1CAN122101, Response to Request for Supplemental Information Concerning Licensing Amendment Request to Revise Technical Specification 3.4.12 and 3.4.13

From kanterella
(Redirected from ML21337A245)
Jump to navigation Jump to search

Response to Request for Supplemental Information Concerning Licensing Amendment Request to Revise Technical Specification 3.4.12 and 3.4.13
ML21337A245
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 12/02/2021
From: Gaston R
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN122101
Download: ML21337A245 (65)
v • d • e


Text

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 1CAN122101 December 2, 2021 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Response to Request for Supplemental Information Concerning Licensing Amendment Request to Revise Technical Specification 3.4.12 and 3.4.13 Arkansas Nuclear One, Unit 1 NRC Docket No. 50-313 Renewed Facility Operating License No. DPR-51 Entergy Operations, Inc. (Entergy) submitted a license amendment request (LAR) for Arkansas Nuclear One, Unit 1 (ANO-1) in Reference 1. The proposed amendment would revise the Dose Equivalent I-131 (DEI) and the Reactor Coolant System (RCS) primary activity limits required by Technical Specification (TS) 3.4.12, RCS Specific Activity. In addition, the primary-to-secondary leak rate limit provided in TS 3.4.13, RCS Leakage, would be revised. These proposed changes are due to non-conservative inputs used in the Steam Generator Tube Rupture (SGTR) accident, the Main Steam Line Break (MSLB) accident, and the Control Rod Ejection Accident (CREA) dose calculations.

The U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed to complete its acceptance review (Reference 2). The requested supplemental information is provided in response to this request in the Enclosure. The supplemental information does not affect the no significant hazards consideration provided in Reference 1.

There are no new regulatory commitments contained in this submittal.

1CAN122101 Page 2 of 3 If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.

I declare under the penalty of perjury that the foregoing is true and correct. Executed on December 2, 2021.

Respectfully, Ronald W. Digitally signed by Ronald W. Gaston Gaston 17:01:01 -06'00' Date: 2021.12.02 Ron Gaston RWG/rwc References

1. Entergy Operations, Inc. (Entergy) letter to the U. S. Nuclear Regulatory Commission (NRC), "License Amendment Request Proposed Technical Specifications 3.4.12 and 3.4.13 revised Dose Calculations," (1CAN092101),

ML21274A874, dated September 30, 2021.

2. NRC letter to Entergy, "Arkansas Nuclear One, Unit 1 - Supplemental Information Needed for Acceptance of Requested Licensing Action RE: Licensing Amendment Request Concerning Revised Dose Calculations (EPID L-2021-LLA-0181) "

(1CNA112101), ML21320A212, dated November 17, 2021.

Enclosure:

Response to Request for Supplemental Information Related to LAR to Revise Technical Specification 3.4.12 and 3.4.13 Attachments:

1. Basic Parameters Used in the Dose Consequence Analyses
2. Steam Generator Tube Rupture (SGTR) Model Information
3. Main Steam Line Break (MSLB) Model Information
4. Control Rod Ejection Accident (CREA) Model Information
5. SGTR RADTRAD Input Files
6. MSLB RADTRAD Input Files
7. CREA RADTRAD Input files cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One

1CAN122101 Page 3 of 3 NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official

Enclosure 1CAN122101 Response to Request for Supplemental Information Related to LAR to Revise Technical Specification 3.4.12 and 3.4.13

1CAN122101 Enclosure Page 1 of 5 RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION RELATED TO LAR TO REVISE TECHNICAL SPECIFICATION 3.4.12 AND 3.4.13 By letter dated September 30, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21274A874), Entergy Operations, Inc. (Entergy) submitted a license amendment request (LAR) for Arkansas Nuclear One, Unit 1 (ANO-1). The proposed amendment would revise the Dose Equivalent I-131 (DEI) and the Reactor Coolant System (RCS) primary activity limits required by Technical Specification (TS) 3.4.12, RCS Specific Activity. In addition, the primary-to-secondary leak rate limit provided in TS 3.4.13, RCS Leakage, would be revised. These proposed changes are due to non-conservative inputs used in the Steam Generator Tube Rupture (SGTR) accident, the Main Steam Line Break (MSLB) accident, and the Control Rod Ejection Accident (CREA) dose calculations.

The U.S. Nuclear Regulatory Commission (NRC) staff performed an acceptance review of the LAR in accordance with Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-109, Revision 3, and Acceptance Review Procedures, dated July 20, 2020 (ADAMS Accession No. ML20036C829), and determined that the application is unacceptable for review, with opportunity to supplement because it is missing sufficient information for the NRC staff to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements for the protection of public health and safety and the environment.

To make the application complete, the NRC staff requests that the licensee supplement the application to address the information requested, as described below.

REGULATORY BASIS FOR REQUEST The regulation in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.67, Accident source term, requires that:

1) A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under § 50.90.

The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

2) The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not

1CAN122101 Enclosure Page 2 of 5 receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

SUPPLEMENTAL INFORMATION NEEDED FOR THE AMENDMENT REQUEST A preliminary review of the LAR has determined that additional information is needed for the NRC staff to begin performing a meaningful review. Although detailed information appears to be provided for other non-radiological analyses, the critical inputs to the dose analyses are not readily displayed in the amendment request. Therefore, the following information is needed:

1. Please provide additional information describing, for each design basis accident affected by the proposed changes, all the basic parameters used in the dose consequence analyses. This information should include the current licensing basis (CLB) values, the revised values, where applicable, as well as the basis for any changes to the CLB values. For clarity, please provide the requested information in separate tables for each affected design basis accident.

Entergy's Response Attachment 1 provides additional information describing, for each design basis accident affected by the proposed changes, all the basic parameters used in the dose consequence analyses.

2. In addition, please provide additional information describing the models and assumptions used in the dose consequence analyses affected by the proposed changes. Alternatively, to support a timely NRC staff review, the calculation packages that typically contain this information may be provided for the affected design basis accidents.

Entergy's Response Source Term The reactor coolant source terms in the CLB analyses are based on coolant measurements from 2007. In an effort to standardize the reactor coolant source term methodologies among the Entergy units, these source terms are being updated to ANSI/ANS-18.1-2020 as they are revised. This standard identifies (i) the radiologically-applicable isotopes and (ii) the relative concentrations among the isotopes to ensure a consistent fleet approach to performing these radiological calculations. In order to minimize periodic updates, this methodology eliminates or minimizes the need for measured plant data that may change over time and is as independent as possible from fuel parameters such as fuel mechanical

1CAN122101 Enclosure Page 3 of 5 design, burnup, and enrichment that can change each cycle. The ANO-1 reactor coolant sources were revised to this standard as part of this application.

Once the normal reactor coolant sources are developed based on this standard, the reactor coolant sources are increased to meet the applicable ANO-1 Technical Specifications (TSs).

Current TS Surveillance Requirement (SR) 3.4.12.2 limits the Dose Equivalent I-131 (DEI) to less than 1.0 microcurie per gram (µCi/g) while TS SR 3.4.12.1 limits the Dose Equivalent Xe-133 (DEX) to less than 2200 µCi/g. The DEI is calculated with the Committed Effective Dose Equivalent (CEDE) dose conversion factors using Table 2.1 of Environmental Protection Agency (EPA) Federal Guidance Report No. 11, while the DEX applies the effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12. Consistent with Regulatory Issue Summary (RIS) 2006-04, the alkali metals are also included in the reactor coolant source terms.

The CREA applies the core source term inventory since this accident involves a gap release.

The coolant and core activities applied in these analyses are listed below.

ANO-1 Reactor Coolant Activities (µCi/g)

DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary Kr-85m 3.51E+00 Kr-85 1.45E+02 Kr-87 5.98E+00 Kr-88 7.01E+00 Xe-131m 1.61E+03 Xe-133m 3.14E+00 Xe-133 7.90E+01 Xe-135m 1.50E+01 Xe-135 2.69E+01 Xe-137 7.00E+01 Xe-138 1.81E+01 Br-84 3.85E+00 1.23E+00 I-131 5.65E-01 5.61E-02 I-132 2.00E+00 1.99E-01 I-133 1.74E+00 1.75E-01

1CAN122101 Enclosure Page 4 of 5 DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary I-134 2.93E+00 3.06E-01 I-135 2.45E+00 2.51E-01 Rb-88 1.49E-01 2.63E-02 Cs-134 2.95E-01 5.33E-02 Cs-135 4.52E-02 8.23E-03 Cs-137 1.88E-01 2.29E-02 ANO-1 Core Activity of Gap Isotopes (Ci)

Core Core Activity Isotope Activity (Ci) Isotope (Ci)

Kr-83m 8.77E+06 I-130 1.36E+06 Kr-85 9.61E+05 I-131 7.22E+07 Kr-85m 1.90E+07 I-132 1.05E+08 Kr-87 3.73E+07 I-133 1.48E+08 Kr-88 5.01E+07 I-134 1.67E+08 Xe-131m 7.55E+05 I-135 1.41E+08 Xe-133 1.48E+08 Cs-134 1.46E+07 Xe-133m 4.60E+06 Cs-136 2.98E+06 Xe-135 3.51E+07 Cs-137 9.88E+06 Xe-135m 3.09E+07 Cs-138 1.38E+08 Xe-138 1.27E+08 Rb-86 1.29E+05 Attachments 2, 3, and 4 address the SGTR, MSLB, and CREA analyses, respectively. These attachments provide information to understand the RADTRAD input files provided in response to request 3, below.

