Text

Enclosure 6 - Millstone, Units 2 and 3 and ISFSI; North Anna, Units 1 and 2 and ISFSI; and Surry, Units 1 and 2 and ISFSI - Summary of Calculations
	ML19249B776
Person / Time
Site:	Millstone, Surry, North Anna, 07200002, 07200055
Issue date:	08/29/2019
From:	; Dominion Energy Nuclear Connecticut, Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	19-296
	Download: ML19249B776 (55)
	v • d • e

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 ENCLOSURE 6

SUMMARY

OF CALCULATIONS (UPDATED}

Dominion Energy Nuclear Connecticut, Inc. (DENC)

Virginia Electric and Power Company (Dominion Energy Virginia)

Millstone Power Station Units 1, 2 and 3 and ISFSI North Anna Power Station Units 1 and 2 and ISFSI Surry Power Station Units 1 and 2 and ISFSI

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 2 of 54

SUMMARY

OF CALCULATIONS (UPDATED)

Calculation Summaries (Updated)

General:

One of the goals of the EAL reanalysis effort was to apply consistent practices and methods between the Dominion fleet plants and to define common response criteria across the fleet for certain EAL initiating conditions (i.e. fuel clad degradation and fuel barrier failure criteria, RCS sample dose rates and sample line dose rates).

Uncertainties in the proposed methodologies are similar to anticipated and acceptable uncertainties known to exist in most radiological assessment methods and techniques in support of emergency response. The primary variables used in the reanalysis of dose rate EAL thresholds response are, (1) source term, (2) shielding geometry, and (3) source volume. New calculated EAL values are based on expected plant conditions.

For instance, core isotopic release fractions are based on realistic recommendations from NUREG 1228,"Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents", October 1988" rather than conservative and bounding design basis guidance of NUREG 1465,"Accident Source Terms for Light-Water Nuclear Power Plants", dated February 1995. Equilibrium coolant concentrations are taken from calculations of Technical Specification RCS Coolant Activity applicable to the fuel clad degradation initiating condition criteria. RCS volumes are updated to assume hot full-power conditions. Single dose rate response thresholds are used wherever applicable across all Fleet plants. Common fleet EAL values for similar category thresholds are deemed important to enhance familiarity between facilities/units to reduce the potential for human error. Calculation summaries have been provided in this Enclosure which contain additional information for critical calculations used in our EAL reanalysis.

The basis for the gaseous Unusual Event IC and associated thresholds has been revised to correspond to any unplanned release of gaseous effluent radioactivity to the environment that will result in release 2 times the allocated site-specific effluent release controlling document limits for 60 minutes or longer. This Unusual Event gaseous release criterion is being used consistently across all operating nuclear units at Dominion Energy nuclear stations at Millstone, North Anna and Surry. The word

'allocated' is required because for some release points, using ODCM methods and limits to determine the UE EALs, the UE values calculated were greater than ALERT EAL threshold values or did not provide a factor of 10 separation from the ALERT EAL threshold. To maintain consistency across the Fleet and reduce confusion and human error potential, a single Initiating Condition (IC) definition for gaseous and liquid releases at the NOLIE level is being used. The Initiating Condition (IC) will be worded:

"Release of gaseous or liquid radioactivity greater than 2 times the allocated ODCM

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 3 of 54 limits for 60 minutes or longer." This method provides a justifiable basis for NOUE thresholds based on established methods and setpoints provided in the facility ODCM.

The proposed NOUE values will classify events based on degradation in the level of safety of the plant and will maintain a near linear . escalation between all four classification levels (i.e., NOUE, ALERT, Site Area Emergency (SAE) and General Emergency (GE)). The IC being used is the same IC definition currently used for gaseous pathways in the North Anna and Surry NEI 99-01, Revision 4 EALs.

Classification thresholds within Table R-1 were generated using the MIDAS dose assessment code. Inputs to MIDAS use most prevalent meteorological data and expected release point parameters. Most prevalent meteorology represents conditions that would most likely exist (based on the most prevalent stability class and average wind speed within that stability class). Dispersion based on most prevalent meteorology differs from that assumed in the ODCM which uses annual average meteorology.

Dispersion based on actual meteorological conditions at the time of the emergency (most prevalent) can be 10 - 20 times higher than the annual average dispersion prescribed for use in an ODCM. Assumptions of one-hour decay since shutdown and a one-hour release duration are applied. Mitigating reduction mechanisms (e.g., decay, sprays, filters) input into MIDAS for each accident type determined the radiological release source term consistent with the guidance provided in NUREG-1228.

Dose rate values specified in Tables F-2, F-3 and F-4 were developed using a method to minimize error(+/-) for the threshold value within defined time (decay) periods. Time periods were chosen to fit monitor response (fast changes in response early following reactor shutdown are broken up into smaller time periods to better approximate expected change). Values were chosen within each time period to minimize error

(<50%) from the highest to lowest response within each time range. Fuel clad barrier loss thresholds are each calculated based on 5% fuel clad damage which represents a significant amount of fuel clad damage comparable to 300 uCi/gm DEl-131.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 4 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-18-08, "MPS1 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

To determine new Emergency Action Level release threshold using updated guidance from NEI 99-01 Rev 6 for Millstone Unit 1 continuous monitored pathway from the Spent Fuel Pool Island Vent.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. MP-22-REC-BAP01, Rev.29, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)."

3. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document", Oct 18, 2016.

4. Calculation RA-0016, Revision OAddendum A, "Radiological Consequences of Release of all Gap Activity in the Spent Fuel Pool at MP1", Nov. 29, 2010.

5. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

6. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614.

7. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

8. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment."

9. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters".

Computer Codes Used:

MIDAS software (Ref 5) was utilized to determine the projected EDE, TEDE and Thyroid CDE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for the ALERT classification. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref.6).

Methodology:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency.

The calculated threshold value considers a source of 100% Kr-85 release from damage of irradiated assemblies in the Unit 1 SFP and meteorology in accordance with NEI 99-

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 5 of 54

01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01.

To determine the ALERT radiological threshold for the MP1 SFPI vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters and normalized source terms of 1 Ci/sec. No mitigating reduction mechanisms (decay, sprays, filters, etc.) were used as input into MIDAS for this particular calculation as iodine and particulate removal mechanisms have no effect to the release of Kr-85. The MIDAS outputs generated represent a radiological prediction normalized to the source entered (e.g., 1 Ci/sec of Kr-85).

The maximum projected EDE, TEDE at or beyond the site boundary distance were obtained from the MIDAS outputs. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor which would yield the referenced dose criteria for the ALERT classification.

This concentration is the rad monitor action level for the ALERT classification. Thyroid COE limits were not included in this evaluation per new guidance in EPA-400 (Ref. 7) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

From the predicted release to obtain levels that achieve the ALERT EAL threshold limit of> 10 mrem TEDE, the corresponding number of irradiated fuel assemblies that would need to be damaged to achieve that source term will be determined.

==

Conclusions:==

The condition where the NOUE threshold is exceeded for 60 minutes is indicative of the inability to terminate radioactive release within prescribed regulatory and license limits and therefore represents a loss of plant control and degraded safety. For the Unusual Event (NOUE) threshold value based on station release limits, the methodology and assumptions established with the Millstone REMODCM were followed.

For the ALERT threshold value determined, the release conditions required to produce 10 mrem TEDE from a pure Kr-85 release using predominant meteorological conditions was calculated using the MIDAS accident dose software code. The ALERT threshold value determined, while equivalent to release conditions that would produce 10 mrem TEDE, requires that 50% of the spent fuel pool irradiated assemblies will need to fail and release within 15 minutes to the environment to produce radiological conditions that produce such dose.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 6 of 54 MP1 SFPI Vent RM-SFPl-02 EAL Thresholds MPl SFPI Vent EAL Escalation 1.0E+02 : .. -- - - - - - - - - - - - - - * - - * - * * - - -

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 7 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-18-02, "MPS2 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the MP2 Ventilation Vent and Millstone Stack based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400 for removal of thyroid COE PAG limits.

References:

1. Nu.clear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents",

1. McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. MP-22-REC-BAP01, Rev.27-01, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)."

6. MG-EV-99-0004, Rev 0, "Units 1, 2, 3 Radiological Boundaries", July 20, 1999.

7. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment."

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. RERM-02906-R2, Rev. 1, "Millstone Unit 2 Vent Radiation Monitor (RM- 81328)

High Range Setpoint", Jan 16, 2003.

10. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document", Oct 18,2016.

11. DWG 25203-20098, Rev. 05, "Main Steam Piping Plan - Containment & Aux.

Bldg", dated 10/10/2003.

12. Millstone Unit 2 Radiation Monitor Manual

13. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

14. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters" MIDAS Dose Software:

MIDAS software (Ref 3.2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration are used for the purpose of calculating emergency action levels for ALERT,

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 8 of 54 Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3).

MIDAS Assumptions and Inputs:

For all MIDAS runs, the stability class=D, ambient lemperature=50°, and the direction=252° (from) lo result in MIDAS shortest distance to the site boundary = 496 m {-0.31 mi} and 620 m {-0.39 mi) in the ENE direction. (5° added to centerline to align over calculation point within MIDAS tabular report at 0.37 miles) .

.. Discharge flow per release point was taken from Ref. 7.

M Release from a LOCA is considered leak of containment GAS.

Based on the information above, one run is required for each of the release points as all input assumptions are identical and can be ral.klJid,to the applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance with NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

The RU1 thresholds based on the REMODCM Instantaneous Release Rate Limits that utilize annual average meteorology are shown to be essentially equal to 1 mrem TEDE using most prevalent met conditions. This shows the same principles of dose and maintains consistency with the Technical Specifications. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the emergency condition. The Unusual Event (UE)

EALs are calculated for release points controlled in the REMODCM, Ref. 3.

To determine the EAL radiological thresholds for MP2 Ventilation Vent and Millstone Stack, and release points, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and stack

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 9 of 54 rad monitors. . An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc). For MP2 Vent and Millstone Stack releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action level thresholds for the MP2 Vent and Millstone Stack.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs .. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. Thyroid COE limits were not included in this evaluation per new guidance in EPA-400 (Ref.13) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

Cone Ius ions:

Following the guidance of NEI 99-01 Revision 6, recommended values for Millstone 2 release point EAL thresholds were calculated.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. To maintain Dominion Fleet commonality, this difference from the prescribed guidance in NEI 99-01 Revision 6 is needed by adding the word 'allocated'.

