout the mixing volume and detector RE-135 is places at the location listed in Table 4-1. The origin of the model is located at the intersection of column 13 and the line shown in Attachment 2. The positive X-axis is to the South, the positive Y-axis is the East, and the positive Z-axis is up.

The RCB was modeled as a cylinder with a volume equivalent to the free space calculated in the RCB above the refueling floor. The cylinder is 90'-9" high with a 52'-6" radius . Detectors RE-126, RE-127, and RE-128 are placed in the location listed in Table 4-1 . The origin is located at the center of the RCB with the positive X-axis being east, positive Y-axis being north, and the positive Z-axis being up.

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

'*jJ ENERCON Monitor Response for REV. 0 Excelfence- Every project. Eve1y day EAL Thresholds PAGE NO. 14 of 23

7. Calculations 7.1 Release Activity from Pool Table 7-1 contains the calculations used to find the release activity from the pool.

Table 7-1 Calculations Nuclide Activity@ Gap Fraction Pool OF (0.1. Release 65h (0.1. 5.1) (0.1. 5.3) 5.5) Activity from Pool @65h Ci Ci 1-130 2.29E+02 0.10 200 1.15E-01 1-131 3.34E+05 0.12 200 2.06E+02 1-132 3.50E+05 0.10 200 1.75E+02 1-133 1.03E+05 0.10 200 5.15E+01 1-135 8.76E+02 0.10 200 4.38E-01 Kr-85m 4.89E+OO 0.10 1 4.89E-01 Kr-85 5.07E+03 0.30 1 1.52E+03 Kr-87 9.34E-11 0.10 1 9.34E-12 Kr-88 3.84E-02 0.10 1 3.84E-03 Xe-131m 4.52E+03 0.10 1 4.52E+02 Xe-133m 1.65E+04 0.10 1 1.65E+03 Xe-133 6.95E+05 0.10 1 6.95E+04 Xe-135m 1.43E+02 0.10 1 1.43E+01 Xe-135 1.45E+04 0.10 1 1.45E+03 The release activity from the pool is the input used in the ORIGEN-S source term input. It is found by using Equation 6.1.

7.2 Source Terms In order to convert the isotope specific activity into an energy spectrum , ORIGEN-S of the SCALE6.1 code package is used to initiate a decay and bin into 19 photon energy groups. The energy groups along with their associated activities are used in the MCNP source definition to model the anticipated radiation emission following shutdown.

The ORIGEN-S input deck, PBFHAEAL4, is provided below in Figure 7-1.

This input has a simple decay case where the inputted isotopic composition in curies is decayed . The isotope is specified in the 73$$ card using the special identifier described in Section F7 .6.2 of the ORIGEN-S manual, and the activity in curies is specified in the 74** card .

The time steps for the decay are given on the 60** card in hours. Although multiple time steps are calculated, the source term with 343 hours0.00397 days 0.0953 hours 5.671296e-4 weeks 1.305115e-4 months decay time is used in this calculation since the

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

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Monitor Response for REV. 0 EAL Thresholds PAGE NO. 15 of 23 original activities were based on 65 hours7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months after shutdown resulting in a total time after shutdown of 408h per Assumption 8Error! Reference source not found .. The output of the decay is iven in terms of photons/s/Energy-Group, which is automatically normalized in the MCNP input.

Figure 7-1 ORIGEN-5 Input Deck for MCNP Source Term Calculation origens E-Call Origen-S Sequence 0$$ a11 71 e t E-Logical Unit Assignments

-Binary Photon Library (71)

FHA Source Term Analysis E-Case Title 3$$2111 a427a164a3319et E-Library Integer Constants

-Units 83** Card Ci (4)

-Gamma Energy Groups (19) 35$$ 0 t E-Not Used 54$$ a8 0 a11 2 e E-Special Calculation Options

-Cutoff Value (Default)

-(a,n) Composition Dependent 56$$07a61 a100a131433020e E-Subcase Control Constants

-Decay Only Subcase (0)

-Number ofTime Intervals (6)

-Number of Nuclides (66)

-Unit of Time in Hours (3) 57**0 a31-16e E-Not Used 95$$ 0 t E-Not Used PBEAL E-Subcase Title Ci Source Tenns E-Subcase Basis 60** 0 24 40 50 60 70 343 E-Time (hours) 61** 5r1-8 1+6 1+4 E-Cutoff Values 65$$ E-Decay Period Print Triggers

'GRAM-ATOMS GRAMS CURIES WATIS-ALL WATIS-GAMMA 3Z 010100 100 3Z 6Z 3Z 1 1 1 101 111 3Z 6Z 3Z 1 1 1 111 111 3Z 6Z 81$$ 2 0 26 1 e E-Gamma Source Constants 82$$ f2 E-Produces Gamma Source Spectrum 83** 1.1OE+07 1.00E+07 8.00E+06 6.50E+06 5.00E+06 4.00E+06 3.00E+06 E-Gamma Energy Groups 2.50E+06 2.00E+06 1.66E+06 1.33E+06 1.00E+06 8.00E+05 6.00E+05 4.00E+05 3.00E+05 2.00E+05 1.00E+05 5.00E+04 1.00E+04 e 84** 2.00E+07 6.43E+06 3.00E+06 1.85E+06 1.40E+06 9.00E+05 4.00E+05 E-Neutron Energy Groups 1.00E+05 1.70E+04 3.00E+03 5.50E+02 1.00E+02 3.00E+01 1.00E+01 (Not Used) 3.05E+OO 1.77E+OO 1.30E+OO 1.13E+OO 1.00E+OO 8.00E-01 4.00E-01 3.25E-01 2.25E-01 1.00E-01 5.00E-02 3.00E-02 1.00E-02 1.00E-05 e 73$$ 360851 360850 360870 360880 541311 541331 E-Nuclide Identifiers 541330 541351 541350 531300 531310 531320 531330 531350 74** 4.89E-01 1.52E+03 9.34E-12 3.84E-03 4.52E+02 1.65E+03 6.95E+04 E-Nuclide Concentrations (Ci) 1.43E+01 1.45E+03 1.15E-01 2.06E+02 1.75E+02 5.15E+01 4.38E-01 75$$ 3 3 3 3 3 3 3 3 3 3 3 3 3 E-Library Kind 2-Actinide 3-Fission Product t

56$$ fO t End

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

. :il ENERCON Monitor Response for REV. 0 Excellence - Every projen Every day.

EAL Thresholds PAGE NO. 16 of 23 The results of this calculation are transmitted as electronic files and summarized below in Table 7-2. See Appendix 1 for the electronic file listing. These values will be used in the MCNP input source definition.

Table 7-2 Binned Total Core Source Term Energy Energy Boundaries Photons/sec Group (MeV) 1 0.01-0.05 2.248E+14 2 0.05-0.1 1.670E+14 3 0.1-0.2 1.805E+12 4 0.2-0.3 4.291 E+11 5 0.3-0.4 1.973E+12 6 0.4-0.6 2.692E+11 7 0.6-0.8 1.896E+11 8 0.8-1 1.238E+06 9 1-1.33 1.006E+06 10 1.33-1.66 3.031E+04 11 1.66-2 9.478E-04 12 2-2.5 8.139E-08 13 2.5-3 O.OOE+OO 14 3-4 O.OOE+OO 15 4-5 O.OOE+OO 16 5-6.5 O.OOE+OO 17 6.5-8 O.OOE+OO 18 8-10 O.OOE+OO 19 10-11 O.OOE+OO totals 3.695E+14 7.3 MCNP Model Geometry The following MCNP models geometry is based on the containment dimensions from References 10 and 11 and the SFP/Auxiliary building dimensions are from References 6, 7, and

8. The model only focuses on volume above the fuel handling floor for the RCB and for the volume of the Auxiliary building between the two containment units for the SFP. Visual Editor (VISED) plots of the model geometry for the RCB are provided in Figure 7-2 and Figure 7-3.

The VISED plots of the model geometry for the SFP/Auxiliary Building are provided in Figure 7-4 and Figure 7-5. The MCNP surface cards with the model dimensions (em) for the RCB and SFP are shown in Figure 7-6 and Figure 7-8, respectively. The cell cards for the RCB and SFP are shown in Figure 7-7 and Figure 7-9 respectively. Areas that are not of interest are given an

CALC. NO.

'a

... ENERCON Fuel Handling Accident NEE-363-CALC-002 Excellence-Every projec t. Every day Monitor Response for REV. 0 EAL Thresholds PAGE NO. 17 of 23 importance of zero (white areas) so MCNP will not track particles in locations that will not contribute to the detector response.

Figure 7-2 Y-Z VISED Plot of RCB 90 ' -9" Figure 7-3 X-Y VISED Plot of RCB

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident F.:il ENERCON Excellence- Every projecr. Eve1y day.

Monitor Response for REV. 0 EAL Thresholds PAGE NO. 18 of 23 Figure 7-4 Y-Z VISED Plot of SFP/Auxiliary Building (looking North)

Model EL. 66 '

EL. 46' EL. 26 '

Column U Figure 7-5 X-Y VISED Plot of SFP/Auxilary Building at 66' El.

Column F Modo) O<igio~

Column U

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident F.::d ENERCON Monitor Response for REV. 0 Excellence- Every projecr. Every day EAL Thresholds PAGE NO. 19 of 23 Figure 7-6 MCNP RCB Model Surface Cards c surfaces 1 rcc 0 0 0 0 0 2766.06 1592.58 $Containment Cylinder H=90'-9" R=52'-6" 2 rcc 0 0 0 0 0 2765.43 1591 .95 $ Inside Steel liner Figure 7-7 RCB MCNP Model Cell Cards c cells 11 1 -1 .21 E-03 -2 imp:p=1 $ Inside Containment liner 22 2 -7.94 -1 2 imp:p=1 $ Steel liner 3/8" thick 99 0 imp:p=O $ Problem Boundary Figure 7-8 SFP/Auxiliary Building MCNP Model Surface Cards c surface 1 rpp 0 2377.44 0 6223 0 1305.56 $AUX 66 EL X=78' Y=204'-2" Z=42'-10" 2 rpp 0 2377.44 1971.04 4627.88-609.6 0 $AUX 46 EL X=78' Y= 87'-2" Z = 20' 3 rpp 0 2377.44 -134112 0 -1219.2 1305.56 $AUX 26 EL X-78' Y=44' X=82'-1 0" Figure 7-9 SFP/Auxiliary Building MCNP Model Cell Cards c cells 11 1 -1.21 E-03 -1 imp:p=1 $66 EL 22 1 -1 .21 E-03 -2 imp:p=1 $46 EL 33 1 -1.21 E-03 -3 imp:p=1 $26 EL 99 0 #11 #22 #33 imp:p=O $ Problem Boundary

CALC. NO. NEE-363-CALC-002 I Fuel Handling Accident

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Monitor Response for REV. 0 EAL Thresholds PAGE NO. 20 of 23 7.4 MCNP Source Definition The core source term is assumed to be uniformly distributed throughout the mixing volume of the building per Assumption 0, and has an energy spectra based on the Origin program output.