1CAN122101 Enclosure Page 5 of 5

3. Due to the small margins to the acceptance criteria for several assessments, the NRC staff plans to perform confirmatory analyses. To increase the efficiency of the staff review, please provide the RADTRAD 3.03 input and output files, if available.

Entergy's Response Attachments 5, 6, and 7 provide a listing of the RADTRAD input files for the SGTR, MSLB, and CREA events, respectively. In discussions with the NRC Staff, it was determined that the output files were not necessary, Electronic copies of the files were provided separately.

1CAN122101 Attachment 1 Basic Parameters Used in the Dose Consequence Analyses

1CAN122101 Page 1 of 8 CONTROL ROOM PARAMETERS Parameter CLB Value Current Value Reason for Change Volume 4.00E+05 ft3 4.00E+05 ft3 No change Normal Air Intake Flow 35,200 cubic feet 35,200 cfm No change per minute (cfm)

Recirculation Flow 1667 cfm 1667 cfm No change Recirculation Filtration 95% Elemental 95% Elemental No change 95% Organic 95% Organic 99% Aerosol 99% Aerosol Filtered Intake Flow 333 cfm 333 cfm No change Intake Filtration 95% Elemental 99% Elemental The two ANO Control 95% Organic 99% Organic Room Emergency Ventilation System 99% Aerosol 99% Aerosol (CREVS) trains are different. One train consists of a single 4-inch charcoal bed with a high-efficiency particulate absorbing (HEPA). The other train consists of two 2-inch charcoal beds each with a HEPA filter in series.

Unfiltered In-leakage SGTR: 85 cfm 82 cfm Consistent (post-isolation) MSLB: 85 cfm in-leakage for all calculations CREA: 82 cfm Isolation Time (based 10 seconds 10 seconds No change on high radiation in intake)

Time of High Radiation SGTR: 11 minutes SGTR: 9.6 minutes Consistent with Detected in Control MSLB: 0 seconds MSLB: 0 seconds scram timing Room Intake CREA: 0 seconds CREA: 0 seconds Occupancy Factor No change 0 - 1 day 1.0 1.0 1 - 4 days 0.6 0.6 4 - 30 days 0.4 0.4 Breathing Rate 3.5E-04 m3/s 3.5E-04 m3/s No change 0 - 30 days

1CAN122101 Page 2 of 8 DISPERSION COEFFICIENTS (seconds/m3)

Current Parameter CLB Value Reason for Change Value Offsite Exclusion Area Boundary No change 0-2 hours 6.8E-04 6.8E-04 Low Population Zone 0-8 hours 1.1E-04 1.1E-04 8-24 hours 1.1E-05 1.1E-05 No change 1-4 days 4.0E-06 4.0E-06 4-30 days 1.3E-06 1.3E-06 Control Room Main Steam Safety Valve 0-2 hours 1.90E-02 1.90E-02 2-8 hours 1.23E-02 1.23E-02 No change 8-24 hours 5.83E-03 5.83E-03 1-4 days 3.80E-03 3.80E-03 4-30 days 3.10E-03 3.10E-03 Atmospheric Dump Valve 0-2 hours 4.10E-03 4.10E-03 2-8 hours 2.59E-03 2.59E-03 No change 8-24 hours 1.12E-03 1.12E-03 1-4 days 8.32E-04 8.32E-04 4-30 days 5.91E-04 5.91E-04 Main Steam Pipe No change 0-2 hours 3.15E-03 3.15E-03

1CAN122101 Page 3 of 8 STEAM GENERATOR TUBE RUPTURE PARAMETERS Parameter CLB Value Current Value Reason for Change Pre-Accident Iodine 60 µCi/g 6 µCi/g Proposed License Spike DEI Amendment Equilibrium Iodine 1.0 µCi/g 0.25 µCi/g Proposed License Inventory Amendment Secondary System DEI 0.1 µCi/g 0.1 µCi/g No change Primary-to-Secondary 150 gpd/SG 150 gpd/SG No change (PS) Leakage Rate Fraction of PS leakage 15% 100% Original value is not That Is Vaporized supported Fraction of Ruptured 100% pre- 100% pre- Original value is not Flow That Is Vaporized scram scram supported, new values based 15% post- 100% for first on Computational Fluid scram 4.4 min post- Dynamics (CFD) analysis scram 40% thereafter Time of Scram 11 min 9.6 min Based on updated Thermal-Hydraulics (T/H) analyses Noble Gas DEX 2200 µCi/g 2200 µCi/g No change RCS Mass 5.14E+05 lbs 5.37E+05 lbs Updated calculation Secondary System 3.76E+04 6.00E+04 Maximum mass is applied to Mass lbs/SG lbs/SG maximize initial source term inventory Percentage of Failed 0% 0% No change Fuel Rods Density used for 62.4 lb/ft3 62.4 lb/ft3 No change Leakage Volume-to-Mass Conversion Single Active Failure Failure of ADV Failure of No change block valve to Atmospheric open Dump Valve (ADV) block valve to open Response Time to 30 min 30 min No change Open Block Valve

1CAN122101 Page 4 of 8 Parameter CLB Value Current Value Reason for Change Initial Ruptured Tube 34.56 lb/s 36.83 lb/s Updated T/H analyses Flow Rate (varies with time)

Isolation Time of 34 min 70 min Based on updated T/H Ruptured Steam analyses and conservative Generator (SG) operator response times Isolation Time of Intact 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> No change SG Accident Induced 335 335 No change Spike Iodine Appearance Multiplier Duration of Iodine 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> No change Spike SG Iodine Partition 100 100 No change Coefficient SG Moisture Carryover 0.1% 0.1% No change Fraction Condenser Partition 10,000 10,000 No change Coefficient Iodine Chemical 97% elemental 97% elemental No change Species 3% organic 3% organic

1CAN122101 Page 5 of 8 MAIN STEAM LINE BREAK PARAMETERS Parameter CLB Value Current Value Reason for Change Pre-Accident Iodine 60 µCi/g 60 µCi/g No change Spike DEI Equilibrium Iodine 1.0 µCi/g 1.0 µCi/g No change Inventory Secondary System DEI 0.1 µCi/g 0.1 µCi/g No change Primary-to-Secondary 0.5 gpm/SG 0.5 gpm/SG No change (PS) Leakage Rate Fraction of PS leakage Affected SG Both SGs Original value is not that is vaporized 100% 100% supported Intact SG 20%

Time of Scram 0 sec 0 sec No change Noble Gas DEX 2200 µCi/g 2200 µCi/g No change RCS Mass 5.14E+05 lbs 5.37E+05 lbs Updated calculation Secondary System 3.76E+04 6.00E+04 Maximum mass is applied to Mass lbs/SG lbs/SG maximize initial source term inventory Percentage of Failed 0% 0% No change Fuel Rods Density used for 62.4 lb/ft3 62.4 lb/ft3 No change Leakage Volume-to-Mass Conversion Single Active Failure Failure of ADV Failure of ADV No change block valve to block valve to open open Response Time to 30 min 30 min No change Open Block Valve Isolation Time of 251.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 251.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> No change Affected S/G Accident Induced 500 500 No change Spike Iodine Appearance Multiplier Duration of Iodine 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> No change Spike

1CAN122101 Page 6 of 8 Parameter CLB Value Current Value Reason for Change Isolation Time of Intact 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> No change SG Intact SG Iodine 100 100 No change Partition Coefficient Intact SG Moisture 0.1% 0.1% No change Carryover Fraction Iodine Chemical 97% elemental 97% elemental No change Species 3% organic 3% organic

1CAN122101 Page 7 of 8 CONTROL ROD EJECTION ACCIDENT PARAMETERS Parameter CLB Value Current Value Reason for Change Source Terms Core Core No change Fraction of Damaged 14% 14% No change Fuel Peaking Factor of 1.8 1.8 No change Damaged Rods Failure Mechanism DNB DNB No change Gap Fraction Noble Gases: Noble Gases: No change 10% 10%

Iodine: 10% Iodine: 10%

Alkali Metals: Alkali Metals:

12% 12%

Time of Scram 0 sec 0 sec No change Primary-to-Secondary Leakage Case RCS Mass 5.14E+05 lbs 5.37E+05 lbs Updated calculation Primary-to-Secondary 150 gpd/SG 39 gpd/SG Proposed License (PS) Leakage Rate Amendment Fraction of PS leakage 15% 100% Original value is not that is vaporized supported Density used for 62.4 lb/ft3 62.4 lb/ft3 No change Leakage Volume-to-Mass Conversion Isolation Time of SGs 38.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 38.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> No change (PS Leakage Case)

Single Active Failure Failure of ADV Failure of ADV No change block valve to block valve to open open Response Time to 30 min 30 min No change Open Block Valve Iodine Chemical 97% elemental 97% elemental No change Species 3% organic 3% organic (PS Leakage Case)

1CAN122101 Page 8 of 8 Parameter CLB Value Current Value Reason for Change Containment Leakage Case Containment Volume 1.81E+06 ft3 1.81E+06 ft3 No change Containment Leakage 0-1 day: 0-1 day: No change Rate 0.2%/day 0.2%/day

> 1 day: > 1 day:

0.1 %/day 0.1 %/day Aerosol Deposition 0.1 hr-1 for 0.1 hr-1 for No change Rate 69 hours7.986111e-4 days <br />0.0192 hours <br />1.140873e-4 weeks <br />2.62545e-5 months <br /> 69 hours7.986111e-4 days <br />0.0192 hours <br />1.140873e-4 weeks <br />2.62545e-5 months <br /> (Containment Leakage Case)

Containment Spray No credit No credit No change Iodine Chemical 95% aerosol 95% aerosol No change Species 4.85% 4.85%

(Containment Leakage elemental elemental Case) 0.15% organic 0.15% organic

1CAN122101 Attachment 2 Steam Generator Tube Rupture (SGTR)

Model Information

1CAN122101 Page 1 of 15 STEAM GENERATOR TUBE RUPTURE (SGTR)

MODEL INFORMATION 1.0 EVENT DESCRIPTION While operating at full power, the double-ended rupture of a steam generator (SG) tube occurs with unrestricted discharge from each end. The initial leak rate exceeds the normal makeup (MU) to the reactor coolant system (RCS) and system pressure decreases. No initial operator action is assumed and the primary system pressure decreases until a reactor trip occurs on low reactor coolant pressure. The turbine trips as a result of the reactor trip. Consistent with Section 5.4 of Appendix F to Regulatory Guide (RG) 1.183, a coincident loss of offsite power (LOOP) is assumed coincident with the reactor scram and turbine trip rather than the tube rupture.