The difference is necessitated because some release pathways at the Surry and North Anna stations following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 10 of 54 Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Millstone 2. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the lowest EAL threshold.

Figura 3 - MP2 Vent RM81S8 and Stack RMB1S9 EAL Escalation MP2 Vent and Stack.EAl Escalation 1.0f *Cd f

L.

I

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 11 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-.

18-03, "MPS3 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the MP3 Ventilation Vent, Millstone Stack, and MP3 ESF Vent based on updated guidance from NEI 99-01; Rev 6 and revision to EPA-400 for removal of thyroid COE PAG limits.

References:

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

1. November 2012

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17 .022218

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents",

5. MP-22-REC-BAP01, Rev.27-01, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)"

6. MG-EV-99-0004, Rev 0, "Units 1, 2, 3 Radiological Boundaries", July 20, 1999

7. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment"

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. Blank

10. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document"

11. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017

12. Millstone Unit 3 Radiation Monitor Manual

13. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

14. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters" MIDAS Dose Software:

MIDAS software (Ref 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 12 of 54 MIDAS Assumptions and Inputs:

,Rf;LEA POINT<

U3 Vent U3 Vent Site.S~

SiteS~ LOCA Gap Site~,lk LOCA Gap

For all MIDAS runs, the stability class=D, ambient temperature=SO", and the direction=252" (fmm}fo result in MIDAS shortest distance to the site boundary =496 m (-0.31 mi} and620 m (-0.39 mi) in the ENE direction. (5" added to centerline to align over calculation point within MIDAS tabular report at 0.37 miles).

- Discharge flow per release point was taken from Ref. 7.

~R@Jg<,l,~~ from a LOCA is considered leak of containment GAS.

Based on the information above. one run is required for each of the release points as all input assumptions are identical and can be all.ilt@sj,to the applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance with NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

RU1 thresholds based on the ODCM Instantaneous Release Rate Limits that utilize annual average meteorology are compared against dose criteria to maintain a logical and consistent escalation between the UE and ALERT thresholds. Both are based on the same principles of dose and maintain consistency with the Technical Specifications.

Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOUE emergency condition.

To determine the EAL radiological thresholds for MP3 Ventilation Vent and Millstone Stack MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and stack rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 13 of 54 release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc). For MP3 Ventilation Vent and Millstone Stack releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the MP3 Ventilation Vent and Millstone Stack.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs .. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced TEDE dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. Thyroid COE limits were not included in this evaluation per new guidance in EPA-400 (Ref. 13) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

==

Conclusions:==

Following the guidance of NEI 99-01 Revision 6, recommended values for Millstone 3 release point EAL thresholds based on the results of this calculation are summarized in Table 1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. To maintain Dominion Fleet commonality, this difference from the prescribed guidance in NEI 99-01 Revision 6 is needed by adding the word 'allocated'.

The difference is necessitated because some release pathways at the Surry and North Anna stations following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety. "

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the rriost prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Millstone 3. This figure demonstrates that the four EALs are sufficiently separated and show escalation from

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 14 of 54 the NOUE level up through the GE level. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the lowest EAL threshold.

The Millstone 3 ESF Vent is isolated upon Safety Injection signal. Therefore, there is no ALERT, SAE , or GE threshold for this pathway since this pathway would be isolated prior to reaching levels sufficient to warrant higher classification.

Figure 3 - MP3 Vent(RE10} and Stack {RE19} EAL Escalation MP3 Vent and Stack EAL Escalation Uf.03 ~ - - - - - - - - - - - - - - - - - - - - -

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 15 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from North Anna Calculation RP-08-22, "North Anna Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99".".

01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the NAPS Ventilation Vent and Process Vent based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614 and all previous files.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. VPAP-2103N, Revision 28, "Off-site Dose Calculation Manual (North Anna)."

6. NA-ENGT-000-CME 97-0010, Rev. 0, "Evaluation of the Required Tech Spec Flow Rate Value for Process Vent Blowers ... " Feb. 10, 1997. *

7. EPIP-4.03, Revision 22, "North Anna Power Station Dose Assessment Team Controlling Procedure".

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. Calculation PA-0225, Revision 0, Addendum (00, DOA, OOB, OOC), "North Anna Radiation Monitor Conversion Factors and EAL Readings".

10. HP-3010.040, Revision 27, "North Anna Power Station - Radiation Monitoring System Setpoint Determination".

11. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

12. RP-AA-151, Revision 0, "Radiation Protection Technical Bases Analyses or Calculations".

13. NAPS UFSAR, Revision 53.03.

14. NE-GL-0035N, "PC-MIDAS Guideline", Revision 11.

15. 1-E-O, Rev. 50, North Anna Emergency Procedure - "Reactor Trip or Safety Injection".

16.DC NA-11-01082, Rev. 01, "Main Steam Radiation Monitor Replacement".