Only the gamma source term is taken into account for this evaluation. The source is defined on the MCNP sdef card using distributions to define the particle location and energy. The source is defined by the three range limits in the X, Y and Z directions. These range limits were set to create a rectangle that is slightly larger than the mixing volume. The cell parameter added to the source term to only allow particles created in cell 11 to be counted. This will not affect the source magnitude, as this part of the evaluation is independent of magnitude. These parameters combined define the mixing volume where the particles can be born. The erg parameter defines the energy spectrum of source particles and is based on the results of the OR IGEN-S calculated in Table 7-2. This distribution is a histogram of energies represented by activities. These are automatically normalized by MCNP to create a probability distribution. The total activity is preserved in the tally multiplier. The MCNP source definition cards are shown below in Figure 7-10. The difference between the RCB and SFP source definition cards is that the three range limits are adjusted to be around their respective building and that the SFP source has a distribution associated with the three cells to allow for the particles to be created in all three cells.

CALC. NO. NEE-363-CALC-002 Fuel Handling Accident

~. Jl ENERCON Excellence-Every project. Every day Monitor Response for REV. 0 EAL Thresholds PAGE NO. 21 of 23 Figure 7-10 MCNP Source Definition Cards RCB:

sdef X=d1 Y=d2 z=d3 erg=dB par-2 cell=11 ~Source Definition Card

-X-range = d1

-Y-range = d2

-Z -range = d3

-Energy= dB

-Inside Cell11 si1 -1593 1593 ~X-range limits sp1 0 1 ~Unifonn probability on X si2 -1593 1593 ~Y-range Limits sp2 0 1 ~Unifonn Probability on Y si3 0 2767 ~Z-range Limits sp3 0 1 ~Unifonn Probability on Z c Fuel Gamma Spectra siB h 1.000e-002 5.000e-002 1.000e-001 2.000e-001 3.000e-001 4.000e-001 ~Source Energy Groups 6.000e-001 B.OOOe-001 1.000e+OOO 1.330e+OOO 1.660e+OOO 2.000e+OOO 2.500e+OOO 3.000e+OOO 4.000e+OOO 5.000e+OOO 6.500e+OOO B.OOOe+OOO 1.000e+001 1.1 00e+001 spB O.OOE+OO 2.24BE+14 1.670E+14 1.B05E+12 4.291 E+11 1.973E+12 2.692E+11 ~Source Emission on 1.B96E+11 1.23BE+06 1.006E+06 3.031 E+04 9.47BE-04 B.139E-OB 0 Energy Basis 000000 SFP/Auxilary Building:

sdef X=d1 Y=d2 z=d3 erg=dB par-2 cel=d4 ~Source Definition Card

-X-range = d 1

-Y -range = d2

-Z-range = d3

-Energy= dB

-Inside Cell 11 si1 0 2377.44 ~X-range limits sp1 0 1 ~Unifonn probability on X si2 -1341.12 6223 ~Y-range Limits sp2 0 1 ~Unifonn Probability on Y si3 -1219.201305.56 ~Z-range Limits sp3 0 1 ~Unifonn Probability on Z si4 L 11 22 33 ~Source cells: 11, 22, 33 sp4 1 1 1 ~Percent of particles from each cell c Fuel Gamma Spectra siB h 1.000e-002 5.000e-002 1.000e-001 2.000e-001 3.000e-001 4.000e-001 ~Source Energy Groups 6.000e-001 B.OOOe-001 1.000e+OOO 1.330e+OOO 1.660e+OOO 2.000e+OOO 2.500e+OOO 3.000e+OOO 4.000e+OOO 5.000e+OOO 6.500e+OOO B.OOOe+OOO 1.000e+001 1.100e+001 spB O.OOE+OO 2.24BE+14 1.670E+14 1.B05E+12 4.291 E+11 1.973E+12 2.692E+11 ~Source Emission on 1.B96E+11 1.23BE+06 1.006E+06 3.031 E+04 9.47BE-04 B.139E-OB 0 Energy Basis 000000

CALC. NO. NEE-363-CALC-002 I Fuel Handling Accident I - ENERCON Monitor Response for REV. 0 Excellence- Every projecr. Evety day.

EAL Thresholds PAGE NO. 22 of 23 7.5 MCNP Tally Specification The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-126, RE-127 and RE-128 for the RCB and RE-135 for the SFP. Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions.

The inputs to this card are the coordinates of the dose points followed by an exclusion zone (reduce variance), as well as a multiplier card, which represents the total core activity in photons/sec. The tally cards are shown in Figure 7-11 .

Figure 7-11: MCNP Tally Cards RCB:

f5c RE-126, RE-127 and RE-128 ~Tally Comment Card f5:p 152.4 -1560.96137.16 20 ~Tally 5 (point detector) 457.2 1524137.16 20 x y z exclusion

-1280.16 -944.88137.16 20 fm5 3.965E+14 ~ Tally Multiplier (Total Activity)

SFP:

f5c RE-135 ~Tally Comment Card f5:p 787.41468.12 91.44 20 ~Tally 5 (point detector) x y z exclusion fm5 3.965E+14 ~Tally Multiplier (Total Activity)

In addition, the flux is multiplied by ANSI/ANS flux-dose conversion factors per Reference 4.

This is specified in MCNP using the deldf cards . These are shown in Figure 7-12. This is the same for both the RCB and SFP/Auxiliary Building models.

Figure 7-12 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors c ------------------------------------------------------------------

c ANSI/ANS-6.1.1-1977 c Gamma Flux to Dose Conversion Factors c (mrem/hr)/(photons/cm2-s) c -----------------------------------------------------------------

deO .01 .03 .05 .07 .1 0 .15 .20 .25 .30 .35 .40 ~Energy Bins for Flux

.45 .50 .55 .60 .65 .70 .80 1. 1.4 1.8 2.2 to Dose Conversion 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5.75 6.25 6.757.59. 11 .

dfO 3.96E-03 5.82E-04 2.90E-04 2.58E-04 2.83E-04 3.79E-04 ~Energy Dependent 5.01 E-04 6.31 E-04 7.59E-04 8.78E-04 9.85E-04 1.08E-03 Flux Multipliers 1.17E-03 1.27E-03 1.36E-03 1.44E-03 1.52E-03 1.68E-03 1.98E-03 2.51 E-03 2.99E-03 3.42E-03 3.82E-03 4.01 E-03 4.41 E-03 4.83E-03 5.23E-03 5.60E-03 5.80E-03 6.01 E-03 6.37E-03 6.74E-03 7.11E-03 7.66E-03 8.77E-03 1.03E-02

CALC. NO. NEE-363-CALC-002 F. *. ENERCON

'~

Fuel Handling Accident ExceJ!ence- Every projecr. Eve1y day.

Monitor Response for REV. 0 EAL Thresholds PAGE NO. 23 of 23 7.6 MCNP Material Cards The MCNP material cards are provided in Figure 7-13. The only materials used in the model are for the air inside the mixing volume and the stainless steel liner. The SFP/Auxliary Building only has the air material card.

Figure 7-13 MCNP Material Cards m1 6012-0.000126 $ Air 7014 -0.76508 8016 -0.234793 m2 6000 -0.0008 $ ss 304 14000 -0.01 15031 -0.00045 24000 -0.19 25055 -0.02 26000 -0.68375 28000 -0.095 7.7 Results The dose rate calculated by MCNP at each detector is provided in Table 7-3. Note that dose rates are in Table 7-3 are in mrem/h. Assumption 10 is used to convert the dose rate toR/h.

Table 7-3 Dose Rate Response at RCB/SFP Detectors (mrem/h)

Location Detector Dose Rate fsd 3 Tally File RCB RE-126 5.73443E+03 0.45% mctan RCB RE-127 5.57293E+03 1.32%

RCB RE-128 5.56944E+03 0.89%

SFP RE-135 3.84319E+03 0.22% metal

8. Computer Software This calculation uses ORIGEN-S of the SCALE Version 6.1.2 code package [2] and MCNP Version 6.1 .0 [3] in accordance with CSP 3.09.

3 Fraction standard deviation .

CALC. NO. NEE-363-CALC-002 ft jl ENERCON APPENDIX A REV. 0 Excellence- Every pfO}ect. Eve1y day.

PAGE NO. 1 of 1 Appendix A- Electronic File Listing Origin output:

04/12/2017 8:45AM 40,025 PBFHAEAL4.out MCNP output:

Directory of\RCB 04/12/2017 2:25 PM 187,152 outs Directory of\SFP 04/12/2017 1:19 PM 239,474 outp

CALC. NO. NEE-363-CALC-002 ATTACHMENT 1

~A::~ ENERCON REV. 0 Excellen ce- Every project. Every day PAGE NO. 1 of 2 SFP Detector Location from Walkdown From: Kendall, Thomas Sent: Monday, February 20, 2017 9:03AM To: Jay Bhatt (jbhatt@enercon .com )

Subject:

Spent fuel pool instrument detector location

Jay, Please forward this to whomever is working it on your end.

The attached photos (most of which are very poor quality ... sorry, apparently I had a slow shutter speed) show the two brown boxes containing the detectors for the spent fuel pool area monitor.

The detectors are approximately 3' above the floor (66') level. Based on the photos and the attached drawing, the detectors are ~1' west oft he eastern edge of the checker-plate decking over the new fuel storage vault. I have shown this location on the attached drawing C-161. Note that it abuts drawing C-160 (spent fuel pool). Together, these two drawings and the photos should provide sufficient detail to construct an adequate model. I recommend that you include one or more of the photos in the final calculation.

Tom Kendall, P.E.

Principal Design Engineer Point Beach Nuclear Plant 920-755-7661 (office) 920-901-0210 (mobile)

..E1A.!i l \. E V.1IJO~ i T c.q,' . D* '

C : 5 10.*~ I'IOQll' I liff l ()olO

Z~ P~ F' U. N .

CALC. NO. NEE-363-CALC-002 ATTACHMENT 1 REV. 0 ENERCON Excellence - Every project. Every day.

PAGE NO.

CALC. NO. NEE-363-CALC-002 ATTACHMENT 2

,_'jJ ENERCON REV. 0 Exceflence - Every project. Eve1y day.