Due to the LOOP, the reactor coolant pumps (RCPs) trip, the main feedwater (MFW) pumps trip, and the condenser becomes unavailable. Emergency feedwater (EFW) actuates automatically to raise SG liquid levels. Closure of the turbine stop valves (TSVs) causes the main steam line pressure to increase and open the atmospheric dump valves (ADVs) and main steam safety valves (MSSVs). The fission products escaping from the ADVs and/or MSSVs are released directly to the atmosphere.

The operator performs normal post-trip verifications and then begins a cooldown of the reactor coolant system (RCS) using the ADV(s) in manual mode. After the RCS temperature has decreased to a value that corresponds to a saturation pressure which is below the lowest MSSV setpoint, the affected SG can be isolated. The RCS cooldown continues by steaming the unaffected SG. RCS pressure is maintained near the minimum adequate subcooling margin (SCM) by throttling high pressure safety injection (which started automatically on a low reactor coolant pressure signal or was started manually by the operator) and cycling the electromatic relief valve (ERV).

The SGTR scenario considers a single active failure and a coincident LOOP. The accident chronology applied in this analysis is reported below.

SGTR Chronology Time Event Comments 0 seconds SGTR occurs (sec) 9.6 minutes Reactor trip on low RCS pressure Per Assumption 5.4 of Appendix F with consequential loss of offsite to RG 1.183 power ADV block valve on intact SG fails to open (single active failure), all releases are through the MSSVs.

1CAN122101 Page 2 of 15 Time Event Comments 9.6 minutes + Control Room isolated on high 10 sec radiation in intakes 30 minutes Operators begin plant cooldown Cooldown is started using the intact SG but operators notice ADV on intact SG fails to open, so cooldown is started using ruptured SG. Operator is dispatched to address ADV on intact SG.

60 minutes Failed ADV block valve manually opened.

70 minutes Ruptured SG is isolated. Tube and RCS temperature decreased to a Primary-to-Secondary (PS) leakage value that corresponds to a no longer released to environment. saturation pressure which is below the lowest MSSV setpoint 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Decay Heat Removal (DHR) conditions reached. PS leakage from intact loop terminated 2.0 RADIOLOGICAL CONSEQUENCES The RCS response to a SGTR was evaluated with Framatomes RELAP5/MOD2-B&W thermal-hydraulics (T/H) computer code with ANO-1s current enhanced once-through steam generators (EOTSGs). This T/H analysis yielded the scram timing, isolation time of the ruptured SG, and the time-dependent break flow rates.

The radiological consequences of the SGTR accident were then analyzed with RADTRAD 3.03.

Due to code limitations, the scenario was broken down into six cases, which were evaluated separately and summed to develop total dose. For each case, the impact on control room dose from the plume was also evaluated using modified isotopic-specific dose conversion factors and was shown to be negligible. These cases are listed in the table below.

Description 1 Secondary coolant iodine release from both SGs Secondary coolant alkali metal release from both 2

SGs Primary coolant iodine release due to PS and 3

ruptured tube leakage Primary coolant alkali metal release due to PS and 4

ruptured tube leakage

1CAN122101 Page 3 of 15 Description Primary coolant noble gas release due to PS and 5

ruptured tube leakage Primary coolant iodine release from accident-6 induced spike due to PS and ruptured tube leakage The RADTRAD model developed to evaluate these cases is illustrated below.

Recirc Filtered Intake CR 1 Intake 3 1

Control CR 2 Room Intake Unfiltered Intake Vaporized 6 P/S leakage 6

4 Environment Environment (Condenser/ 5 (Condenser/

MSSV/ADV) MSSV/ADV) 5 Fuel Vaporized 9 8 P/S leakage Vaporized Fuel Leakage Tube For Accident 10 Rupture Steaming Induced Spike Steaming 6

4 2 3 Intact Primary Ruptured S/G System 7 S/G Liquid Liquid Tube Rupture Liquid 2.1 COMPARTMENTS The volumes of the compartments are listed below. Since the RCS inventories are typically reported in terms of mass, the RADTRAD runs applied the masses of these compartments in grams (rather than cubic feet [ft3]) with corresponding flows between the compartments in grams/minute (rather than cubic feet per minute [cfm]).

1CAN122101 Page 4 of 15 Compartment Name Volume Comments 1 Control Room 4.00E+05 ft3 2 RCS 2.44E+08 g Equivalent to 5.37E+05 pounds (lbs) 3 Ruptured SG Liquid 2.72E+07 g Equivalent to 6.00E+04 lbs 4 Intact SG Liquid 2.72E+07 g Large enough to minimize depletion 5 Fuel 1.00E+10 g effects over the 8-hour iodine spike The control room is modeled with a recirculating filter that starts upon isolation at ten seconds after the scram with the following parameters.

Time Flow Filters (Hours)

Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

0.163 1667 cfm 99% 95% 95%

2.2 PATHWAYS Pathway 1: Filtered Control Room (CR) Intake This pathway represents the filtered intake into the CR. This pathway starts ten seconds after the scram and is modeled with the following flows and filters.

Time Flow Filters (Hours)

Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

0.163 333 cfm 99% 99% 99%

1CAN122101 Page 5 of 15 Pathway 2: Un-Filtered CR In-leakage This pathway represents the unfiltered intake into the CR. The ANO-1 design basis in-leakage rate of 82 cfm was applied in this analysis. The unfiltered control room in-leakage is modeled with the following flows and filters.

Time Flow (Hours) 0.0 35,200 cfm 0.163 82 cfm Pathway 3: CR Exfiltration This pathway represents the exfiltration from the CR. This flow is the sum of the flows entering the control room (Pathways 1 and 2).

Time Flow (Hours) 0.0 35,200 cfm 0.163 415 cfm Pathway 4: Vaporized PS Leakage (Ruptured SG)

This pathway represents the PS leakage into the ruptured SG that is vaporized. All of the PS leakage is assumed to be vaporized and immediately released to the environment. The leakage rate for this pathway is 150 gallons per day (gpd) based on Technical Specification (TS) 3.4.13.

This leakage rate is based on 62.4 lb/ft3 consistent with Section 5.2 of Appendix F to RG 1.183.

This leakage is assumed to continue into the ruptured SG until the SG is isolated at 70 minutes.

3 62.4 453.59 150 3

= 394 24 60 7.48 Time Flow (Hours) (grams/min) 0.0 394 1.167 0.0

1CAN122101 Page 6 of 15 Pathway 5: Vaporized PS Leakage (Intact SG)

This pathway represents the PS leakage into the intact SG that is vaporized and immediately released to the environment. As discussed for Pathway 4, the flowrate is 150 gpd or 394 grams per minute.

Time Flow Filters (Hours) (grams/min) Aerosol Elemental Organic 0.0 394 0% 0% 0%

237.8 0.0 0% 0% 0%

Pathways 6 & 7: Ruptured Tube Leakage Based on the Framatome T/H analysis, the ruptured tube flow rates are listed below.

Break Flow Rate Time Average Break Flow Start (sec) End (sec) lbs/sec grams/min 0 580 36.83 1.0022E+06 580 840 31.53 8.5823E+05 840 1200 31.34 8.5298E+05 1200 1800 36.15 9.8370E+05 1800 2400 36.45 9.9200E+05 2400 3000 33.33 9.0709E+05 3000 3600 33.88 9.2210E+05 3600 4200 30.51 8.3043E+05 Based on the computational fluid dynamics (CFD) analyses, 40% or less of this tube leakage is expected to be vaporized after 4.4 minutes post-scram (i.e., 14 minutes). Therefore, the breakdown of flows between Pathways 6 and 7 were calculated as follows.

1CAN122101 Page 7 of 15 Time Starting Vaporization Average Break Flow Time Fraction (grams/min)

Start End Hours Vapor Liquid (sec) (sec) (Pathway 6) (Pathway 7) 0 580 0.0000E+00 100% 1.0022E+06 0.0000E+00 580 840 1.6111E-01 100% 8.5823E+05 0.0000E+00 840 1200 2.3333E-01 40% 3.4119E+05 5.1179E+05 1200 1800 3.3333E-01 40% 3.9348E+05 5.9022E+05 1800 2400 5.0000E-01 40% 3.9680E+05 5.9520E+05 2400 3000 6.6667E-01 40% 3.6284E+05 5.4425E+05 3000 3600 8.3333E-01 40% 3.6884E+05 5.5326E+05 3600 4200 1.0000E+00 40% 3.3217E+05 4.9826E+05 4200 1.1667E+00 0.0000E+00 0.0000E+00 While the above table is applied to the iodine and alkali metals, the vaporization fraction for noble gases is modeled to be 100% such that Pathway 7 is always zero and Pathway 6 represents the entire break flow for this nuclide group.