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 16 of 54 Computer Codes Used:

MIDAS Dose Software MIDAS software (Ref. 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref.3).

MIDAS Assumptions and Inputs:

- - RELEASE POINT EAL,

Mer Mer WS-LL .WS4.JL Rate>*

Release Flow . *>* Decay A~cident **. *

~o.~r~ Spray\, Filter *** Qurali(!fl

.... (mph} {mph) .... ..... {cfm)> <110:

. ** i : *** ... /. , . 1* I *** .* cc**

.* *. < 1 **

- - - {hr) <

Vent.A. .ALERT 4.9 NIA 1,yg/cc 40,000* 1 LOCA-Gasleak Gap y y 1 VentA SAE 4.9 N/A 1 J,tCj/cc 40,000 1 LOCA- Gas Leak Gap y y 1 VentA GE 4.9 NIA 1 J,tCJ/cc 40,000 1 LOCA-Gas Leak Gap y y 1 VentB ALERT 4.9 NIA 1 UCj/cc 12,000* 1 LOCA- Gas Leak Gap y y 1 VentB SAE 4.9 NIA 1 IJ.Cj/cc 12,000 1 LOCA- Gas leak Gap y y 1 VentB GE 4.9 NIA 1 uCj/cc 12,000 1 LOCA- Gas Leak Gap y y 1 ProcYnt ALERT 4.9 8.1 1 .uci/cc 300 1 LOCA- RCS Leak... Gap y y 1 ProcYffi SAE 4.9 8.1 1 UCj/cc 300 1 LOCA- RCS Leak Gap y y 1 ProcVnt GE 4.9 8.1 1 "' itcc 300 1 LOCA- RCS Leak Gao y y 1

For all MIDAS runs, the stability* class=D, ambienttemperature=SO*, and the d1rection=252' {from}to resultm - MIDAS shortest distance to the site boundar;=5000 {t(-0.94 mi). (5° added to align calculation pointwithin MIDAS tabularreportat 1 mile}.

- AccidenlflowsfromRef.15,AttachmentA

- - Two plant accident conditions apply to effluents discharged from the Process Vent, (1) LOCA/GAP/Spray/Filler/RCS Leak-that represents activity in the RCS that is leaking iJJ}pjl)~A.l..!15,l;M.g..gJ (2U,QG,6lG,~lt;R5J)J~('f.ftt~flG.<1sJ..,tk1Kfrnm-CNMI-that represents discharge from CNMT during early onset of a LOCA before isolation. In MIDAS, the isotopic mix for both of these scenarios is the same with a total credited OF of :woo.

Based on the inforrnation above, one run is requiredfor each oflhe release points as all input assumptions are .identical and can be r.atlQ.!W.to the applicable E.A.Lthreshold.

Methodology:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance to NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

RU1 thresholds, based on the ODCM Instantaneous Release Rate Limits that utilize annual average meteorology, are compared against dose criteria to maintain a logical and consistent escalation between the UE and ALERT thresholds. Both are based on the same principles of dose and maintain consistency with the Technical Specifications.

Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOUE emergency condition. The Unusual Event (UE) EALs are calculated for release points controlled in the ODCM, Ref. 5.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 17 of 54 To determine the EAL radiological thresholds for Ventilation Vent and Process Vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and process vent rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc).

For Ventilation Vent and Process Vent releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the Ventilation Vent and Process Vent.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs. These doses were divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. The lowest predicted concentration between each dose analyzed is selected as the applicable EAL limit.

==

Conclusions:==

Following the guidance of NEI 99-01 Revision 6, recommended values for North Anna release point EAL thresholds based on the results of this calculation are summarized in Table R-1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological

Serial No. : 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 18 of 54 conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 6 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from North Anna. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Additionally, it can be seen that the NOUE threshold is set at or above the setpoint limit, thus assuring the operator will be alerted due to radiation monitors going into alarm prior to or when the NOUE level is exceeded.

Frgure 6- NAPS Vent and Process Vent EAL Escatation NAPS Vents and Process Vent EAL Escalation H )£"'09 =--------------------

BE+OS Z.EiMS

u :f+CS l

g l .CE..ai +/--** ...................................................................................................... +* * / J t'

j ii i

"~

v;;.~M-!n j l.O.E;,Q; ,. ~.gi.1-uo J

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 19 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Surry Calculation RP 01, "Surry Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist techni.c-al reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the SPS Ventilation Vent and Process Vent based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1 .5.17 .022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20161219 and all previous files.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. VPAP-2103S, Revision 20, "Off-site Dose Calculation Manual (Surry)."

6. EPIP-4.03, Revision 20, "Surry Power Station - Dose Assessment Team Controlling Procedure".

7. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

8. Calculation PA-0224, Revision 0, Addendum (00, OOA, OOB, OOC,D, and E), Surry Power Station Radiation Monitor Emergency Action Levels (EALs) for the Process Vent and Ventilation Vent #2, Steam Line, and Auxiliary Feed water Exhaust".

9. HP-3010.040, Revision 36, "Surry Power Station - Radiation Monitoring System Setpoint Determination".

10. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

11. RP-AA-151, Revision 0, "Radiation Protection Technical Bases Analyses or Calculations".

12.DC SU-10-01083, Rev. 01, "NRC Radiation Montiors Replacement Project".