PAGE NO. 1 of 1 Western Edge of SFP Marked in red is the western edge of the SFP at the 66'-0" EL. as shown on C-160 and C-161 . This is used as a boundary throughout the SFP model.

C-160 :

y .

T

c--J, .

I**

\

'o C-162:

CALC NEE-363-CALC-002 ATTACHMENT 3 NO.

~ :. CALCULATION PREPARATION IJ ENERCON CHECKLIST E>:cellence-Every project. Every day.

REV. 0 CHECKLIST ITEMS1 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? D D ~

The Calculation is performed in accordance with ENERCON procedures.

2. Are the proper forms being used and are they the latest revision? ~ D D The Calculation is performed in accordance with ENERCON procedures.

3. Have the appropriate client review forms/checklists been completed? D D ~

OAR will be performed ather calculation submittal.

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure? ~ D D

5. Is all information legible and reproducible? ~ D D

6. Is the calculation presented in a logical and orderly manner? ~ D D

7. Is there an existing calculation that should be revised or voided? D ~ D There is no existing calculation that should be revised or voided.

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D ~ D No existing calculation would be applicable.

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 ~ D D apply to the calculation revision being performed.

An existing FHA calculation CN-CRA-08-14 is used for some design inputs. The relevant design inputs , assuptions and engineering judgements used in that calculation are valid to NEE-363-CALC-002.

10. Is the format of the calculation consistent with applicable procedures and expectations? ~ D D 11 . Were design input/output documents properly updated to reference this calculation? D D ~

There are no design output documents.

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

13. Does the calculation provide a clear concise statement of the problem and objective

~ D D of the calculation?

14. Does the calculation provide a clear statement of quality classification? ~ D D

15. Is the reason for performing and the end use of the calculation understood? ~ D D

16. Does the calculation provide the basis for information found in the plant's license basis?

D ~ D This does not provide a basis for license basis Page 1 of5

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~ jj CALCULATION PREPARATION

' or ENERCON CHECKLIST Excellence- Every projecr. Every doy.

REV. 0 CHECKLIST ITEMS1 YES NO N/A

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

D D fZJ This does not provide a basis for license basis

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D D fZJ This calculation does not provide basis for design basis.

19. If so, is this documented in the calculation? D D fZJ This calculation does not provide basis for design basis.

20. Does the calculation otheJWise support information found in the plant's design basis D fZJ D documentation? I This does not provide support for information found in the design basis documentation.

21. If so, is this documented in the calculation? I D I D fZJ This does not provide support for information found in the design basis documentation.

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DESIGN INPUTS I I I 23 . Are design inputs clearly identified? I fZJ I D I D

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33. Does the calculation reference the latest revision of the design input source? I fZJ I D I D Page 2 of5

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CHECKLIST REV. 0 CHECKLIST ITEMS 1 YES NO N/A

34. Were all applicable plant operating modes considered? 1:8] 0 0 ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? 1:8] 0 0

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37. Are any engineering judgments used?

0 1:8] 0 Engineering judgments is not used as a design input.

38. Are engineering judgments clearly identified as such? 0 0 1:8]

Engineering Judgement is not used as a design input.

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, 0 0 1:8]

engineering principles, physical laws or other appropriate criteria?

Engineering Judgement is not used as a design input.

METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's 1:8]

licensing basis?

0 0 This calculations scope is outside the plant licensing basis.

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

0 0 1:8]

This calculations scope is outside the plant licensing basis.

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1:8] 0 0 BODY OF CALCULATION

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1:8] 0 0

45. Is there reasonable justification provided for the use of equations not in common use?

0 0 1:8]

There are no uncommon equations used in the calculation .

46. Are the mathematical operations performed properly and documented in a logical fashion?

1:8] 0 0

47. Is the math performed correctly? 1:8] 0 0 48 . Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied?

1:8] 0 0

49. Has proper consideration been given to results that may be overly sensitive to very small changes in input?

1:8] 0 0 Page 3 of 5

CALC NEE-363-CALC-002 ATTACHMENT 3 NO.

,'jJ ENERCON CALCULATION PREPARATION II CHECKLIST Excellence- Every project. Every day.

REV. 0 CHECKLIST ITEMS1 YES NO N/A SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation?

~ D D MCNP and Scale are used

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met?

~ D D

52. Are the codes properly identified along with source vendor, organization, and revision level?

~ D D

53. Is the computer code applicable for the analysis being performed? ~ D D

54. If applicable , does the computer model adequately consider actual plant conditions? I ~ D D

55. Are the inputs to the computer code clearly identified and consistent with the inputs

~ D D and assumptions documented in the calculation? I I

56. Is the computer output clearly identified? I ~ I D D

57. Does the computer output clearly identify the appropriate units? I ~ I D I D

58. Are the computer outputs reasonable when compared to the inputs and what was expected? I ~

I D I D

59. Was the computer output reviewed for ERROR or WARNING messages that could

~ D D invalidate the results? I I I RESULTS AND CONCLUSIONS I I I

60. Is adequate acceptance criteria specified? I D I D I ~

This calculations scope is outside the plant licensing basis.

61 . Are the stated acceptance criteria consistent with the purpose of the calculation , and D D ~

intended use? I I I This calculations scope is outside the plant licensing basis .

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable D D I ~

licensing commitments and industry codes, and standards? I I This calculations scope is outside the plant licensing basis.

63. Do the calculation results and conclusions meet the stated acceptance criteria? I D I D I ~

This calculations scope is outside the plant licensing basis.

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

~

I D

I D Page 4 of 5

CALC NEE-363-CALC-002 ATTACHMENT 3 NO.

~J

. EN E R C ON CALCULATION PREPARATION b*aJh:nrf' - *hNy prc,JeCI. Every d,-1*.

CHECKLIST REV. 0 CHECKLIST ITEMS 1 I YES I NO I N/A

65. Are the calculation results and conclusions reasonable when considered against the 0 0 D stated inputs and objectives? I I I

66. Is sufficient conseNatism applied to the outputs and conclusions? I 0 I 0 I D

67. Do the calculation results and conclusions affect any other calculations? I 0 I 0 I D No other calculations are affected by this calculation.

68. If so, have the affected calculations been revised? j D I 0 I 0 No other calculations are affected by this calculation.

69. Does the calculation contain any conceptual, unconfirmed or open assumptions 0 0 0 requiring later confinnation? I I I There are no open assumptions requiring confirmation later.

70. If so, are they properly identified? I D I 0 I 0 There are no open assumptions requiring confinnation later.

DESIGN REVIEW I I I 71 . Have alternate calculation methods been used to verify calculation results? I D I 0 I 0 No a Design Review was peliormed.

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: Ryan Skaggs Print Name and Sign Date Page 5 of5

Design Inform ati on Tran smittal From: Point Beach Design Engineering To : ENERCON Document/ EC/ Tracking Date: 4/10/2017 DIT No: n/a EC 0288568 Number:

Document

Title:

Calculations NEE-363-CALC-001 (Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds) and NEE-363-CALC-001 (Fuel Handling Accident Monitor Reseonse for EAL Thresholds}

Quality Facility/ Unit: PBNP/0 QR Classification

SUBJECT:

This Design Information Transmittal (DIT) is to provide ENERCON with requested inputs for use in calculating EAL thresholds.

Check if applicable:

D This DIT confirms information previously transmitted on N/A under DIT-001 Draft Attachment 1 and 2.

Additional information has been added since DIT-001 Draft was sent.

D This information is preliminary. See explanation below.

SOURCE OF INFORMATION:

All sources documented in attached text.

DESCRIPTION OF INFORMATION:

Various source terms , geometries, dimensions, etc. of the fuel, reactor vessel, and core internals. Details provided in the attached text.

To ensure complete documentation, this DIT shall be included with any QA documents (e.g. calculations) citing it as a source of information.

DISTRIBUTION (Recipients should receive all attachments unless otherwise indicated. All attachments are uncontrolled unless otherwise ind icated)

Jay Bhatt (ENERCON)

Ryan Skaggs (ENERCON)

PREPARED BY (The Preparer and Approver may be the same person.)

Tom Kendall Principal Engineer

'~/

.-*/; _. ,_.- - -

ttfi¢/tt Preparer Name Position Signature Date VERIFIED BY (Design verification is required if the information is not a verified design output. Verification is also required if the information is developed , interpreted, or extracted from an unverified source. Otherwise, N/A).

/ f/

Steve Bach Senior Engineer

~~AL12vf~A w.1~/L ~

Verifier Name Position Signature Date APPROVED BY (The cognizant Engineering Supervisor has release authority. Consult the Design Interface Agreement or local procedures to determine who else has release authority.)

Civ/Mech Design Jane Marean

.Eng . Sueervisor

(\~ ~~~ '-(*1 f . i f Approver Name Position Si nature\ Date A copy of the DIT (along with any attachments not on file) should be included with the associated EC or document record.

EN-AA-1 00-1 003-F02 Revision 0 Page 1 of 1

8 Containment High Range gamma detectors read out in R/hr, and have a range of 1 R/hr to 10 R/hr (source: Vendor Technical Manual 978, General Atomics)

Parameters Value Unit Reference*

Fuel Type Westinghouse 14x14 Table 2-1

Fuel Rods per Assy 179 Table 2-1 Assembly Array 14x14 Table 2-1 Assembly Width 7.761 [in] Table 2-1 0.96 Eval 2016-0005 Density{% of theoretical STD (rounded to 2 Fuel) significant digits)

Fuel Pellet OD 0.3659 [in] Table 2-1 Fuel Rod OD 0.422 [in] Table 2-1 Clad Thickness 0.0243 [in] Table 2-1 Active Length 143.25 [in] Table 2-1 Diameter of Active Fuel 96.3 [in] Derived**

RPV Thickness (including clad) 6.656 [in] Dwg. 117802E Liner Thickness 1/4 [in] Drawing C-123

Unless otherwise noted above, all data in this table is from the Reload Transition Safety Report for the Point Beach Units 1 and 2 Fuel Upgrade to 422 Vantage Plus Fuel ("RTSR"), Revision 2, filed as Report NPC 1999-05997.

- There are 121 fuel assemblies in the core (Eval 2016-0005). Each assembly is 7.761 inches square per the above table:

A = 121 x (7.761 in) 2 = 7,288 in 2 The equivalent diameter is therefore:

D eq

~ = 96.3 in

= 2~------;----

The vertical distance from the top of the active fuel to the centerline of the reactor outlet nozzles is approximately SO" [1189E025 Rev 0 and 6669E93 Rev 1].