Pathway 8: Ruptured SG Steaming This pathway represents the steaming of the ruptured SG. Prior to receiving the trip signal at 9.6 minutes, the design full power steaming rate is 5.6E+06 lbs/hr/SG.

453.59 5,600,000 = 4.2347 60 As the core decay heat is removed through the ruptured SG, the secondary coolant is vaporized and released via the MSSV or ADV. The steaming rate is modeled to drop after the scram until the ruptured S/G is isolated at 70 minutes.

Time Flow (Hours) (grams/min) 0.0 4.234E+07 0.16 2.5352E+06 1.167 0

1CAN122101 Page 8 of 15 Before the scram, the condenser provides a substantial reduction in the source term release via partitioning and removal via the air removal system. These flows are consequently multiplied by 0.0001 for the iodine and alkali metal cases.

After the scram, these flows are multiplied by 0.01 and 0.001 for the iodine and alkali metal cases, respectively to model the partitioning of iodine and moisture carryover of the alkali metals in the SGs.

Pathway 9: Intact SG Steaming This pathway represents the steaming of the intact SG. As the core decay heat is removed through the single intact SG, the secondary coolant is vaporized and released via the MSSV or ADV. The steaming rate is terminated when the intact SG is isolated.

Time Flow (Hours) (grams/min) 0 4.234E+07 0.16 2.5352E+06 2 3.611E+05 237.8 0 Before the scram, the condenser provides a substantial reduction in the source term release via partitioning and removal via the air removal system. These flows are consequently multiplied by 0.0001 for the iodine and alkali metal cases.

After the scram, these flows are multiplied by 0.01 and 0.001 for the iodine and alkali metal cases, respectively to model the partitioning of iodine and moisture carryover of the alkali metals in the SGs.

Pathway 10: Fuel Leakage for Accident-Induced (AI) Spike This pathway represents the iodine release from the fuel for the AI spike. As described in the source term section, the source term in the core compartment (Compartment 5) is based on the iodine generation rate into the RCS needed to maintain the equilibrium DEI (1.0 Ci/gm) with a 1 gram per minute flow from the core compartment. Consequently, for the accident-induced spike, the flow rate is increased by 335 times the initial rate.

1CAN122101 Page 9 of 15 Time Flow (Hours) (grams/min) 0 335 8 0 2.3 SOURCE TERMS Primary and Secondary Compartments Based on the RCS mass of 2.44E+08 grams and the secondary coolant masses of 2.72E+07 grams/SG, the total activity in each compartment is reported below based on the specific inventories. The secondary coolant is initialized with a source term fraction of 0.50 set in either Compartment 3 or 4 (faulted or intact SG) such that the value in the table below is the total secondary activity.

ANO-1 Reactor Coolant Activities (Ci)

DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary Kr-85m 8.564E+02 Kr-85 3.538E+04 Kr-87 1.459E+03 Kr-88 1.710E+03 Xe-131m 3.928E+05 Xe-133m 7.662E+02 Xe-133 1.928E+04 Xe-135m 3.660E+03 Xe-135 6.564E+03 Xe-137 1.708E+04 Xe-138 4.416E+03 Br-84 9.394E+02 6.691E+01 I-131 1.379E+02 3.052E+00 I-132 4.880E+02 1.083E+01 I-133 4.246E+02 9.520E+00 I-134 7.149E+02 1.665E+01

1CAN122101 Attachment 2 Page 10 of 15 DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary I-135 5.978E+02 1.365E+01 Rb-88 3.636E+01 1.431E+00 Cs-134 7.198E+01 2.900E+00 Cs-135 1.103E+01 4.477E-01 Cs-137 4.587E+01 1.246E+00 Fuel Sources The fuel source is applied for the AI spike case. The core compartment is initialized with this source term to represent the release rate from the core based on a 1 gram/minute release to offset the system losses. To model a 335-fold iodine spike, Pathway 10 is modeled with a 335 gram/minute flow rate for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The source term for the core compartment is calculated by summing the losses from letdown, radioactive decay and PS leakage for an initial coolant activity at 1 Ci/g DEI. Then, with a compartment volume of 1E+10 grams and a 1 gram/minute flow (Pathway 10), the core inventories can be calculated by merely multiplying the loss rate (in Ci/minute) by 1E+10.

Decay Total Initial Letdown Leakage Loss Rate Loss Rate Source Isotope Constant Activity (min-1) (min-1) (Ci/min) (Ci)

(min-1) (min-1) (Ci)

I-131 1.360E-03 3.210E-06 5.984E-05 1.423E-03 1.379E+02 1.962E-01 1.962E+09 I-132 1.360E-03 3.210E-06 5.023E-03 6.386E-03 4.880E+02 3.116E+00 3.116E+10 I-133 1.360E-03 3.210E-06 5.553E-04 1.918E-03 4.246E+02 8.146E-01 8.146E+09 I-134 1.360E-03 3.210E-06 1.316E-02 1.452E-02 7.149E+02 1.038E+01 1.038E+11 I-135 1.360E-03 3.210E-06 1.747E-03 3.111E-03 5.978E+02 1.860E+00 1.860E+10 The iodine species of the initial and spike activity is assumed to the 97% elemental and 3% organic.

2.4 OTHER FACTORS Exclusion Area Boundary (EAB)

Per Section 4.1.3 of RG 1.183, the breathing rate of persons offsite is assumed to be 3.5E-04 m3/s.

1CAN122101 Page 11 of 15 The EAB dispersion coefficient is unchanged from the current value. Although it is a 2-hour value, it is modeled throughout the entire accident duration so that the worst-case 2-hour sliding window can be accurately identified.

Time /Q value (Hours) (s/m3) 0.0 6.8E-04 720 0.0 Low Population Zone (LPZ)

Per Section 4.1.3 of RG 1.183, for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons at this location is assumed to be 3.5E-04 m3/s. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is 1.8E-04 m3/s. After that and until the end of the accident, the rate is assumed to be 2.3E-04 m3/s.

Time Breathing (Hours) Rate (m3/s) 0.0 3.5E-04 8 1.8E-04 24 2.3E-04 720 0.0 The LPZ dispersion coefficients are unchanged from the current values.

Time /Q value (Hours) (s/m3) 0.0 1.1E-04 8 1.1E-05 24 4.0E-06 96 1.3E-06 720 0.0 Control Room Per Section 4.2.6 of RG 1.183, the dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual is assumed to be 3.5E-04 m3/s.

1CAN122101 Page 12 of 15 Time Occupancy Breathing (Hours) Factor Rate (m3/s) 0.0 1.0 24 0.6 3.5E-04 96 0.4 720 0 0.0 The dispersion coefficients are based on the release point and the receptor locations. The control room /Qs applicable to this event are listed below for the worst-case (closest) intake.

These values represent the closest MSSV or ADV to the nearest control room intake in either loop.

Control Room Dispersion Coefficients (s/m3)

Time Main Steam Atmospheric (Hours) Safety Valve Dump Valve 0.0 1.90E-02 4.10E-03 2 1.23E-02 2.59E-03 8 5.83E-03 1.12E-03 24 3.80E-03 8.32E-04 96 3.10E-03 5.91E-04 720 0 0 The condenser releases are modeled with the same /Q as the ADV. In order to maximize the releases, the normally-closed motor-operated block valve isolating the ADV on the intact loop fails to open so that the releases are via the MSSV after the scram until an operator manually opens the block valve at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The associated /Qs modeled to the environment are listed below.

1CAN122101 Page 13 of 15 Control Room Dispersion Coefficients (s/m3)

Time MSSV/ADV Description (Hours) 0.0 4.10E-03 Condenser release -

same /Qs as ADV 0.16 1.90E-02 MSSV release upon scram 1.0 4.10E-03 ADV release 2 2.59E-03 8 1.12E-03 24 8.32E-04 96 5.91E-04 720 0 3.0 RESULTS RADTRAD 3.03 was executed for the six cases identified above based on an initial DEI of 1.0 µCi/g in the RCS and secondary coolant DEI of 0.1 µCi/g.

Base RADTRAD Results for Each Case (Rem TEDE)

DEI = 1.0 Ci/g (RCS) and 0.1 Ci/g (Secondary)

RADTRAD RESULTS File Name Description EAB LPZ CR Airborne Dose iodine release from secondary 1 sec_iodine side 4.3784E-03 8.7894E-04 7.1424E-03 alkali release from secondary 2 sec_alkali side 3.9997E-04 9.0236E-05 1.6387E-03 3 rcs_iodine iodine release from rcs 2.2519E-01 3.6608E-02 7.0076E-01 4 rcs_alkali alkali release from rcs 1.0635E-01 1.7343E-02 4.0931E-01 5 rcs_ngas noble gas release from rcs 3.4516E-01 5.5919E-02 2.7679E-01 additional iodine release from rcs in accident-induced spike 7.7425E+00 1.2968E+00 7.4640E+00 6 ai_spike case

1CAN122101 Page 14 of 15 Cloud Dose 1 c_sec_iodine 3.5329E-05 2 c_sec_alkali 8.0403E-08 3 c_rcs_iodine 1.2545E-03 4 c_rcs_alkali 1.5199E-05 5 c_rcs_ngas 1.0418E-02 6 c_ai_spike 3.3945E-02 To develop appropriate DEI limits for the TSs, these results were ratioed. A reduction of the pre-accident spike from 60 to 6 µCi/g was determined to be sufficient to reduce the control room dose to below 5 rem TEDE. For the pre-existing spike case, the doses from the iodine releases from the RCS above (which are based on a DEI of 1.0 µCi/g) were multiplied by a factor of 6 representing an initial DEI of 6 µCi/g. The control room dose set the limit for this case. For the AI spike case, the initial iodine activity was reduced to 0.25 µCi/g to meet the acceptance criterion at the EAB.