13. NE-GL-0035S, "PC-MIDAS Guideline", Revision 11.

MIDAS Dose Software:

MIDAS software (Ref 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 20 of 54 duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications.

MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3f MIDAS Assumptions and Inputs:

Vent Y.ent ALERT 4.7 NIA 1 u.Cilcc 34 000 1 LOCA - Gas leak Gap Y Y 1 Vent Vent SAE 4.7 NIA 1 _uCilcc 34 000 1 LOCA - Gas Leak Gap Y Y 1 Ventjl~.t GE 4.7 NIA 1 llCI/cc 34,000 1 LOCA- Gas leak Gap Y Y 1 ProcVnt ALERT 4.7 9.0 1.uGi/cc 300 LOCA-RCSleak*" Gap Y Y Proc )mt SAE 4.7 9.0 1 u_Cj/cc 300 LOCA- RCS Leak Gap Y Y Proc ,\Int GE 4.7 9.0 1 _u_~i/cc 300 LOCA - RCS leak Gao Y Y

~ For all MIDAS runs, the stability dass=E, ambient temperature=50", and the direction=252° (from) to result in MIDAS shortest distance to the site boundary= 503 m (-0.31 mi). (5. added to centerline to align over calculation point within MIDAS tabular report at 0.37 miles).

""The Process Vent system takes input from the Aux. ID- atmosphere, Aerated Waste system, Containment Vacuum Pump discharge, Waste Gas Decay tanks, and from the Gas Stripper Surge drum. Two likely plant accident conditions could apply to effluents discharged from the Process Vent, (1) LOCAiGAP/Spray/Filter/RCS Leak - that represents activity in the RCS that is leaking in1oJhe.AwLfllruL.9LL2lJ*.OCALGAe/NQ*.Spr.a.YlEi!ierLG.a~LL.eaK.frJim.CNMI - that represents discharge from CNMT during early onset of a LOCA before isolation. In MIDAS, the isotopic mix for both of these scenarios is the same with a total credited DF of 3000.

Based on the information above, one run is required for each of the release points as all input assumptions are identical and can be rn.tlo. to the applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance to NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1. The RU1 thresholds are based on the REMODCM Instantaneous Release Rate Limits that utilize annual average meteorology. They are based on the same principles of dose and maintain consistency with the Technical Specifications. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOUE emergency condition.

The Unusual Event (UE) EALs are calculated for release points controlled in the ODCM, Ref. 5.

To determine the ALERT, SAE and GE EAL radiological thresholds for Ventilation Vent and Process Vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 21 of 54 expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and process vent rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc).

For Ventilation Vent and Process Vent releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the Ventilation Vent and Process Vent.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs. These doses were divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the ALERT, SAE and GE emergency classifications. The lowest predicted concentration between each dose analyzed is selected as the applicable EAL limit.

==

Conclusion:==

Following the guidance of NEI 99-01 Revision 6, recommended values for Surry release point EAL thresholds based on the results of this calculation are summarized in Table 1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 22 of 54 within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Surry. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Additionally, it can be seen that the NOUE threshold is set at or above the setpoint limit, thus assuring the operator will be alerted due to radiation monitors going into alarm prior to or when the NOUE level is exceeded.

Serial No. : 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 23 of 54 Figure 3 - SPS Vent and Process Vent. EAL Escalation SPS ve,nt and Process Vent EAL Escalation H1E~ - - , , - - - - - - - - - - - - - - - - - - - - - - - - - -*

1 .2E+07

1.1£+06 a 2.SE+OS

tOE~

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 24 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Surry Calculation RA-0063, Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor response to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

1. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

2. Drawing 11448-FM-1 E, Rev. 13, Sheet 1, "Mach. Loe. - Reactor Cont. Sections "A-A", "E-E" & "Z-Z" Surry Power Station - Unit 1.

3. Drawing 11448-FE-46C, Rev. 16, "Conduit Plan Reactor Containment El. 47' -4" Surry Power Station - Unit 1."

4. PA-0163, Rev, 0, Add. D, "Calculation of the Surry AST LOCA Dose Consequences to Support the Gothic Containment Reanalysis for GSl-191."

5. RA-0008, Rev. 0 thru Add. B, Core Isotopic Inventories for Surry Dose Consequence Analyses Based on the Alternate Source Term, May, 2010.

6. Drawing 11448-FE-46D, Rev. 12, "Conduit Plan Reactor Containment El. 47' - 4" Surry Power Station - Unit 1."

7. SEALTB Rev. 4, "Emergency Action Level Technical Bases Document", December 2013.

8. Drawing 11448-FP-13D, Rev. 14, Sheet 4, "Containment & Recirc Spray System Sh 4."

9. Radiological Health Handbook, January 1970.

10. Drawing 11448-FM-1 G, Rev. 14, Sheet 1, "Mach. Loe. - Reactor Cont. Sections "C-C" & "D-D" Surry Power

11. Station - Unit 1."

12. SEAL MATRICES Rev. 4, "Surry Power Station Emergency Action Level Matrix".

Computer Codes Used:

Microshield version 7.02

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 25 of 54 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

Microshield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and/or

Conclusions:

Dose Rates vs Decay Time for the 5% Clad Damage Decay time (hrs) Dose Rate (R/hr) 0 4.27E+02 1 1.24E+02 2 8.97E+01 4 6.05E+01 8 3.52E+01 16 2.12E+01 24 1.76E+01 36 1.53E+01 48 1.40E+01 72 1.20E+01 Figure 1:Dose Rate,,s Dec,ayTimerorlhe5%Clad Damage Dose Rate {R/hr)vs DetayTimf! (hrs)

t ______________________________________ - --- !