Table 5-1 Design Basis Core Shutdown Source Term (Calculation CN-CRA-08-21 Revision 1)

Isotope Ci Isotope Ci Kr-85 6.15E+05 Te-127 4.54E+06 Ce-143 8.03E+07 Kr-85m 1.36E+07 Ba-139 9.42E+07 Ce-144 6.72E+07 Kr-87 2.68E+07 Ba-140 9.05E+07 Pu-238 1.33E+05 Kr-88 3.60E+07 Te-127m 7.48E+05 Pu-239 1.45E+04 Xe-131m 5.55E+05 Te-129 1.33E+07 Pu-240 2.25E+04 Xe-133 1.02E+08 Te-129m 2.52E+06 Pu-241 5.73E+06 Xe-133m 3.21E+06 Te-131m 9.95E+06 Np-239 9.65E+08 Xe-135 2.17E+07 Te-132 7.30E+07 Y-90 5.01E+06 Xe-135m 2.20E+07 Sb-127 4.63E+06 Y-91 6.56E+07 Xe-138 9.05E+07 Sb-129 1.42E+07 Y-92 6.82E+07 1-130 1.05E+06 Sr-89 5.03E+07 Y-93 7.67E+07 1-131 5.10E+07 Sr-90 4.80E+06 Nb-95 8.87E+07 1-132 7.47E+07 Sr-91 6.30E+07 Zr-95 8.76E+07 1-133 1.06E+08 Sr-92 6.73E+07 Zr-97 8.80E+07 1-134 1.19E+08 Ru-103 7.79E+07 La-140 9.69E+07 1-135 1.01E+08 Ru-105 5.42E+07 La-142 8.25E+07 Rb-86 9.95E+04 Ru-106 2.54E+07 Pr-143 7.75E+07 Cs-134 9.52E+06 Rh-105 5.08E+07 Nd-147 3.33E+07 Cs-136 2.14E+06 Te-99m 8.47E+07 Am -241 6.16E+03 Cs-137 6.27E+06 Mo-99 9.62E+07 Cm-242 1.70E+06 Cs-138 9.89E+07 Ce-141 8.52E+07 Cm-244 1.58E+05

Upper Internals/Upper Fuel Hardware To approximate volume of metallic reactor internals is needed to estimate the shielding such components provide. A review of available documents found that calculation N-93-015 performed a detailed evaluation of the water displacement volume of such components. Calculation N-93-015 has now been superseded by another calculation that does not provide a detailed accounting of the reactor internal metal volumes. However, based on a review of the contents of N-93-015, it is apparent that the estimates of metallic component volumes remains valid and appropriate for use in establishing EAL initiating conditions.

Calculation N-93-015 Revision 0 lists the following volumes:

Volume of steel from the top of the upper core plate to % pipe is 21.019 ft 3 (pg 3I)

Volume of steel from% pipe to I ft below the reactor vessel flange is 41.837 ft 3 (pg 3I)

Volume of the Upper Core Plate is 4.797 ft 3 (pg 26).

Volume of the Upper Support Plate is I.30083 ft 3/inch x 4"=5.2 ft 3 (pg 30)

Decay time for Fuel Handling Accident I) The latest time after reactor shutdown that we typically are handling fuel (i.e. end of the core re-load evolution). The basis for this value is a review of recently scheduled and/or completed refueling outages.

a. Unit 2 35th refueling:

i. Rx shutdown at 0034 on 3/18/20I7 (source: Station Logs) ii. Core reloaded by I931 on 4/4/2017 (source: Station Logs) iii. Duration: 403 hours0.00466 days 0.112 hours 6.66336e-4 weeks 1.533415e-4 months

b. Unit I 36th refueling:

i. Rx shutdown at OI07 on 3/I2/20I6 (source: Station Logs) ii. Core reloaded by 0404 on 3/26/20I6 (source: Station Logs) iii. Duration: 339 hours0.00392 days 0.0942 hours 5.605159e-4 weeks 1.289895e-4 months

c. Unit 2 34th refueling

i. Rx shutdown at 0048 on I0/3/20I5 (source: Station Logs) ii. Core reloaded by 04I4 on IO/I8/20I5 (source: Station Logs) iii. Duration: 364 hours0.00421 days 0.101 hours 6.018518e-4 weeks 1.38502e-4 months The maximum duration from these three outages was therefore 403 hours0.00466 days 0.112 hours 6.66336e-4 weeks 1.533415e-4 months . Rounding up to nearest full day (17 days) gives 408 hours0.00472 days 0.113 hours 6.746032e-4 weeks 1.55244e-4 months . This is judged to be an adequate approximation for the purposes of establishing the EAL IC.

2) The earliest time to statt fuel movement. As with the latest time to complete fuel handling, this is based on a review of recently scheduled or completed refueling outages.

a. Unit 2 35th refueling (scheduled to start 3/18/2017):

i. Rx shutdown at 0034 on 3/18/2017 (source : Station Logs) ii. Core off-load started at 2031 on 3/25/17 (source: Station Logs) iii. Duration: 188 hours0.00218 days 0.0522 hours 3.108466e-4 weeks 7.1534e-5 months

b. Unit 136th refueling:

i. Rx shutdown at 0107 on 3/12/2016 (source: Station Logs)

ii. Permission to offload the core logged at 1214 on 3/17/2016 (source: Station Logs) iii. Duration: 131 hours0.00152 days 0.0364 hours 2.166005e-4 weeks 4.98455e-5 months

c. Unit 2 34th refueling

i. Rx shutdown at 0048 on 10/3/2015 (source: Station Logs) ii. Permission to offload the core logged at 1010 on 10/12/2015 (source: Station Logs) iii. Duration: 225 hours0.0026 days 0.0625 hours 3.720238e-4 weeks 8.56125e-5 months The minimum duration was 131 hours0.00152 days 0.0364 hours 2.166005e-4 weeks 4.98455e-5 months . Rounding down to the nearest full day (5 days) gives 121 hours0.0014 days 0.0336 hours 2.000661e-4 weeks 4.60405e-5 months . This is judged to be an adequate approximation for the purposes of establishing the EAL IC, though shorter durations (e .g. for head lift) may also be acceptable.

Fuel Assembly Activity from Calculation CN-CRA-08-14 Revision 0 (Table 4.6.1.4.1-2):

Nuclide Core Activity Activity of 1 (121 assys) average fuel assy (tiJ 65h (Ci) (tiJ 65h (Ci) 1-130 2.77E+04 2.29E+02 1-131 4.15E+07 3 .34E+05 1-132 4.23E+07 3.50E+05 1-133 1.25E+07 1.03E+05 1-135 1.06E+05 8.76E+02 Kr-85m 5.92E+02 4.89E+OO Kr-85 6.14E+05 5.07E+03 Kr-87 1.13E-08 9.34E-11 Kr-88 4.65E+00 3.84E-02 Xe-131m 5.47E+05 4.52E+03 Xe-133m 2.00E+06 1.65E+04 Xe-133 8.41E+07 6.95E+05 Xe-135m 1.73E+04 1.43E+02 Xe-135 1.75E+06 1.45E+04 Worst Case Radial Peaking Factor from Calculation CN-CRA-08-14 Revision 0 (Table 4.6.1.4.1-2):

RPF = 1.7 Rods per Assembly and Total Rods from Reload Transition Safety Report for the Point Beach Units 1 and 2 Fuel Upgrade to 422 Vantage Plus Fuel ("RTSR"), Revision 2, filed as Report NPC 1999-05997:

Rods per Assembly 179 Total Fuel Rods

ENCLOSURE 3 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENTAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, *

"DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" CALCULATION 2013-0018, REVISION 3 RADIOLOGICAL EFFLUENT INITIATING CONDITION VALUES FOR EMERGENCY ACTION LEVEL RS1 AND RG1 25 pages follow

CALCULATION COVER SHEET (Page 1 of 1}

Document Information:

Calculation (Doc) No: 2013-00 18 IControlled Documents Revision: 3

Title:

RADIOLOGICAL EFFLUENT rNITIATING CONDITION VALUES FOR EMERGENCY ACTION LEVEL RS 1 AND RG 1 Type: CALC Sub-Type: CALC Discipline: Radiological Facility: PBNP I Unit: 0 Safety Class: D SR ~ Quality Related 0 Non-Nuclear Safety D Important to Safety 0 Not Important to Safety Special Codes: 0 Safeguards 0 Proprietary Vendor Doc No: N/A IVendor Name or Code:N/A Executive Summary (optional): Correct typographical errors for RE-232 8 train (consistently listed as "A" in three different tables), and insert two lines in the table on page 19 for a General Emergency for RE-307 that were inadvertently omitted in Revision

2. Errors identified in AR 022 I2948 Review and Approval:

Associated EC Number: 289448 EC Revision: 0 ARJ Other Document Number: 02212948 Description of Calculation Rev~ ion: Update EC Document Revision: 3 Prepared by: --i N~rf I T. C. Kendall Date: 7/17/2017 (signature) __..... (print name)

Reviewed by: t1k_ ~.Jk:. _... Carl Onesti Date: "'1 / Z.L It 7 (signature) (print name)

Type of Review: [g) Design Verification D Review D Owner Acceptance Review Method Used (Por DV Only): 0 Design Review D Alternate Calculation Approved by: U~)dl., ,, ___ ~9~.,.-~v\ Date: ~7 j 17 (sign~ture) \ (print name)

EN-AA-100-1004-F01, Revision 0

CALCULATION REVISION

SUMMARY

SHEET (Page 1 of 1)

Calculation Number: 2013-0018 Rev. Affected Paqes Reason for Revision 0 All (1-21) Initial Issue I All Update header, footer 2 Delete unused references 16 Correct typographical error in table for ADV releases 2 All (cover sheet, Update header, footer, delete unused references, update revision history, pages revised references, expand to include Unusual Event and 1-23): Alert EAL ICs, delete 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months release calculations, incorporate revised ODCM value for X/Q, include Alert EAL IC for both Revision 4 and Revision 6 ofNEI 99-01.

3 All Update headers, cover sheet with new revision Page 19 Correct typographical error for RE-232 train B in the table Page 19 Add lines for RE-307 in the table that were inadvertently omitted in the previous revision Page 20 Correct typographical error for RE-232 train B in the table Page 21 Correct typographical error for RE-232 train B in the table Pgs 2, 4, 7, 8, 10-14, Restore proper links (pointers) to input. Links to the 16, and 17 sequentially numbered inputs had been off by 1 in the previous revision.

EN-AA-100-1004-F02, Revision 1

Calculation 2013-0018 Revision 3

Purpose:

The purpose of this calculation is to calculate effluent monitor readings that correspond to the initiating condition (I C) values for emergency action level (EAL) RU I, RA 1, RS 1 and RG 1. This calculation is valid for Unit 1 and Unit 2 extended power uprate conditions (1800 MWth).