Pre-Existing Spike RADTRAD Results for Each Case (Rem TEDE)

DEI = 6.0 Ci/g (RCS) and 0.1 Ci/g (Secondary)

RADTRAD (DEI = 1.0) PRE-EXISTING SPIKE DOSE RESULTS (REM TEDE) 6.00 DOSE RESULTS (REM TEDE)

Airborne Dose EAB LPZ CR EAB LPZ CR 1 iodine release from secondary side 4.3784E-03 8.7894E-04 7.1424E-03 1.00 4.3784E-03 8.7894E-04 7.1424E-03 2 alkali release from secondary side 3.9997E-04 9.0236E-05 1.6387E-03 1.00 3.9997E-04 9.0236E-05 1.6387E-03 3 iodine release from rcs 2.2519E-01 3.6608E-02 7.0076E-01 6.00 1.3511E+00 2.1965E-01 4.2046E+00 4 alkali release from rcs 1.0635E-01 1.7343E-02 4.0931E-01 1.00 1.0635E-01 1.7343E-02 4.0931E-01 5 noble gas release from rcs 3.4516E-01 5.5919E-02 2.7679E-01 1.00 3.4516E-01 5.5919E-02 2.7679E-01 6 additional iodine release from rcs in ai spike case 7.7425E+00 1.2968E+00 7.4640E+00 0.00 0.0000E+00 0.0000E+00 0.0000E+00 Cloud Dose to Control Room 1 iodine release from secondary side 3.5329E-05 1.00 0.0000E+00 0.0000E+00 3.5329E-05 2 alkali release from secondary side 8.0403E-08 1.00 0.0000E+00 0.0000E+00 8.0403E-08 3 iodine release from rcs 1.2545E-03 6.00 0.0000E+00 0.0000E+00 7.5270E-03 4 alkali release from rcs 1.5199E-05 1.00 0.0000E+00 0.0000E+00 1.5199E-05 5 noble gas release from rcs 1.0418E-02 1.00 0.0000E+00 0.0000E+00 1.0418E-02 6 additional iodine release from rcs in ai spike case 3.3945E-02 0.00 0.0000E+00 0.0000E+00 0.0000E+00 1.8074E+00 2.9388E-01 4.9174E+00

1CAN122101 Page 15 of 15 Accident-Induced Spike RADTRAD Results for Each Case (Rem TEDE)

DEI = 0.25 Ci/g (RCS) and 0.1 Ci/g (Secondary)

RADTRAD (DEI = 1.0) ACCIDENT-INDUCED SPIKE DOSE RESULTS (REM TEDE) 0.25 DOSE RESULTS (REM TEDE)

Airborne Dose EAB LPZ CR EAB LPZ CR 1 iodine release from secondary side 4.3784E-03 8.7894E-04 7.1424E-03 1.00 4.3784E-03 8.7894E-04 7.1424E-03 2 alkali release from secondary side 3.9997E-04 9.0236E-05 1.6387E-03 1.00 3.9997E-04 9.0236E-05 1.6387E-03 3 iodine release from rcs 2.2519E-01 3.6608E-02 7.0076E-01 0.25 5.6298E-02 9.1520E-03 1.7519E-01 4 alkali release from rcs 1.0635E-01 1.7343E-02 4.0931E-01 1.00 1.0635E-01 1.7343E-02 4.0931E-01 5 noble gas release from rcs 3.4516E-01 5.5919E-02 2.7679E-01 1.00 3.4516E-01 5.5919E-02 2.7679E-01 6 additional iodine release from rcs in ai spike case 7.7425E+00 1.2968E+00 7.4640E+00 0.25 1.9356E+00 3.2420E-01 1.8660E+00 Cloud Dose to Control Room 1 iodine release from secondary side 3.5329E-05 1.00 0.0000E+00 0.0000E+00 3.5329E-05 2 alkali release from secondary side 8.0403E-08 1.00 0.0000E+00 0.0000E+00 8.0403E-08 3 iodine release from rcs 1.2545E-03 0.25 0.0000E+00 0.0000E+00 3.1363E-04 4 alkali release from rcs 1.5199E-05 1.00 0.0000E+00 0.0000E+00 1.5199E-05 5 noble gas release from rcs 1.0418E-02 1.00 0.0000E+00 0.0000E+00 1.0418E-02 6 additional iodine release from rcs in ai spike case 3.3945E-02 0.25 0.0000E+00 0.0000E+00 8.4863E-03 1.8074E+00 4.0758E-01 2.4482E+00 2.7553E+00

1CAN122101 Attachment 3 Main Steam Line Break (MSLB)

Model Information

1CAN122101 Page 1 of 12 MAIN STEAM LINE BREAK (MSLB)

MODEL INFORMATION 1.0 EVENT DESCRIPTION While operating at full power, the double-ended rupture of a main steam line outside containment occurs with unrestricted discharge from each end. Upon detection of a steam line break, the affected steam generator (SG) is isolated by closing its main steam isolation valve (MSIV). The reactor scrams on low Reactor Coolant System (RCS) pressure. The entire SG inventory on the affected loop is assumed to be released by the time the MSIV is fully closed.

Due to the Loss-of-Offsite Power (LOOP), the reactor coolant pumps (RCPs) trip, the main feedwater (MFW) pumps trip, and the condenser becomes unavailable. Emergency feedwater (EFW) actuates automatically to raise steam generator liquid levels. Closure of the turbine stop valves (TSVs) causes the main steam line pressure to increase and open the atmospheric dump valves (ADVs) and main steam safety valves (MSSVs). The fission products escaping from the ADVs and/or MSSVs are released directly to the atmosphere.

The MSLB scenario considers a single active failure and a coincident loss-of-offsite power. The accident chronology applied in this analysis is reported below.

MSLB Chronology Time Event Comments 0 seconds MSLB occurs outside containment Coincident LOOP

~0 seconds Reactor trip ADV block valve on intact SG fails to open (single active failure); all releases are through the MSSVs.

10 seconds Control Room Isolated 1 minute Faulted SG secondary side Operators notice failed block valve completely released and dispatch an operator to correct MSIVs are closed. the issue in the field.

30 minutes Failed ADV manually opened Cooldown is started using the intact SG 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Decay Heat Removal (DHR) conditions reached. Primary-to-Secondary (PS) leakage from intact loop terminated

1CAN122101 Page 2 of 12 Time Event Comments 251.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCS temperature below 212 ºF.

PS leakage into faulted loop is terminated.

2.0 RADIOLOGICAL CONSEQUENCE MODEL The radiological consequences of the MSLB accident were then analyzed with RADTRAD 3.03.

Due to code limitations, the scenario was broken down into nine cases, which were evaluated separately and summed to develop to total dose. For each case, the impact on control room dose from the plume was also evaluated using modified isotopic-specific dose conversion factors and was shown to be negligible. These cases are listed in the table below.

ANO-1 MSLB Cases Description 1 Secondary coolant iodine release from faulted SG 2 Secondary coolant iodine release from intact SG 3 Secondary coolant alkali metal release from faulted SG 4 Secondary coolant alkali metal release from intact SG 5 Primary coolant iodine release due to PS leakage 6 Primary coolant alkali metal release due to PS leakage 7 Primary coolant noble gas release due to PS leakage Primary coolant iodine release from pre-existing spike due to PS 8

leakage Primary coolant iodine release from accident-induced spike due to PS 9

leakage The RADTRAD model developed to evaluate these cases is illustrated below.

1CAN122101 Page 3 of 12 Recirc Filtered Intake CR 1 Intake 3 1

Control CR 2 Room Intake Unfiltered Intake 6 6 6 Environment Environment Environment 5

(MSSV/ADV) (MSSV/ADV) (Main Steam Steam/Water 5 Pipe)

Partitioning & Fuel Carryover Vaporized 9

P/S leakage Vaporized Fuel Leakage P/S leakage 7 For Accident 10 4 Steaming Faulted SG Induced Spike Blowdown 4 2 3 Intact Primary Faulted S/G 8 System S/G 6

Liquid Liquid Liquid Liquid P/S leakage P/S leakage 2.1 COMPARTMENTS The volumes of the compartments are listed below. Since the RCS compartments are typically reported in terms of mass, the RADTRAD runs applied the masses of these compartments in grams (rather than cubic feet [ft3]) with corresponding flows between the compartments in grams/minute (rather than cubic feet per minute [cfm]).

Compartment Name Volume Comments 1 Control Room 4.00E+05 ft 3 2 RCS 2.44E+08 g Equivalent to 5.37E+05 pounds (lbs) 3 Ruptured SG Liquid 2.72E+07 g Equivalent to 6.00E+04 lbs 4 Intact SG Liquid 2.72E+07 g Large enough to minimize depletion 5 Fuel 1.00E+10 g effects over the 8-hour iodine spike

1CAN122101 Page 4 of 12 The control room is modeled with a recirculating filter that starts upon isolation at ten seconds after the scram with the following parameters.

Time Flow Filters Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

10 sec 1667 cfm 99% 95% 95%

2.2 PATHWAYS Pathway 1: Filtered Control Room (CR) Intake This pathway represents the filtered intake into the CR. This pathway starts 10 seconds after the scram and is modeled with the following flows and filters.