~:lel* ............................................................................................................................................................................................................................... ~

~-- .............................................................................................................................................................................................................................. :i

~:::::t>------***_******_-**._*

_*

_*

_*

_*

_*

_*

_*

_*.*****_******_****-----*---1. _

l.~ ;

s~+.1--------------------

l.Q:i:S.~

~-rof~l +-j Ii

--=,;;::-:----

!o',.0-!(;,:(} *:rm-H-mrm**m-r--HHmm-m**-H***r----*-

Z ~ ~ ~

0 - -.. -*-*--r-*----........,

M ,~ - 8~

i

~ - - **-***---*----*--"'""-3~r-Sh~--- ---*--*--_J

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 26 of 54 e Rates vs Decay Time for the 20% Clad Failure Decay time (hrs) Dose Rate (R/hr) 0 1.71E+03 1 4.97E+02 2 3.59E+02 4 2.42E+02 8 1.41E+02 16 8.45E+01 24 7.02E+01 36 6.11E+01 48 5.57E+01 72 4.79E+01 Figure 2: Dose Rare vs D=ylime for tire 20% Clad failure Dose Rate (R/hr)vs Decay Time (hrs)

!iltaE~::::1---------------------

- um.¢! ~ - - - - * - * - - - - - - - - - - - - - - - - * - - - - - * - - * * * * * - - - * * * - * * * - * - * * - * * * - * * * * * * * - - - -

tl!li*tl ............................................................................................................................................................................................................................................

~.00:E*~"i ...*************.******************.******************.*.******************************

Dc,aBat;

-= C.'llS~

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.HXlh¢.1f--~

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I

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 27 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from North Anna Calculation RA-0064, Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor response to a large break LOCA with quench (containment) and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the North Anna EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

2. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

3. Drawing 12050-FM-1A, Rev. 19, Mach. Loe. - Reactor Cont. Sh. 1 Plan EL 291'-10" North Anna Power Station - Unit 2."

4. NEAL MATRICES Rev. 7, "North Anna Power Station Emergency Action Level Matrix".

5. PA-0186, Rev. 0, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry," March 4, 2002.

6. PA-0186, Rev. 0, Add. A, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry," Sept. 7, 2006.

7. Calculation 11715-ES-017, Rev. 0, "North Anna 1 & 2 Containment Free Volume,"

Aug.31,1971.

8. Drawing 12050-FM-1 C, Rev. 18, "Mach. Loe. - Reactor Cont. Sh. 3, Plan - El. 241' -

O" North Anna Power Station - Unit 2." *

9. NEALTBD Rev. 7, "Emergency Action Level Technical Bases Document", March 2015.

10. Radiological Health Handbook, January 1970.

11. NA-W0-000-00426795-01, Work Order Task for Reactor Containment Elevation 291 Area High Rad Monitor, March 18, 2000.

12. Drawing 11715-FM-1 G, Rev. 18, Sheet 1, "Mach. Loe. - Reactor Cont. SH7 Sections 3-3 & 4-4 North Anna Power Station."

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 28 of 54 Computer Codes Used:

Microshield version 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

Microshield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and/or

Conclusions:

Dose Rates vs Decay Time for the 5% Clad Damage Decay time (hrs) Dose Rate (R/hr)

O 5.81E+02 1 1.69E+02 2 1.22E+02 4 8.29E+01 8 4.91E+01 16 3.04E+01 24 2.56E+01 36 2.26E+01 48 2.07E+01 72 1.78E+01 Figure 1: Dose Rate vs Decay lime for tne 5% Clad Damage Dose Rate {R/hr) vs Decay Time (hrs) 7.00E-+02 -,-;- - - - - - - - - - - - - - - - -

6.00E+o2 -;._!_- - - - - - - - - - - - - - - - - -

5.00E-+02 ->------------*--------

Dose Rate 4.00E+o2 ~ - - - - - - - - - - - - - - - - -

(R/hr) 3.00E+o2 ..;i-------------------

2.00E-+02 *H*-*-****-***********************-********--*****-****************--**************

l.OOE+o2 O.OOE-+00 t.~::::::::::::~~~~~~.=.:::::.:::::.~.~=--~,~--=~.------,

0 10 20 30 40 50 60 70 80 Hours After Shutdown

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 29 of 54 Dose Rates vs Decay Time for the 20% Clad Failure Decay time (hrs) Dose Rate (R/hr)

O 2.32E+03 1 6.75E+02 2 4.88E+02 4 3.31E+02 8 1.96E+02 16 1.21E+02 24 1.03E+02 36 9.01E+01 48 8.29E+01 72 7.16E+01 Figure 2: Dose Rate vs Decay Time for the 20% Clad Faffure Dose Rate (R/hr) vs Decay Time (hrs) 2.SOE+03 2.00E+o3 Dose Rate 1.SOE+o3 (R/hr) 1.00E+o3 5.00E+o2 O.OOE+oO 0 10 20 30 40 50 60 70 80 Hours After Shutdown

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 30 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Calculation RA-0074, Millstone Unit 2 Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is p*rovided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor system response for Millstone Unit 2 to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the Millstone Unit 2 EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. Vendor Calculation 3D00-005, Rev. 3, "Millstone Unit 2 Containment Heat Sinks,"

December 2006.