Revision 2 of this calculation expands the previous scope to include the bases for RUI and RAJ, and includes a bases for these EAL ICs as determined using the guidance ofNEI 99-0 I Revisions 4 and Revision 6. The guidance methodology for determining RS 1 and RG I is the same in both revisions of the NEI guidance. In addition, the value of x/Q in this calculation has been replaced with that derived and used by the ODCM to be consistent with the NEI 99-01 guidance. Lastly, the Unit 1 containment purge exhaust values eliminated, and the Unit 2 values have been revised based on a much lower flow-rate that reflects operation of the letdown gas stripper building. The containment purge exhaust lines are disconnected and blanked off during operation above MODE 5.

Methodology:

The methodology of this calculation is described in NEI 99-01. Two different revisions of this industry guidance will be used in anticipation oft:ransitioning the station license bases from the current Revision 4 to Revision 6 at a later date. The approved guidance differs between the two revisions only for ICs RUl and RAl.

There are no acceptance criteria for this calculation.

Assumptions:

Validated Assumptions:

A 1. It is assumed that the source term can be represented by the gaseous release data contained in the Point Beach Nuclear Plant Final Safety Analysis Repmi (FSAR) Table I. 7-1, Point Beach Nuclear Plant Calculated Total Annual Gaseous Releases (Cilyr).

Basis: Reference R14, Section 6, Abnormal Rad Levels I Radiological Effluent ICs I EALs.

While the Developer notes state that Offsite Dose Calculation Manual (ODCM) methodology should be used, the expanded guidance contained in Appendix A.7 acknowledges that the ODCM does not contain source terms for halogens (i.e. Thyroid). As a result, it states that "Other source terms may be appropriate".

A2. It is assumed that the release duration is one hour.

Basis: While Reference R18 establishes a four hour release duration when performing dose assessments, this is not applicable for pre-calculated EAL initiating condition set points. NEI 99-01 (Reference R14) developer notes for ICs AS 1 and AG 1 state that "the monitor readings should correspond to a dose ... for one hour of exposure".

Page 1 of 23

Calculation 2013-0018 Revision 3 A3. All releases via the ABV, DAY, containment purge vent, atmospheric dump valve (ADV), and main steam safety valve (MSSV) are classified as a release through a vent or other building penetration and not a stack release.

Basis:These release points are all effectively lower than two and one-halftimes the height of adjacent solid structures. The release height for the ABV, DA V, and containment purge vent is Elevation 168 ' and the height of the adjacent solid structure is Elevation 161 '6". The release height for the ADV and MSSV is Elevation 170' and the height of the adjacent solid structure is Elevation I 61 ' 6". (Reference Rl 7, Inputs 119 and 120)

Page 2 of 23

Calculation 2013-0018 Revision 3

References:

Rl. Point Beach Nuclear Plant Offsite Dose Calculation Manual (ODCM) Revision 19 R2 . Calculation CN-SEE-III-08-10, "Point Beach Units 1 and 2 RHRS Cooldown Analysis for EPU to 1806 MWT NSSS Power," Revision 5 R3. Calculation N-89-009, "Decay Heat Rate Curve," Revision 2 R4. Copes-Vulcan Drawing D-350250, "6 Inch Class 600," Revision 2 R5. Drawing C-28, "Roof Plan," Revision 9 R6. Drawing M-75, "Area-6 Auxiliary Building Miscellaneous Sections," Revision 8 R7. EOP-0 Unit 1 (Rev. 64) and Unit 2 (Rev. 63), "Reactor Trip or Safety Injection,"

R8. EOP-2 Unit 1 (Rev. 25) and Unit 2 (Rev. 25), "Faulted Steam Generator Isolation,"

R9. EPA 400-R-92-001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," October 1991 RIO. EPA-520/1-88-020, "Limiting Values ofRadionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," September 1988 R11. HPCAL 3.4, "SPING Calibration," Revision 37, 9/30/2014 Rl2. HPCAL 3.7, "Steam Line Radiation Monitor Calibration," Revision 21, 8/23/2016 R13. NAMS Equipment Data for MS-02010, HX-1A SG Header Safety, and MS-02005, HX-1B SG Header Safety.

R14. NEI 99-01 [Revisions 4 and 6], "Development of Emergency Action Levels for Non-Passive Reactors," Specific revision *will be cited when used.

R15. Point Beach Nuclear Plant Final Safety Analysis Repmi (FSAR)

R16. Steam Tables, "Prope1iies of Saturated and Superheated Steam- From 0.08865 to 15,500 Lb per Sq In. Absolute Pressure," Combustion Engineering R17. U.S. Nuclear Regulatory Commission Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Revision 1, November 1982 (Reissued February 1983).

R18. NPM 2006-0327, "4 Hour Unknown Estimated Release Duration for Dose Assessment," June 27,2006 Page 3 of 23

Calculation 2013-00 18 Revision 3 II. The activity fraction of each radionuclide in the effluent pathway for the Auxiliary Building Vent (ABV), the Drumming Area Vent (DAY), the Containment Purge Exhaust, and the Steam Line Atmospheric I Safety Exhaust are listed below. The fractions were calculated by first summing the total noble gas, iodine, and pat1iculate activities to obtain a total annual activity release and then by dividing the individual isotopic activities by the total annual activity release to obtain the dimensionless fraction . (Reference R 15, Table I. 7-1, "Point Beach Nuclear Plant Calculated Total Annual Gaseous Releases (Ci/yr)") Data below does not include H-3, C-14, Sr-89, and Sr-90 because they are weak beta emitters and are not detected by the applicable radiation monitor in the vent pathway. (Input I21) 12.

Steam Line ABV/DAV Containment Safety/ Atmospheric Nuclide Activity Purge Activity Activity (J.Lci/cc)* (mci/cc)**

(mcilcc)***

Ar-41 - 2.5E+01 -

Kr-83m 7.2E-01 5.4E-02 1.2E-04 Kr-85m 3.0E+OO 4.1E-01 S.OE-04 Kr-85 2.6E+OO 1.7E+Ol 4.2E-04 Kr-87 2.2E+OO 1.3E-01 3.4E-04 Kr-88 6.4E+OO 6.3E-01 l.OE-03 Kr-89 2.0E-01 9.8E-04 3.4E-05 Xe-131m 1.8E-01 7.3E-01 3.0E-05 Xe-133m 1.4E+00 1.6E+OO 2.2E-04 Xe-133 5.6E+01 1.4E+02 8.8E-03 Xe-135m 5.6E-01 l.OE-02 8.6E-05 Xe-135 7.2E+OO 1.7E+00 1.2E-03 Xe-137 3.8E-01 2.1E-03 6.0E-05 Xe-138 1.8E+OO 3.3E-02 2.8E-04 I-131 7.4E-02 3.1E-03 3.0E-03 I-133 l.OE-01 4.9E-04 4.2E-03 Co-58 1.2E-03 7.5E-04 -

Co-60 5.4E-04 3.4E-04 -

Mn-54 3.6E-04 2.2E-04 -

Fe-59 1.2E-04 7.5E-05 -

Cs-134 3.6E-04 2.2E-04 -

Cs-137 6.0E-04 3.8E-04 -

Total activity 8.3E+01 1.9E+02 2.0E-02 Page 4 of 23

Calculation 2013-0018 Revision 3 Nuclide ABV/DAV Containment Purge Steam Line Activity Fraction* Activity Fraction** Safety/ Atmospheric Activity Fraction***

Ar-41 - 1.3E-O 1 -

Kr-83m 8.7E-03 2.9E-04 6.1 E-03 Kr-85m 3.6E-02 2.2E-03 2.5E-02 Kr-85 3.1E-02 9.1E-02 2.1E-02 Kr-87 2.7E-02 6.9E-04 1.7E-02 Kr-88 7.7E-02 3.4E-03 4.9E-02 Kr-89 2.4E-03 5.2E-06 1.7E-03 Xe-131m 2.2E-03 3.9E-03 1.5E-03 Xe-133m 1.7E-02 8.5E-03 1.1E-02 Xe-133 6.8E-01 7.5E-01 4.3E-01 Xe-135m 6.8E-03 5.3E-05 4.2E-03 Xe-135 8.7E-02 9.1E-03 5.9E-02 Xe-137 4.6E-03 l.lE-05 3.0E-03 Xe-138 2.2E-02 1.8E-04 1.4E-02 I-131 8.9E-04 1.7E-05 1.5E-01 1-133 1.2E-03 2.6E-06 2.1E-01 Co-58 1.4E-05 4.0E-06 -

Co-60 6.5E-06 1.8E-06 -

Mn-54 4.3E-06 1.2E-06 -

Fe-59 1.4E-06 4.0E-07 -

Cs-134 4.3E-06 1.2E-06 -

Cs-137 7.2E-06 2.0E-06 -

Per Table 1.7-1, the DA V activity fraction is the same as the ABV activity fi*action because exit velocities and locations are essentially identical.

- Per Table 1.7-1 , the Unit 1 and Unit 2 activity fractions are the same.

- - From Table 1.7-1 , the Turbine Building Ventilation activity fraction data is used for the steam line. Tllis source term is for steam releases from the secondary system.

13. The secondary side water volume of a single steam generator at zero power is 2,877 fe for Unit 1 and 2,704 ftl for Unit 2. (Reference R15, Table 4.1-4, Steam Generator Design Data)

14. The hot zero power temperature is 547 °F. (Reference R15, Table 14.0-1 , "Summary ofinitial Conditions and Computer Codes Used")

15. The specific volume for saturated liquid at 547 °F is 0.0216 fe/lb 01 * (Reference RI6, Table 1, "Saturated Steam: Temperature Table")

16. The reactor coolant system heat capacity is 1E+06 BTU/°F. (Reference R2)

Page 5 of 23

Calculation 2013-0018 Revision 3 I7. The atmospheric dump valve (ADV) flow capacity at 1,085 psig is 333,200 Ibm/hr. (Reference R4)

I8. The main steam safety valve (MSSV) flow capacity is 817,000 Ibm/hr. (Reference R13)

I9. The lowest MSSV pressure setpoint is 1,085 psig. (Reference R13)

IlO. The upper bound decay heat rate as a function of time after shutdown for the Point Beach Nuclear Plant at 1800 MWth is listed below. (Reference R3)

Time (sec) Time increment (sec) Q'decav (Btu/lu:)

I 1 4.2E+08 1.5 0.5 4.0E+08 2 0.5 3.9E+08 4 2 3.7E+08 6 2 3.5E+08 8 2 3.2E+08 10 2 3.1E+08 15 5 2.9E+08 20 5 2.8E+08 40 20 2.4E+08 60 20 2.3E+08 80 20 2.1E+08 100 20 2.0E+08 150 50 1.9E+08 200 50 1.8E+08 400 200 1.5E+08 600 200 1.4E+08 800 200 1.3E+08 1000 200 1.2E+08 1500 500 1.1E+08 2000 500 l.OE+08 3600 1600 8.6E+07*

4000 400 8.3E+07

Linearly interpolated from two adjacent values Ill . The enthalpy for evaporation at 54 7 op is 64 7 Btu/Ibm and at 200 op is 978 Btu/Ibm. (Reference Rl6, Table 1, "Saturated Steam: Temperature Table")

Page 6 of 23

Calculation 2013-0018 Revision 3 112. The specific volume of steam at 1085 psig is 0.4006 fe!lbm. (Input 19; Reference R16, Table 1, "Saturated Steam: Temperature Table")

113. The discharge flow rates for the auxiliary building vent (ABV) and the drumming area vent (DA V) are 66,400 CFM and 43, I 00 CFM. The containment purge exhaust ventilation penetrations are blanked off during unit operation, and therefore releases via this pathway are not valid. However, If the gas stripper building fans are operating, then they discharge 13,000 cfrn to the Unit 2 containment purge/vent release path (Reference R I, Table 10-1, "Gaseous Effluent Pathways").