Time Flow Filters Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

10 sec 333 cfm 99% 99% 99%

Pathway 2: Un-Filtered CR In-leakage This pathway represents the unfiltered intake into the CR. The ANO-1 design basis in-leakage rate of 82 cfm was applied in this analysis. The unfiltered control room in-leakage is modeled with the following flows and filters.

Time Flow 0.0 35,200 cfm 10 sec 82 cfm Pathway 3: CR Exfiltration This pathway represents the exfiltration from the CR. This flow is the sum of the flows entering the CR (Pathways 1 and 2).

Time Flow 0.0 35,200 cfm 10 sec 415 cfm

1CAN122101 Page 5 of 12 Pathway 4: Vaporized PS Leakage (Faulted SG)

This pathway represents the PS leakage into the faulted SG that is vaporized. Since all of the PS leakage is assumed to be vaporized, this flow represents all of the PS leakage into the faulted SG. The leakage rate for this pathway is 0.5 gallons per minute (gpm). Since both PS leakage paths have identical vaporization fractions (i.e., 100%) and /Qs, the total 1 gpm leakage from Technical Specification (TS) 5.5.9.b.2 is simply split between the SGs.

As described in Section 5.2 of Appendix E to Regulatory Guide (RG) 1.183, the density of this PS leakage should be based on a cooled liquid at a density of 62.4 pound mass per cubic foot (lbm/ft3). On these bases, the 0.5 gpm of PS leakage is calculated to be 1892 grams per minute. This leakage is assumed to continue into the faulted SG until the RCS has been cooled to below 212 ºF at 251.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3 62.4 453.59 0.5 = 1892 7.48 3 Time Flow (Hours) 0.0 1892 grams/min 251.8 0.0 grams/min Pathway 5: Vaporized PS Leakage (Intact SG)

This pathway represents the PS leakage into the intact SG that is vaporized. Since all of the PS leakage is assumed to be vaporized, this flow represents all of the PS leakage into the intact S/G.

As discussed for Pathway 4, the flowrate is 0.5 gpm or 1892 grams per minute. This leakage is assumed to continue into the intact SG until the RCS has been cooled to decay heat removal entry conditions at 237.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Time Flow (Hours) 0.0 1892 grams/min 237.8 0.0 grams/min Pathways 6 & 8: Liquid PS Leakage These pathways are no longer applied since this analysis assumes 100% vaporization of all P/S leakage.

1CAN122101 Page 6 of 12 Time Flow (Hours) 0.0 0.0 grams/min Pathway 7: Faulted SG Blowdown This pathway represents the blowdown of the faulted loop. Once the main steam line breaks, the inventory in the faulted loop is assumed to be released to the environment through this pathway within the first minute. Consequently, this pathway has a very high flow rate for the first minute (0.0167 hours0.00193 days <br />0.0464 hours <br />2.761243e-4 weeks <br />6.35435e-5 months <br />). This flow would equate to 2.72E+07 gram/min. To ensure that all the inventory is released, this flow is increased to 2.72E+10 for an additional 2 seconds.

Time Flow (Hours) 0.0 2.72E+07 grams/min 0.0167 2.72E+10 grams/min 0.01722 0.0 grams/min Pathway 9: Intact SG Steaming This pathway represents the steaming of the intact SG. As the core decay heat is removed through the single intact SG, the secondary coolant is vaporized and released via the MSSV or ADV.

Time Flow (Hours) 0.0 2.5352E+06 grams/min 2.0 3.611E+05 grams/min 237.8 0.0 After the scram, these flows are multiplied by 0.01 and 0.001 for the iodine and alkali metal cases, respectively to model the partitioning of iodine and moisture carryover of the alkali metals.

Pathway 10: Fuel Leakage for Accident-Induced (AI) Spike This pathway represents the iodine release from the fuel for the accident-induced spike. As described in the source term section, the source term in the core compartment (Compartment 5) is based on the iodine generation rate into the RCS needed to maintain the equilibrium DEI

1CAN122101 Page 7 of 12 (1.0 Ci/gm) with a 1 gram per minute flow from the core compartment. Consequently, for the accident-induced spike, the flow rate is increased by 500 times the initial rate.

Time Flow (Hours) (grams/min) 0 500 8 0 2.3 SOURCE TERMS Primary and Secondary Compartments Based on the RCS mass of 2.44E+08 grams and the secondary coolant masses of 2.72E+07 grams/SG, the total activity in each compartment is reported below based on the specific activities. The secondary coolant is initialized with a source term fraction of 0.50 set in either Compartment 3 or 4 (faulted or intact SG) such that the value in the table below is the total secondary activity.

ANO-1 Reactor Coolant Activities (Ci)

DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary Kr-85m 8.564E+02 Kr-85 3.538E+04 Kr-87 1.459E+03 Kr-88 1.710E+03 Xe-131m 3.928E+05 Xe-133m 7.662E+02 Xe-133 1.928E+04 Xe-135m 3.660E+03 Xe-135 6.564E+03 Xe-137 1.708E+04 Xe-138 4.416E+03 Br-84 9.394E+02 6.691E+01 I-131 1.379E+02 3.052E+00 I-132 4.880E+02 1.083E+01

1CAN122101 Attachment 3 Page 8 of 12 DEI DEX Isotope 1.0 0.1 2200 Primary Secondary Primary I-133 4.246E+02 9.520E+00 I-134 7.149E+02 1.665E+01 I-135 5.978E+02 1.365E+01 Rb-88 3.636E+01 1.431E+00 Cs-134 7.198E+01 2.900E+00 Cs-135 1.103E+01 4.477E-01 Cs-137 4.587E+01 1.246E+00 Fuel Sources The fuel source is applied for the AI spike case. The core compartment is initialized with this source term to represent the release rate from the core based on a 1 gram/min release to offset the system losses. To model a 500-fold iodine spike, Pathway 10 is modeled with a 500 gram/min flow rate for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> The source term for the core compartment is calculated by summing the losses from letdown, radioactive decay and PS leakage for an initial coolant activity at 1.0 Ci/g DEI. Then, with a compartment volume of 1E+10 grams and a 1 gram/min flow (Pathway 10), the core inventories can be calculated by merely multiplying the loss rate (in Ci/min) by 1E+10.

Total Decay Loss Initial Letdown Leakage Loss Rate Source Isotope Constant Rate Activity (min-1) (min-1) (Ci/min) (Ci)

(min-1) (Ci)

(min-1)

I-131 1.360E-03 3.210E-06 5.984E-05 1.423E-03 1.379E+02 1.962E-01 1.962E+09 I-132 1.360E-03 3.210E-06 5.023E-03 6.386E-03 4.880E+02 3.116E+00 3.116E+10 I-133 1.360E-03 3.210E-06 5.553E-04 1.918E-03 4.246E+02 8.146E-01 8.146E+09 I-134 1.360E-03 3.210E-06 1.316E-02 1.452E-02 7.149E+02 1.038E+01 1.038E+11 I-135 1.360E-03 3.210E-06 1.747E-03 3.111E-03 5.978E+02 1.860E+00 1.860E+10 The iodine species of the initial and spike activity is assumed to the 97% elemental and 3% organic.

1CAN122101 Page 9 of 12 2.4 OTHER FACTORS Exclusion Area Boundary (EAB)

Per Section 4.1.3 of RG 1.183, the breathing rate of persons offsite is assumed to be 3.5E-04 m3/s.

The EAB dispersion coefficient is unchanged from the current value. Although it is a 2-hour value, it is modeled throughout the entire accident duration so that the worst-case 2-hour sliding window can be accurately identified.

Time /Q value (Hours) (s/m3) 0.0 6.8E-04 720 0.0 Low Population Zone (LPZ)

Per Section 4.1.3 of RG 1.183, for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons at this location is assumed to be 3.5E-04 m3/s. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is 1.8E-04 m3/s. After that and until the end of the accident, the rate is assumed to be 2.3E-04 m3/s.

Time Breathing (Hours) Rate (m3/s) 0.0 3.5E-04 8 1.8E-04 24 2.3E-04 720 0.0

1CAN122101 Page 10 of 12 The LPZ dispersion coefficients are unchanged from the current values.

Time /Q value (Hours) (s/m3) 0.0 1.1E-04 8 1.1E-05 24 4.0E-06 96 1.3E-06 720 0.0 Control Room Per Section 4.2.6 of RG 1.183, the dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual is assumed to be 3.5E-04 m3/s.

Time Occupancy Breathing (Hours) Factor Rate (m3/s) 0.0 1.0 24 0.6 3.5E-04 96 0.4 720 0 0.0 The dispersion coefficients are based on the release point and the receptor locations. The control room /Qs applicable to this event are listed below for the worst-case (closest) intake.

These values represent the closest MSSV, ADV, or main steam pipe to the nearest control room intake in either loop.

1CAN122101 Page 11 of 12 Control Room Dispersion Coefficients (s/m3)

Time Main Steam Atmospheric Main Steam (Hours) Safety Valve Dump Valve Pipe 0.0 1.90E-02 4.10E-03 3.15E-03 2 1.23E-02 2.59E-03 2.16E-03 8 5.83E-03 1.12E-03 8.90E-04 24 3.80E-03 8.32E-04 6.61E-04 96 3.10E-03 5.91E-04 5.01E-04 720 0 0 0 Since the MSIV is expected to close in the faulted loop, the only release out of the main steam pipe would be the secondary coolant in the faulted loop. This dispersion value is consequently applied for the release of the secondary coolant in the faulted loop in Cases 1 and 3.