2. Nuclear Energy Institute Document NEI 99-01, Rev. 6, "Development of Emergency Actions Levels for Non Passive Reactors," November 2012.

3. Reference Manual MP-26-EPA-REF02, Rev. 24, "Millstone Unit 2 Emergency Action Level (EAL) Technical Basis Document," March 2016.

4. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

5. Drawing 25203-28014, Rev. 15, "Millstone Unit 2 Instrument Location Containment Plan El. 14'-6" & 38'-6".

6. Drawing 25203-27021, Rev. 3, "Millstone Nuclear Power Station Unit No. 2 General Arrgt- Containment & Aux. Bldg. Section A-A".

7. 7. Engineering Calculation M2AST-03105R2, Rev. 0, "Millstone 2 Alternate Source Term," January 2002.

8. 8. Radiological Health Handbook, January 1970.

9. Computer Code MicroShield, Version 7.02, Grove Software, Inc.

10. Millstone Unit 2 EALs MP-26-EPI-FAP06-002, Rev. 7.

11. Drawing 25203-20104, Rev. 3, "Millstone Nuclear Power Station Unit No. 2 Area 5 Piping Containment Spray % H2 Purge".

12. Drawing 25203-27022, Rev. 9, "General Arrangement Containment & Aux. Bldg.

Section B-B".

13. Engineering Calculation NUC-181, Rev. 1, "MP-2 Design-Basis Loss of Coolant Accident- Radiation Source Terms," June 1998.

14. Vendor Technical Manual VTM-303-007A, "Energy Response Test & Dose Rate Calibration of Model RD-23 Det." June 1986.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 31 of 54 Computer Codes Used:

MicroShield 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

The region of containment measured by each of the radiation monitors is modeled using a simplified rectangular box configuration. This geometry is used in MicroShield Version 7.02 [Reference 9] with the receptor location being the radiation monitor location.

MicroShield is used in this analysis to calculate the expected response from the Containment High Range Radiation Monitors.

Results and Conclusions The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 5% clad damage.

5% Clad Damage Table Decay Time(hrs) Dose Rate (R/hr) 0 266 1 78 2 55 4 34 8 16 16 6.0 24 3.7 36 2.6 48 2.1 72 1.7

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 32 of 54

-Figure 13.2 5% Fuel Clad Damage Dose Rate (R/hr) vs. Decay Time (hrs) 100 10 40 so 70 BO HoursAftcrSttutdawn The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 5% clad damage. These results were generated by multiplying the 5% fuel clad damage by a factor of 4, since the concentrations of dispersed nuclides are 4 times greater between the 20% and 5% calculations.

20% Clad Damage Table Decay Time(hrs) Dose Rat (R/hr) 0 1065 1 314 2 223 4 138 8 64 16 23 24 14 36 10 48 8.5 72 6.9

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16 /4 7/55/56 Enclosure 6 Page 33 of 54 Agurel3.4 20% Fuel Clad Damage Dose Rate (R/hr) vs. Decay lime (hrs) ltlOO 100 Dose Rate (R/hrt lJ) 10 20 30 40 SO 70 80 Hours After Shutdawn

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 34 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Calculation RA-0075, Millstone Unit 3 Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor system response for Millstone Unit 3 to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the Millstone Unit 3 EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. Vendor Calculation ES-227, Rev. 0, "Containment Structure Free Volume,"

November 1979.

2. Millstone Unit 3 EALs MP-26-EPI-FAP06-003, Rev. 11.

3. Computer Code MicroShield, Version 7.02, Grove Software, Inc.

4. Engineering Calculation RERM-04345R3, Rev. 0, "Millstone Unit 3 Containment High Range Radiation Monitors' Accident Responses to a LOCA," April 2008.

5. Nuclear Energy Institute Document NEI 99-01, Rev. 6, "Development of Emergency Actions Levels for Non-Passive Reactors," November 2012.

6. Reference Manual MP-26-EPA-REF03, Rev. 21, "Millstone Unit 3 Emergency Action Level (EAL) Technical Basis Document," May 2016.

7. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

8. Drawing 25212-27012, Rev. 16, "Millstone Nuclear Power Station Unit No. 3 Machine Location - Containment Structure - Plan El 51 '-4"."

9. Engineering Calculation NUC-181, Rev. 1, 'MP-2 Design-Basis Loss of Coolant Accident - Radiation Source Terms," June 1998.