114. The dose conversion factor for combined exposure pathways (DCFwa) during the early phase of a nuclear incident are listed below. (Reference R9, Table 5-1 for all except Ar-41 and Kr-83m; Reference R1 0, Table 2.3 for Ar-41 and Kr-83m)

DCF Nuclide (rem - cm3) I (J.!Ci - h)

Ar-41 8.00E+02 Kr-83m 1.50E-02 Kr-85m 9.30E+01 Kr-85 1.30E+OO Kr-87 5.10E+02 Kr-88 1.30E+03 Kr-89 1.20E+03 Xe-131m 4.90E+OO Xe-133m 1.70E+01 Xe-135m 2.50E+02 Xe-133 2.00E+01 Xe-135 1.40E+02 Xe-137 l.IOE+02 Xe-138 7.20E+02 1-131 5.30E+04 1-133 1.50E+04 Co-58 1.70E+04 Co-60 2.70E+05 Mn-54 1.20E+04 Fe-59 2.30E+04 Cs-134 6.30E+04 Cs-137 4.10E+04 Page 7 of 23

Calculation 2013-0018 Revision 3 Il5. The dose conversion factors for dose equivalent to the thyroid (DCFTHv) from inhalation of radioiodine are listed below. (Reference R9, Table 5-2)

DCF Nuclide (rem - cm 3 ) I (!lCi - h) 1-131 1.3E+06 1-133 2.2E+05 I 16. The AS 1 initiating condition is offsite dose resulting in an actual or imminent release of gaseous radioactivity exceeding I 00 mrem TEDE or 500 rnrem Thyroid CDE for the actual or projected duration of the release. (Reference R14)

Il7. The AG 1 initiating condition is offsite dose resulting in an actual or imminent release of gaseous radioactivity exceeding 1000 mrem TEDE or 5000 mrem Thyroid CDE for the actual or projected duration of the release. (Reference R14)

I 18. The site x/Q value for the all calculations, consistent with that used in the ODCM, is 1.09E-6 seclm3 (Reference Rl Table 10-2, and Reference R14 Revisions 4 and 6 AU!, AAI, AS!, and AGI)

Il9. The release height for the DA V, ABV, and Containment purge vent, and the maximum adjacent building height either upwind or downwind from these vents are listed below. (Reference R15, Table 1.2-4)

DAV release height elevation= 168' ABV release height elevation = 168' Containment purge vent release height elevation= 168' Maximum adjacent building (Facade) height elevation= 161' 6" I20. The release height for the Steam Line Atmospheric I Safety Exhaust vents and the maximum adjacent building height either upwind or downwind from these vents are listed below.

(References R5 and R6, Input Il9)

ADV release height elevation= 170' MSSV release height elevation= 170' Maximum adjacent building (Facade) height elevation= 161' 6" Page 8 of 23

Calculation 2013-0018 Revision 3 121. The detector type and calibration source radionuclide for each of the monitors used to provide readings for the AS 1.1 and AG 1.1 IC ' s is listed below.

Monitor Detector Type 2RE-307, Containment Purge Exhaust [Letdown GM Tube (Reference R15 , Table 11.5-3)

Gas Stripper Building] Mid-Range Gas Cs-137 (Reference Rl1) 2RE-309, Containment Purge Exhaust [Letdown GM Tube (Reference R 15, Table 11.5-3)

Gas Stripper Building] High Range Gas Kr-85 (Reference R11)

RE-317, Auxiliary Building Exhaust Mid-Range GM Tube (Reference R15, Table 11.5-3)

Gas Cs-137 (Reference R11)

RE-319, Auxiliary Building Exhaust High Range GM Tube (Reference R15, Table 11.5-3)

Gas Kr-85 (Reference Rll)

RE-327, Drumming Area Exhaust Mid-Range Gas GM Tube (Reference Rl5, Table 11.5-3)

Cs-137 (Reference Rll) 1(2) RE-231, Steam Line 1A(2A) GM Tube (Reference Rl5 , Table 11.5-2B)

Cs-137 (Reference Rl2) 1(2)RE-232, Steam Line 1B(2B) GM Tube (Reference Rl5, Table 11.5-2B)

Cs-137 (Reference R12) 122. Reference Rl Table 10-1 provides the default monitored release path alarm setpoints:

GASEOUS EFFLUENT MONITORS DISCHARGE CALCULATED PATHWAY FLOW RATE DEFAULT (cfm) SETPOINT (J.lCi/cc)

1. Auxiliary Building Vent RE-214 & SPING 23 66,400 6.75E-04

2. Combined Air Ejector RE-225 20 2.24E+OO

3. Unit Air Ejector 1(2) RE-215 10 1.79E+01

4. Containment Purge/Vent Unit 1 1RE-212 & SPING 21 25,000* 7.17E-03 Unit 2 2RE-212 & SPING 22 38,000** 4.72E-03 Unit 1{2) 1(2) RE-212 35 * ** 5.12E+00

5. Gas Stripper Building RE-224 13,000 3.45E-03

6. Drumming Area Vent RE-221 & SPING 24 43,100 1.04E-03

Containment Purge alone**

- Containment Purge together with GS building

- - Force Vent in operation without containment purge or GS building Page 9 of 23

Calculation 2013-0018 Revision 3 Calculation:

Pmi 1: Steam line release flow rate The basis for limiting the steam line release flow rate to the equivalent of one main steam safety valve is explained here. In the event of a faulted steam generator, the emergency operating procedures (EOPs) will direct the Operator to isolate the affected steam generator. (Reference R8, step 4) Isolation involves the closure of the following valves: main steam isolation, main feedwater isolation, auxiliary feedwater (AFW) isolation, standby steam generator isolation, atmospheric steam dump, steam generator blowdown isolation, turbine driven AFW pump steam supply, and radwaste system steam supply fi*om the affected SG. (Reference R8)

The maximum steam generator secondary side liquid mass at hot zero power is 133,194 Ibm. (Inputs 13, I4, and I5) This is less than half of the hourly capacity of a single atmospheric steam dump valve at full lift (Input 17), and less than 20% of the hourly capacity of a single MSSV at full lift (Input I8). If the steam generator were intact (not ruptured, it would be boiled dry and fully depressurized by either of these valves being stuck open in less than half of the one hour release duration being considered.

However, since the steam generator would be ruptured, continued break flow from the reactor coolant system (RCS) would keep boiling off as long as the RCS remained above 200 °F. From the hot zero power temperature of 547 °F (Input I4), that is a decrease of approximately 350 °F. Using the RCS heat capacity of 1E+6 Btuf F (Input I6), this would require for cool down a heat removal capacity of:

6 a 10 Btu 8 Qcooldown = 350 Fx op =3.5x10 Btu The heat removal capacity required for the decay heat generated for the one hour time period after the reactor trip was calculated by determining the heat generated during each of the tabulated time periods listed for the decay heat rate after shutdown. (Input I 10) This calculation was performed by multiplying the average heat rate over the time interval by the time increment. The average heat rate is the arithmetic mean of the heat rates at the beginning and the end of the interval. Table I lists the results of these calculations. The total heat removal capacity required for the decay heat generated for the one hour time after reactor trip is 1.3E+08 Btu.

Page 10 of 23

Calculation 2013-0018 Revision 3 Table I Integrated Decay Heat Calculation for One Hour Post Reactor Trip Time Average Time increment Q'decay Q'decay Q'decay (sec) (sec) (Btu/hr) (Btu/lu*) (Btu) 1 1 4.2E+08 4.2E+08 1.2E+05 1.5 0.5 4.0E+08 4.1E+08 5.7E+04 2 0.5 3.9E+08 4.0E+08 5.5E+04 4 2 3.7E+08 3.8E+08 2.1E+05 6 2 3.5E+08 3.6E+08 2.0E+05 8 2 3.2E+08 3.4E+08 1.9E+05 10 2 3.1E+08 3.2E+08 1.8E+05 15 5 2.9E+08 3.0E+08 4.2E+05 20 5 2.8E+08 2.9E+08 4.0E+05 40 20 2.4E+08 2.6E+08 1.4E+06 60 20 2.3E+08 2.4E+08 1.3E+06 80 20 2.1E+08 2.2E+08 1.2E+06 100 20 2.0E+08 2.1E+08 1.1E+06 150 50 1.9E+08 2.0E+08 2.7E+06 200 50 1.8E+08 1.9E+08 2.6E+06 400 200 1.5E+08 1.7E+08 9.2E+06 600 200 1.4E+08 1.5E+08 8.1E+06 800 200 1.3E+08 1.4E+08 7.5E+06 1000 200 1.2E+08 1.3E+08 6.9E+06 1500 500 1.1E+08 1.2E+08 1.6E+07 2000 500 l.OE+08 1.1E+08 1.5E+07 3600 1600 8.6E+07 9.3E+07 4.1E+07 Total 1.3E+08 Summing the heat removal requirements for a cool down to 200 °F and the integrated decay heat over one hour gives 4.8E+08 Btu.