If the MSIV in the faulted loop failed to close, the vaporized RCS PS leakage would also be released from the main steam pipe; however, since the main steam pipe dispersion coefficients are lower than the ADV and MSSV values, this would not be a bounding single active failure.

Instead, the normally-closed motor-operated block valve isolating the ADV on the intact loop fails to open so that the releases are via the MSSV until an operator manually opens the block valve at 30 minutes.

Modeled Control Room Dispersion Coefficients (s/m3)

Time Main Steam MSSV/ADV (Hours) Pipe 0.0 3.15E-03 1.90E-02 0.5 4.10E-03 2 2.59E-03 8 1.12E-03 24 8.32E-04 96 5.91E-04 720 0 0 3.0 RESULTS RADTRAD 3.03 was executed for the nine cases identified above. The initial DEI for the pre-accident spike case was 60.0 µCi/g in the RCS and secondary coolant DEI of 0.1 µCi/g.

1CAN122101 Attachment 3 Page 12 of 12 The initial DEI for the accident-induced spike case was 1.0 µCi/g in the RCS and secondary coolant DEI of 0.1 µCi/g.

ACCIDENT-INDUCED SPIKE (DEI=1 PRE-EXISTING SPIKE (DEI=60 µCi/g)

µCi/g)

File Name EAB LPZ CR EAB LPZ CR Airborne Dose 1 sec_iodine_faulted 3.5890E-02 5.8057E-03 2.4879E-01 3.5890E-02 5.8057E-03 2.4879E-01 2 sec_iodine_intact 2.9590E-03 6.4807E-04 2.9885E-03 2.9590E-03 6.4807E-04 2.9885E-03 3 sec_alkali_faulted 2.1332E-02 3.4508E-03 2.3006E-01 2.1332E-02 3.4508E-03 2.3006E-01 4 sec_alkali_intact 2.3636E-04 6.3784E-05 3.4517E-04 2.3636E-04 6.3784E-05 3.4517E-04 5 rcs_iodine 4.2820E-03 2.8573E-03 9.8209E-03 4.2820E-03 2.8573E-03 9.8209E-03 6 rcs_alkali 2.1764E-03 2.0108E-03 8.5908E-03 2.1764E-03 2.0108E-03 8.5908E-03 7 rcs_ngas 2.3656E-03 1.4162E-03 2.6481E-03 2.3656E-03 1.4162E-03 2.6481E-03 8 pe_spike 2.5264E-01 1.6858E-01 5.7943E-01 0.0000E+00 0.0000E+00 0.0000E+00 9 ai_spike 0.0000E+00 0.0000E+00 0.0000E+00 1.1867E+00 6.9882E-01 2.0445E+00 Cloud Dose to Control Room 1 c_sec_iodine_faulted 1.5381E-04 1.5381E-04 2 c_sec_iodine_intact 2.5039E-05 2.5039E-05 3 c_sec_alkali_faulted 1.5155E-06 1.5155E-06 4 c_sec_alkali_intact 4.7938E-08 4.7938E-08 5 c_rcs_iodine 2.2257E-05 2.2257E-05 6 c_rcs_alkali 1.5658E-06 1.5658E-06 7 c_rcs_ngas 6.3859E-05 6.3859E-05 8 c_pe-spike 1.3132E-03 0.0000E+00 9 c_ai-spike 0.0000E+00 3.6689E-03 Total 3.2188E-01 1.8483E-01 1.0843E+00 1.2559E+00 7.1507E-01 2.5517E+00

1CAN122101 Attachment 4 Control Rod Ejection Accident (CREA)

Model Information

1CAN122101 Page 1 of 9 CONTROL ROD EJECTION ACCIDENT (CREA)

MODEL INFORMATION 1.0 EVENT DESCRIPTION While operating at full power, a high-worth control rod is quickly ejected from the core, resulting in a reactor trip on high flux and a substantial number of fuel rods entering Departure from Nucleate Boiling (DNB). These rods are assumed to fail and release their gap activity to either the primary system or the reactor building. A loss-of-offsite power (LOOP) is assumed concurrent with the scram, making the condenser unavailable for decay heat rejection. The atmospheric dump valves (ADVs) and main steam safety valves (MSSVs) are used to cool down the plant. The fission products escaping from the ADVs and/or MSSVs are released directly to the atmosphere. The accident chronology applied in this analysis is reported below for each potential leakage path.

CREA Chronology - Primary-to-Secondary (PS) Leakage Pathway Time Event Comments 0 seconds CREA occurs Reactor trip on high neutron flux Source term release begins LOOP occurs ADV block valve on one SG fails to open (single active failure) releases are through the MSSVs for the affected loop 10 sec Control Room isolated on high radiation Operators notice failed block in intakes valve and dispatch an Source term release into the coolant operator to correct the issue in ends the field.

30 minutes Failed ADV block valve manually opened. Releases from the affected loop are via the ADV.

38.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> Shutdown cooling is initiated.

Releases are terminated CREA Chronology - Containment Leakage Pathway Time Event Comments 0 seconds CREA occurs Reactor trip on high neutron flux Source term release begins LOOP occurs

1CAN122101 Page 2 of 9 Time Event Comments 10 seconds Source term release into the containment ends 10 seconds Control Room isolated on high radiation in intakes 30 days Releases are terminated 2.0 RADIOLOGICAL CONSEQUENCE MODEL The radiological consequences of the CREA accident were then analyzed with RADTRAD 3.03.

For each case, the impact on control room dose from the plume was also evaluated using modified isotopic-specific dose conversion factors and was shown to be negligible. The RADTRAD model developed to evaluate these cases is illustrated below.

Recirc Filtered Intake CR 1 Intake 3 1

Control CR 2 Room Intake Unfiltered Intake 4 3 5

Environment Containment (CMT CMT Leakage) Leakage 4 2 Environment 4 Primary (MSSV/ADV) Vaporized System P/S leakage 2.1 COMPARTMENTS The volumes of the compartments are listed below. Since the RCS compartment is typically reported in terms of mass, the RADTRAD runs applied the mass of this compartment in grams

1CAN122101 Page 3 of 9 (rather than cubic feet [ft3]) with corresponding flows from this compartment in grams/minute (rather than cubic feet per minute [cfm]).

Compartment Name Volume Comments 1 Control Room 4.00E+05 ft3 2 RCS 2.44E+08 g Equivalent to 5.37E+05 pounds (lbs) 3 Containment 1.81E+06 ft3 The control room is modeled with a recirculating filter that starts upon isolation at ten seconds after the scram with the following parameters.

Time Flow Filters Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

10 sec 1667 cfm 99% 95% 95%

2.2 PATHWAYS Pathway 1: Filtered Control Room (CR) Intake This pathway represents the filtered intake into the CR. This pathway starts ten seconds after the scram and is modeled with the following flows and filters.

Time Flow Filters Aerosol Elemental Organic 0.0 0 cfm 0% 0% 0%

10 sec 333 cfm 99% 99% 99%

Pathway 2: Un-Filtered CR In-leakage This pathway represents the unfiltered intake into the CR. The ANO-1 design basis in-leakage rate of 82 cfm was applied in this analysis. The unfiltered CR in-leakage is modeled with the following flows and filters.

Time Flow 0.0 35,200 cfm 10 sec 82 cfm

1CAN122101 Page 4 of 9 Pathway 3: CR Exfiltration This pathway represents the exfiltration from the CR. This flow is the sum of the flows entering the CR (Pathways 1 and 2).

Time Flow 0.0 35,200 cfm 10 sec 415 cfm Pathway 4: Vaporized PS Leakage This pathway represents the total PS leakage into both steam generators (SGs). All of this PS leakage is assumed to be vaporized and released. The leakage rate for this pathway is 150 gallons per day (gpd) per SG based on Technical Specification (TS) 3.4.13. This leakage rate is based on 62.4 lb/ft3 consistent with Section 7.2 of Appendix H to Regulatory Guide (RG) 1.183. This leakage is assumed to continue until the reactor is brought to cold shutdown consistent with Table 6 to RG 1.183. This shutdown time is 38.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> with both SGs available.

3 62.4 453.59 2 150 3

= 788 24 60 7.48 Time Flow (Hours) (grams/min) 0.0 788 38.25 0.0 Pathway 5: Containment Leakage This pathway represents the containment leakage into the environment. This rate is 0.2% of the containment air weight per day at a pressure of Pa per ANO-1 TS 5.5.16. This rate is applied in this calculation for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent with Section 6.2 of Appendix H to RG 1.183, this leakage rate drops by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This leakage is assumed to continue for 30 days consistent with Table 6 to RG 1.183.

Time Flow (Hours) (percent/day) 0.0 0.2 24 0.1 720 0.0

1CAN122101 Page 5 of 9 2.3 SOURCE TERMS The source terms are based on 14% of the core departing from nucleate boiling. These control rods were assumed to have a bounding peaking factor of 1.8. Consistent with Section 1 of Appendix H to RG 1.183, the gap fractions for the iodine and noble gases are 10%. Since Appendix H to RG 1.183 is silent on the applicable gap fraction for alkali metals, this calculation applies a value of 12% consistent with non-LOCA gap fractions in Table 3 of RG 1.183. These released source terms are reproduced below.

ANO-1 CREA Released Activity (Ci)

Fraction of Core Core Peaking Gap Released Isotope Activity (Ci) Damaged Factor Fraction Activity (Ci)

Kr-83m 8.77E+06 14.00% 1.8 10.00% 2.210E+05 Kr-85 9.61E+05 14.00% 1.8 10.00% 2.422E+04 Kr-85m 1.90E+07 14.00% 1.8 10.00% 4.788E+05 Kr-87 3.73E+07 14.00% 1.8 10.00% 9.400E+05 Kr-88 5.01E+07 14.00% 1.8 10.00% 1.263E+06 Xe-131m 7.55E+05 14.00% 1.8 10.00% 1.903E+04 Xe-133 1.48E+08 14.00% 1.8 10.00% 3.730E+06 Xe-133m 4.60E+06 14.00% 1.8 10.00% 1.159E+05 Xe-135 3.51E+07 14.00% 1.8 10.00% 8.845E+05 Xe-135m 3.09E+07 14.00% 1.8 10.00% 7.787E+05 Xe-138 1.27E+08 14.00% 1.8 10.00% 3.200E+06 I-130 1.36E+06 14.00% 1.8 10.00% 3.427E+04 I-131 7.22E+07 14.00% 1.8 10.00% 1.819E+06 I-132 1.05E+08 14.00% 1.8 10.00% 2.646E+06 I-133 1.48E+08 14.00% 1.8 10.00% 3.730E+06 I-134 1.67E+08 14.00% 1.8 10.00% 4.208E+06 I-135 1.41E+08 14.00% 1.8 10.00% 3.553E+06 Cs-134 1.46E+07 14.00% 1.8 12.00% 4.415E+05 Cs-136 2.98E+06 14.00% 1.8 12.00% 9.012E+04 Cs-137 9.88E+06 14.00% 1.8 12.00% 2.988E+05 Cs-138 1.38E+08 14.00% 1.8 12.00% 4.173E+06 Rb-86 1.29E+05 14.00% 1.8 12.00% 3.901E+03

1CAN122101 Page 6 of 9 The iodine species of the released activity is assumed to the 97% elemental and 3% organic for the RCS release. For the containment release, the iodine species is assumed to be 95% aerosol, 4.85% elemental, and 0.15% organic.

2.4 OTHER FACTORS Exclusion Area Boundary (EAB)

Per Section 4.1.3 of RG 1.183, the breathing rate of persons offsite is assumed to be 3.5E-04 m3/s.

The EAB dispersion coefficient is unchanged from the current value. Although it is a 2-hour value, it is modeled throughout the entire accident duration so that the worst-case 2-hour sliding window can be accurately identified.

Time /Q value (Hours) (s/m3) 0.0 6.8E-04 720 0.0 Low Population Zone (LPZ)

Per Section 4.1.3 of RG 1.183, for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons at this location is assumed to be 3.5E-04 m3/s. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is 1.8E-04 m3/s. After that and until the end of the accident, the rate is assumed to be 2.3E-04 m3/s.

Time Breathing (Hours) Rate (m3/s) 0.0 3.5E-04 8 1.8E-04 24 2.3E-04 720 0.0 The LPZ dispersion coefficients are unchanged from the current values.

1CAN122101 Page 7 of 9 Time /Q value (Hours) (s/m3) 0.0 1.1E-04 8 1.1E-05 24 4.0E-06 96 1.3E-06 720 0.0 Control Room Per Section 4.2.6 of RG 1.183, the dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual is assumed to be 3.5E-04 m3/s.

Time Occupancy Breathing (Hours) Factor Rate (m3/s) 0.0 1.0 24 0.6 3.5E-04 96 0.4 720 0 0.0 The dispersion coefficients are based on the release point and the receptor locations. The control room /Qs applicable to this event are listed below for the worst-case (closest) intake.

These values represent the closest MSSV or ADV to the nearest control room intake in either loop. For the containment release, the highest values of the dispersion factors for a diffuse release from the ANO-1 containment to either control room intake are applied.

1CAN122101 Page 8 of 9 Control Room Dispersion Coefficients (s/m3)

Time Main Steam Atmospheric Containment (Hours) Safety Valve Dump Valve Release 0.0 1.90E-02 4.10E-03 3.55E-03 2 1.23E-02 2.59E-03 2.49E-03 8 5.83E-03 1.12E-03 9.85E-04 24 3.80E-03 8.32E-04 8.30E-04 96 3.10E-03 5.91E-04 6.31E-04 720 0 0 0 In order to maximize control room doses, the normally-closed motor-operated block valve on one loop is assumed to fail to open so that the releases are via the MSSV after the scram until an operator manually opens the block valve after 30 minutes. Since the releases from each loop are approximately equal, the control room /Q for the first 30 minutes would be the average of the MSSV and ADV values.

Control Room Dispersion Coefficients (s/m3)

Time MSSV/ADV Containment (Hours) Release 0.0 1.155E-02 3.55E-03 0.5 4.10E-03 2 2.59E-03 2.49E-03 8 1.12E-03 9.85E-04 24 8.32E-04 8.30E-04 96 5.91E-04 6.31E-04 720 0

1CAN122101 Page 9 of 9 3.0 RESULTS RADTRAD 3.03 was executed for the scenarios identified above.

PS LEAKAGE (39 GPD/SG)

Pathway File Name EAB LPZ CR Airborne Dose ps39.out 3.2868E+00 2.1977E+00 4.9540E+00 Cloud Dose c_ps39.out 6.9013E-03 Total 3.2868E+00 2.1977E+00 4.9609E+00 CONTAINMENT LEAKAGE Pathway File Name EAB LPZ CR Airborne Dose cmt.out 4.7183E+00 2.2730E+00 3.1183E+00 Cloud Dose c_cmt.out 5.7096E-03 Total 4.7183E+00 2.2730E+00 3.1240E+00

1CAN122101 Attachment 5 SGTR RADTRAD Input Files

1CAN122101 Page 1 of 2 SGTR RADTRAD INPUT FILES Nuclide Inventory Files Release File Name Compartment Primary rcs.nif Secondary secondary.nif Core (Accident- ai_spike.nif Induced Spike case)

Release Fraction Tables Release Type File Name Noble Gases ngas.rft Iodine iodine.rft Alkali Metals alkali.rft Dose Conversion Factor Tables Dose Type File Name Inhalation adult.inp Shine cloud.inp Plant Scenario Files RADTRAD Case Name Description Airborne Shine Secondary coolant iodine release from both 1 sec_iodine.psf c_sec_iodine.psf steam generators Secondary coolant alkali metal release from 2 sec_alkali.psf c_sec_alkali.psf both steam generators Primary coolant iodine release due to PS 3 rcs_iodine.psf c_rcs_iodine.psf and tube leakage

1CAN122101 Page 2 of 2 RADTRAD Case Name Description Airborne Shine Primary coolant alkali metal release due to 4 rcs_alkali.psf c_rcs_alkali.psf PS and tube leakage Primary coolant noble gas release due to 5 rcs_ngas.psf c_rcs_ngas.psf PS and tube leakage Primary coolant iodine release from 6 accident-induced spike due to P/S and tube ai_spike.psf c_ai_spike.psf leakage

1CAN122101 Attachment 6 MSLB RADTRAD Input Files

1CAN122101 Page 1 of 2 MSLB RADTRAD INPUT FILES Nuclide Inventory Files Release File Name Compartment Primary rcs.nif Secondary secondary.nif Core (Acciden- ai_spike.nif Induced Spike case)

Release Fraction Tables Release Type File Name Noble Gases ngas.rft Iodine iodine.rft Alkali Metals alkali.rft Dose Conversion Factor Tables Dose Type File Name Inhalation adult.inp Shine cloud.inp Plant Scenario Files RADTRAD Case Name Description Airborne Shine Secondary coolant iodine release from 1 sec_iodine_faulted.psf c_sec_iodine_faulted.psf faulted steam generator Secondary coolant iodine release from 2 sec_iodine_intact.psf c_sec_iodine_intact.psf intact steam generator Secondary coolant alkali metal release 3 sec_alkali_faulted.psf c_sec_alkali_faulted.psf from faulted steam generator Description RADTRAD Case Name

1CAN122101 Page 2 of 2 Airborne Shine Secondary coolant alkali metal release 4 sec_alkali_intact.psf c_sec_alkali_intact.psf from intact steam generator Primary coolant iodine release due to 5 rcs_iodine.psf c_rcs_iodine.psf P/S leakage Primary coolant alkali metal release 6 rcs_alkali.psf c_rcs_alkali.psf due to PS leakage Primary coolant noble gas release due 7 rcs_ngas.psf c_rcs_ngas.psf to PS leakage Primary coolant iodine release from 8 pe_spike.psf c_pe_spike.psf pre-existing spike due to PS leakage Primary coolant iodine release from 9 accident-induced spike due to PS ai_spike.psf c_ai_spike.psf leakage

1CAN122101 Attachment 7 CREA RADTRAD Input Files

1CAN122101 Page 1 of 1 CREA RADTRAD INPUT FILES Nuclide Inventory Files Release File Name Compartment Core Gap crea.nif Release Fraction Tables Release Type File Name Primary-to-secondary leakage and crea.rft containment Dose Conversion Factor Tables Dose Type File Name Inhalation adult.inp Shine cloud.inp Plant Scenario Files RADTRAD Case Name Description Airborne Shine 1 Primary-to-secondary leakage pathway ps39.psf c_ps39.psf 2 Containment leakage pathway cmt.psf c_cmt.psf

Retrieved from "https://kanterella.com/w/index.php?title=1CAN122101,_Response_to_Request_for_Supplemental_Information_Concerning_Licensing_Amendment_Request_to_Revise_Technical_Specification_3.4.12_and_3.4.13&oldid=3402763"