If one MSSV were to lift continuously at its full rated flow (Input 18), removing only the latent heat of vaporization over this period (average of 812 Btu/Ibm, Input I 11 ), the total heat removal would be:

QMssv = 812 Btu/lbm x 817,000 lbm/lu* x 1 hr = 6.6E+08 Btu This is greater than the sum of the heat removal required to cool the entire RCS to below 200 °F and to remove all of the decay heat generated for the one hour time period post reactor trip. Therefore, assuming Page 11 of 23

Calculation 2013-0018 Revision 3 additional MSSVs are stuck open is meaningless. There is insufficient energy in the system to continue boiling the RCS at anything close to the rated capacity of a single MSSV, and assuming a single MSSV stuck open is adequate to bound any credible release rate through the main steam safety valve.

If one ADV were to lift continuously at its full rated flow (Input 17), removing only the latent heat of vaporization over this period (average of 812 Btu/Ibm, Input Ill), the total heat removal would be:

QADV = 812 Btu/Ibm X 333,200 lbm/hr X 1 hr = 2.7E+08 Btu This is less than the sum of the heat removal required to cool the entire RCS to below 200 °F and to remove all of the decay heat generated for the one hour time period post reactor trip. The use of an MSSV is needed to provide the additional heat removal capacity that is needed. The previous paragraph showed that the use of one MSSV is sufficient for the entire heat removal requirement for the one hour time period post reactor trip.

Calculate Volumetric Release Rates (Inputs 17, 18 and 112)

The volumetric flow rate for the ADV is:

V'= 333,200lbm x( lht )x0.4006 ft3 =2,225CFM hr 60min lbm Rounding this to two significant figures for convenience and consistency of use in the site's emergency dose assessment methodology gives 2,200 CFM.

The volumetric flow rate for one MSSV is:

V'= 817,000lbm x( lhr )x0.4006 ft3 =5,455CFM hr 60 min Ibm Rounding this to two significant figures for convenience and consistency of use in the site's emergency dose assessment methodology gives 5,500 CFM.

Page 12 of 23

Calculation 2013-0018 Revision 3 Part 2: Initiating Condition Value for General Emergency and Site Area Emergency The monitor readings that correspond to the AG 1 initiating condition for the ABV, DA V, containment purge exhaust, the ADV, and the MSSV is calculated by performing the following:

I. Multiplying each nuclide ' s activity ft*action (Validated Assumption AI , Input II) by its corresponding DCFw8 (Input Il4) and then summing over all nuclides;

2. Multiplying each nuclide's activity fraction (Validated Assumption AI, Input II) by its corresponding DCFrHY (Input I15) and then summing over all nuclides;

3. Multiplying each ofthe sums determined in steps 1 and 2 by the release path flow rate (Input I3 and Part 1 of this calculation), the atmospheric dispersion factor (Input 18, Validated Assumption A3), and the exposure time (Validated Assumption A2 and Part 1 of this calculation);

4. Dividing the I rem TEDE AG 1 IC and the 5 rem Thyroid CDE AG 1 IC (Input Il7) by the applicable factors determined in step 3;

5. Selecting the minimum value of the two factors determined in step 4.

The equation to determine the 1 rem TEDE AG 1 initiating condition is shown below.

. rea M omtor

d"mg ('"'

r<~I"/ cc ) = -.--- -- - - lrem F x L:(JxDCFw8 )x X x T x 0.000472 Q

Where P fe

= release path flow rate, /min f =activity fraction, dimensionless DCFws =dose conversion factor for the combined exposure pathways, rem-cc/flCi-hr x/Q = annual average atmospheric dispersion factor, s/m3 T = exposure time, hr 0.000472 = conversion factor, min-m3/s-ft3 Similarly, the equation to determine the 5 rem Thyroid CDE AG 1 initiating condition is shown below.

. rea d"mg ( r<~I M omtor ,,.,.1cc) = -.------------ 5 rem -- -

Fx L (Jx DCFTHY )x X x Tx 0.000472 Q

Where P = release path flow rate, ft3/min f = activity fraction, dimensionless DCFrHY =dose conversion factor for the dose equivalent to the thyroid, rem-cc/f.LCi-hr x/Q = annual average atmospheric dispersion factor, s/m3 Page 13 of 23

Calculation 2013-0018 Revision 3 T = exposure time, hr 0.000472 =conversion factor, min-m 3/s-ft3 The same equations are used to calculate the monitor readings for the AS I IC initiating condition by replacing the values used in the equations with 0.1 rem and 0.5 rem (Input !16), respectively.

Table 2 shows the results of the calculations performed for the factor 2,(! X DCFw 8 ) for each release pathway.

Table 2 2,(! X DCFw 8 ) Factor Calculation Nuclide DCFwB ABV/DA Cols. Cont Purge Cols. ADV/MSS Cols.

V Activity AxB Activity AxD V Activity AxF Fraction Fraction Fraction A B c D E F G Ar-41 8.0E+02 - - 1.3E-01 1.1E+02 -

Kr-83m l.SE-02 8.7E-03 1.3E-04 2.9E-04 4.3E-06 6.1E-03 8.9E-05 Kr-85m 9.3E+01 3.6E-02 3.4E+OO 2.2E-03 2.0E-01 2.5E-02 2.3E+OO Kr-85 1.3E+OO 3.1E-02 4.1E-02 9.1E-02 1.2E-01 2.1E-02 2.7E-02 Kr-87 5.1E+02 2.7E-02 1.4E+01 6.9E-04 3.5E-01 1.7E-02 8.5E+OO Kr-88 1.3E+03 7.7E-02 l.OE+02 3.4E-03 4.4E+OO 4.9E-02 6.4E+01 Kr-89 1.2E+03 2.4E-03 2.9E+OO 5.2E-06 6.3E-03 1.7E-03 2.0E+OO Xe-131m 4.9E+OO 2.2E-03 l.lE-02 3.9E-03 1.9E-02 1.5E-03 7.2E-03 Xe-133m 1.7E+01 1.7E-02 2.9E-01 8.5E-03 l.SE-01 l.IE-02 1.8E-O 1 Xe-133 2.0E+01 6.8E-01 1.4E+01 7.5E-01 1.5E+01 4.3E-OI 8.7E+OO Xe-135m 2.5E+02 6.8E-03 1.7E+OO 5.3E-05 1.3E-02 4 .2E-03 l.IE+OO Xe-135 1.4E+02 8.7E-02 1.2E+01 9.1E-03 1.3E+OO 5.9E-02 8.3E+OO Xe-137 1.1E+02 4.6E-03 S.OE-01 l.IE-05 1.2E-03 3.0E-03 3.3E-01 Xe-138 7.2E+02 2.2E-02 1.6E+ 01 1.8E-04 1.3E-01 1.4E-02 9.9E+OO 1-131 5.3E+04 8.9E-04 4.7E+01 1.7E-05 8.8E-01 1.5E-O 1 7.8E+03 1-133 1.5E+04 1.2E-03 1.8E+O 1 2.6E-06 3.9E-02 2.1E-01 3.1E+03 Co-58 1.7E+04 1.4E-05 2.5E-01 4.0E-06 6.8E-02 - -

Co-60 2.7E+05 6.5E-06 1.8E+OO l.SE-06 4.9E-OI - -

Mn-54 1.2E+04 4.3E-06 5.2E-02 1.2E-06 1.4E-02 - -

Fe-59 2.3E+04 1.4E-06 3.3E-02 4.0E-07 9.2E-03 - -

Cs-134 6.3E+04 4.3E-06 2.7E-01 1.2E-06 7.4E-02 - -

Cs-137 4.1E+04 7.2E-06 3.0E-01 2.0E-06 8.3E-02 - -

Totals 2.3E+02 1.3E+02 1.1E+04 Page 14 of 23

Calculation 2013-0018 Revision 3 Table 3 shows the results of the calculations performed for the factor 'i,(f X DCFTHY) for each release pathway.

Table 3

'i,(f X DCFTHY) Factor Calculation Nuclide DCFmv ABV/DA Cols. Cont Purge Cols. ADV/MSS Cols.

V Activity AxB Activity AxD V Activity AxF Fraction Fraction Fraction A B c D E F G 1-131 1.3E+06 8.9E-04 1.2E+03 1.7E-05 2.2E+Ol l.SE-0 1 1.9E+05 l-133 2.2E+05 1.2E-03 2.7E+02 2.6E-06 5.8E-Ol 2.1E-Ol 4.6E+04 Totals 1.4E+03 2.2E+Ol 2.4E+05 Page 15 of 23

Calculation 20 13-00 18 Revision 3 Table 4 shows the results of the calculations performed for the RS 1 and RG 1 IC's when a one hour release duration is assumed. All values have been rounded to 2 significant figures.

Table 4 AS I and AG 1 IC Calculations for One Hour Release Duration 5 rem 0.5 rem I rem Thyroid 0.1 rem Thyroid TEDEIC CDEIC TEDEIC CDEIC Vent (f!Ci/cc) (f!Ci/cc) ().!Ci/cc) ().!Ci/cc)

ABV F = 66,400 CFM (Input 113)

I(fxDCE:vs )=2.3E+02 rem-cc

. (Table 2) j£1-hr 130 100 13 10 I(jxDCf;.HY)= 1.4E+03 rem. -cc (Table 3) j£1- hr xJQ = 1.09E-06 s/m3 (Input I18)

T = 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Val. Assump. 2)

DAY F = 43,100 CFM (Input Il3)

I(fxDCFws )= 2.3E+02 rem. -cc (Table 2) jiCI- hr 190 160 19 16 I(fx DCF;.HY)= 1.4E+03 rem. - cc (Table 3) jiCI-hr x/Q = 1.09E-06 s/m3 (Input 118)

T = 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Val. Assump. 2)

Containment Purge Vent - Purge Only F = 25,000 CFM (Input 113)

I(fxDCE:vs )= 1.3E+02 rem. -cc (Table 2) jiCI- hr 600 18,000 60 1,800 I(jxDCf;.HY )= 2.2E+01 rem-cc . (Table 3) jiCI- hr x/Q = 1.09E-06 s/m3 (Input I18)

T = 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Val. Assump. 2)

Containment Purge Vent- Purge+ GS Building F = 38,000 CFM (Input 113)

I(fxDCE:vs )= 1.3E+02 rem. -cc (Table 2) jiCI- hr 400 12,000 40 1,200 I(f DCF;.HY)= 2.2E+0 1 rem-cc X . (Table 3) jiCI- hr xJQ = 1.09E-06 s/m3 (Input I18)

T = 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Val. Assump. 2)

Page 16 of 23

Calculation 2013-0018 Revision 3 5 rem 0.5 rem 1 rem Thyroid 0.1 rem Thyroid TEDEIC CDEIC TEDEIC CDEIC Vent ( uCi/cc) ( uCi/cc) ( uCi/cc) (J-LCi/cc)

U2 Containment Purge Vent- Gas Stripper Bldg only F = 13,000 CFM (Input 113)

I( fx DCFw8 ) = 1.3E+02 rem. -cc (Table 2)

£1-hr 1200 34,000 120 3,400 I(jxDCf;,HY ) = 2.2E+01 rem. -cc (Table 3)

£1- hr xJQ = 1.09E-06 s/m3 (Input 118)

T = 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Val. Assump. 2)

ADV F =2,200 CFM (Calc. Part 1)

I(fxDCFwa )= l.IE+04 rem-cc. (Table 2) jiCI - hr 80 19 8 1.9 I(jxDCf;,HY ) =2.4E+05 rem. -cc (Table 3)

J-LCI- hr xJQ = 1.09E-06 s/m3 (Input 118)

T = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Calc. Part 1)

MSSV F = 5,500 CFM (Calc. Pmt 1)

I(JxDCFwa) = 1.1E+04 re~-cc (Table 2)

£1- hr 32 7.4 3.2 0.74 I(fx DCF;,HY) = 2.4E+05 rem. -cc (Table 3) jLCI- hr xJQ = 1.09E-06 s/m3 (118)

T = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (Calc. Part 1)

These results will not be affected by a transition from NEI 99-01 Revision 4 to NEI 99-01 Revision 6.

However, Revision 6 directs that the Alett EAL IC (AA1) be based on the GE EAL IC, and 2 orders of magnitude below it (i.e. Ill 00 1h of the GE setpoint). The results/conclusions sections will reflect this with separate tables for Revision 4 and Revision 6 guidance.

Page 17 of 23

Calculation 2013-0018 Revision 3 Part 3: Initiating Condition Value for Unusual Event and Ale1t Revision 4 of Reference Rl4 directs that the UE and Alert EAL ICs should be set at 2x and 200x the ODCM calculated nominal setpoints respectively. However, several of the release point monitor detector setpoints calculated in the Point Beach ODCM have a factor of 4 reduction (i.e. 25% of the nominal setpoint) to account for the possibility of a simultaneous release from multiple paths.

As discussed in Appendix A.3.1 of Reference Rl4 Revision 4, such additional conservatisms are not desirable, though they may be retained if appropriate to ensure proper coordination between the UE, Ale1t, SAE, and GE EAL ICs.

Therefore the 0.25 multiplier that was applied to the PAB Vent and Drumming Area Vent stack will be corrected before determining the appropriate EAL ICs:

Pathway Monitor Almm Setpoint EAL res

(!lCi/cc) (!lCi/cc)

ODCM Conected UE (AUI) Ale1t(AAI)

Auxiliary Building RE-214 & SPINO 23 6.75E-04 2.7E-03 5.4E-03 5.4E-Ol Vent Containment PurgeNent Unit I* IRE-212 & SPINO 21 7.17E-03 N/A 1.4E-02 1.4E+OO Unit 2** 2RE-212 & SPINO 22 4.72E-03 9.4E-03 9.4E-01 Unit 1(2)*** 1(2) RE-212 5.12E+OO l.OE+Ol l.OE+03 Gas Stripper RE-224 3.45E-03 1.4E-02 2.8E-02 2.8E+OO Building Drumming Area RE-221 & SPINO 24 1.04E-03 4.2E-03 8.4E-03 8.4E-01 Vent

Containment Purge alone

- Containment Purge together with GS building ventilation

- - Force Vent in operation without containment purge or GS building When transitioning to NEI 99-01 Revision 6 guidance, the UE setpoints will remain unchanged.

However, the Alert setpoints shown above will be supplanted by values that are 1% of the GE setpoints determined in the previous pmt of this calculation.

Page 18 of 23

Calculation 20 I 3-0018 Revision 3 Results and

Conclusions:

The results were calculated using a source term that is consistent with the ODCM and the guidance of NEI 99-01. The results are applicable for a reactor thermal power operating level of 1800 MWth and are applicable generally to other thermal power operating levels due to the use of activity fractional data and not absolute activity values.

The monitor readings that correspond to the AG1 initiating condition value of 1 rem TEDE or 5 rem Thyroid CDE are listed below.

Monitor Reading 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with only 600 JlCi/cc I containment purge in operation (25,000 cfm) 1(2)RE-309, Containment Purge Exhaust High Range Gas with only 600 J1Ci/cc containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid Range Gas with both 400 1-LCi/cc I purge and GS building ventilation in operation (38,000 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with both 400 JlCi/cc purge and GS Building in operation (38,000 cfm)

I 2RE-309, Containment Purge Exhaust High Range Gas with only 1200 JlCi/cc*

GS Building in operation (13,000 cfm)

RE-317, Auxiliary Building Exhaust Mid-Range Gas 100 JlCi/cc RE-319, Auxiliary Building Exhaust High Range Gas 100 J1Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas Off scale high **

I 1(2) RE-231, Steam Line 1A(2A); 1(2)RE-232, Steam Line 1B(2B)

Atmospheric Dump Valve (ADV) release 19 J1Ci/cc Main Steam Safety Valve (MSSV) release 7.4 J1Ci/cc

1200 JlCi/cc is off scale high for RE-307. Therefore, no value is given for this same oper ating configuration for that instrument.

- The upper range limit on R-327 is 100 JlCi/cc, so this instrument would be over-ranged by the nominal EAL IC of 160 JlCi/cc. However, this is close enough to the upper range of the instrument that it is judged appropriate to, in the absence of real -time dose projection capability and no other indications, to make a classification declaration based on a valid off-scale high reading of this instrument.

Page 19 of 23

Calculation 2013-0018 Revision 3 The monitor readings that correspond to the ASl initiating condition value of 100 mrem TEDE or 5 rem Thyroid CDE are listed below for a one hour release duration.

Monitor Reading 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with only 60 !-tCi/cc containment purge in operation (25,000 cfm) 1(2)RE-309 Containment Purge Exhaust High-Range Gas, with only 60 !-tCi/cc containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid Range Gas with both 40 11Ci/cc purge and GS building ventilation in operation (38,000 cfm) .

2RE-309, Containment Purge Exhaust High Range Gas with both 40 f.lCi/cc purge and GS building ventilation in operation (38,000 cfm) .

2RE-309, Conta inment Purge Exhaust High Range Gas with only 120 11Cilcc GS building ventilation in operation (13,000 cfm).

RE-317, Auxiliary Building Exhaust Mid-Range Gas 10 ~J,Ci /cc RE-319, Auxiliary Building Exhaust High Range Gas 10 ~J,Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas 16 ~J,Ci /cc I 1(2) RE-231 , Steam Line 1A(2A); 1(2)RE-232, Steam Line 1B(2B)

Atmospheric Dump Valve (ADV) release 1.9 uCilcc Main Steam Safety Valve (MSSV) release 0.74 uCi/cc Page 20 of 23

Calculation 2013-00 18 Revision 3 The monitor readings that correspond to the AA1 initiating condition (1/100 of AG1} under NEI99-01 Revision 6 are listed below. These will require submittal and approval of a License Amendment Request to implement.

Monitor Reading 1(2}RE-307 Containment Purge Exhaust Mid-Range Gas, with only 6.0 !lCi/cc containment purge in operation (25,000 cfm) 1(2}RE-309 Containment Purge Exhaust High-Range Gas, with only 6.0 !lCi/cc containment purge in operation {25,000 cfm}

2RE-307, Containment Purge Exhaust Mid Range Gas with both 4.0 !lCi/cc purge and GS building ventilation in operation (38,000 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with both 4.0 J.!Ci/cc purge and GS building ventilation in operation {38,000 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with only 12 J.!Ci/cc GS building ventilation in operation (13,000 cfm).

RE-317, Auxilimy Building Exhaust Mid-Range Gas 1.0 llCi/cc RE-319, Auxiliary Building Exhaust High Range Gas 1.0 llCi/ cc RE-327, Drumming Area Exhaust Mid-Range Gas I .6 j..LCi/cc 1(2) RE-231, Steam Line 1A(2A); 1(2)RE-232, Steam Line I 1B(2B)

Atmospheric Dump Valve (ADV) release 0.19 llCi/cc Main Steam Safety Valve (MSSV) release 0.074 j..LCi/cc Page 21 of 23

Calculation 20 13-00 I 8 Revision 3 The monitor readings that correspond to the AA1 initiating condition (200X ODCM Alarm Setpoints) under NEI 99-01 Revision 4 are listed below:

Monitor Reading RE-315, Auxiliary Building Exhaust Low Range Gas 5.4E-O 1 J..LCi/cc RE-317, Auxiliary Building Exhaust Low Range Gas 5.4E-O 1 J..LCi/cc 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with 1.4E+OO flCi/cc only containment purge in operation (25,000 cfm) 1(2)RE-309 Containment Purge Exhaust High-Range Gas, with 1.4E+OO flCi/cc onl)' containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid-Range Gas with both 9.4E-01 f.!Ci/cc purge and GS building ventilation in operation (38,000 cfm).

2RE-307, Containment Purge Exhaust Mid-Range Gas with only 2.8E+OO flCi/cc GS building ventilation in operation (13,000 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with only 2.8E+OO flCi/cc GS building ventilation in operation (13,000 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with only 1.0E+03 flCi/cc forced vent of containment (35 cfm).

RE-327, Drumming Area Exhaust Mid-Range Gas 8.4E-O 1 flCi /cc Page 22 of 23

Calculation 20 13-0018 Revision 3 The monitor readings that correspond to the AUl initiating condition {2X ODCM Alarm Setpoints) under NEI99-01 Revision 4 are listed below:

Monitor Reading RE-315, Auxiliary Building Exhaust Low Range Gas 5.4E-03 ~Ci/cc RE-317, Auxiliary Building Exhaust Low Range Gas 5.4E-03 ~Ci/cc 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with 1.4E-02 ~Ci/cc only containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid-Range Gas with both 9.4E-03 1J.Ci/cc purge and GS building ventilation in operation (38,000 cfm).

2RE-305, Containment Purge Exhaust Low-Range Gas with both 9.4E-03 ~Ci/cc purge and GS building ventilation in operation (38,000 cfm).

2RE-307, Containment Purge Exhaust Mid-Range Gas with only 2.8E-02 ~Ci/cc GS building ventilation in operation (13,000 cfm).

2RE-307, Containment Purge Exhaust Mid-Range Gas with only l.OE+01 1J.Ci/cc forced vent of containment (35 cfm).

2RE-309, Containment Purge Exhaust High Range Gas with only l.OE+Ol 1J.Ci/cc forced vent of containment (35 cfm).

RE-325, Drumming Area Exhaust Low-Range Gas 8.4E-03 ~Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas 8.4E-03 ~Ci/cc Page 23 of 23