10. Radiological Health Handbook, January 1970.

11. Drawing 25212-27015, Rev. 12, "Millstone Nuclear Power Station Unit No. 3 Machine Location - Containment Structure - Section 3-3."

12. Engineering Calculation M3AST-01942R3, Rev. 1, "Millstone 3 Alternate Source Term," May 2006.

13. Vendor Technical Manual VTM-303-007A, "Energy Response Test & Dose Rate Calibration of Model RD-23 Det." June 1986.

Computer Codes Used:

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 35 of 54 MicroShield 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

The source volume bounded by the annular crane wall is modeled using two right cylinder volumes, with a rectangular volume representing the pressurizer cubicle black body. This geometry is used in MicroShield Version 7.02 [Reference 3] with the receptor locations being the radiation monitor locations. MicroShield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and

Conclusions:

The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 5% clad damage.

5% Clad Damage Table Decay Time Dose Rate (R/hr) Dose Rate (R/hr)

(hrs) (3RMS*RE05A) (3RMS*RE04A) 0 703 550 1 191 149 2 132 103 4 83 65 8 40 31 16 16 13 24 8.5 6.8 48 7.3 5.8 72 6.0 4.8

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 36 of 54 Figure 13.2 5% Fuel Clad Damage Dose Rate (R/hr) vs. Decay Time (hrs)

~ :JRMS .. REOSA

-&-:3RMS'"RE04A The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 20% clad damage. These results were generated by multiplying the 5% fuel clad damage by a factor of 4, since the concentrations of dispersed nuclides are 4 times greater between the 20% and 5% calculations.

20% Clad Damage Table Decay Time Dose Rate (R/hr) Dose Rate (R/hr)

(hrs) (3RMS*RE05A) (3RMS*RE04A) 0 2814 2202 1 764 599 2 530 415 4 332 260 8 160 126 16 66 53 24 44 35 36 34 27 48 29 23 72 24 19

Serial No. : 19-296 Docket Nos. : 50-336/423/338/339/280/281 72-2/16/4 7 /55/56 Enclosure 6 Page 37 of 54 flr.ur* 13.4 20% Fuel Clad Damage Dose Rate (R/hr) vs. Dec:aynme (hrs)

- V *o. '""

(11/h<)

100

--=i-~3 AMS*iu:os.+\

..,C-JRMS,-~£"04:A.

l.0

. . .~

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 38 of 54 Calculation Summary Detector Response to an RCS Sample The following pertinent information has been extracted from Fleet Calculation RA-0059, Detector Response to an RCS Sample for EAL Classification of Fuel Clad Degradation and Barrier Loss. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to determine the detector response to various depressurized RCS samples measured with a gamma detectorat a distance of 1 ft. This calculation supports the Emergency Action Levels (EALs) for NAPS, SPS, MPS2 and MPS3. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. SEAL MATRICES Rev. 4, "Surry Power Station Emergency Action Level Matrix".

2. NEAL MATRICES Rev. 5, "North Anna Power Station Emergency Action Level Matrix".

3. MP-26-EPI-FAP06-002 Rev. 9, "Millstone Unit 2 Emergency Action Levels".

4. MP-26-EPI-FAP06-003 Rev. 8, "Millstone Unit 3 Emergency Action Levels".

5. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents"

6. Federal Guidance Report 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water and Soil"

7. 1304952001-UR-0001 Rev. 0 Add. A, "Primary Coolant Design/ Technical Specification Activity Concentrations"

8. RA-0008 Rev. 0 Add. 0, "Core Isotopic Inventories for Surry Dose Consequence Analyses Based on the Alternate Source Term", May 2010.

9. PA-0089 Rev. 0 through Add. B, "Surry Steam Generator Tube Rupture [SGTR]

Dose Calculations at the EAB, the LPZ and in Control Room", August 2000.

10. Robert C. Weast, ed., "CRC Handbook of Chemistry and Physics 60th edition", pg.

F-324, CRC Press, Inc.

11. PA-0194 Rev. 0 Add. 0, "Radiological Consequences of a Steam Generator Tube Rupture at North Anna Based on the Alternate Source Term", April 2003.

12. PA-0186 Rev. 0 Add. 0, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry"

13. PA-0081 Rev. 0 Add. 0, "North Anna SGTR Doses at the EAB, LPZ, and in Control Room", February 1991.

14. 06-ENG-04217R3 Rev. 0 CCN 1, "MP3 SPU Primary Coolant Design and Technical Specification Activity Concentrations etc."

15.M3AST-01942R3 Rev.1 CCN 1, "Millstone 3Alternate Source Term", May 2006.

16. M3ASTSGTR-04072R3 Rev. 1, "MP3 Uprated AST Steam Generator Tube Rupture Dose Consequences Analysis", April 2007.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 39 of 54

17. M2AST-04080R2 Rev. 0 Add. 1, "MP2 Coolant Activity for Accident Analyses",

March 2005.

18.M2AST-03105R2 Rev 0, "Millstone 2 Alternate Source Term", December 2002.

19. SEALTB Rev. 4, "Emergency Action Level Technical Bases Document", December 2013.

20. NEALTBD Rev. 5, "Emergency Action Level Technical Bases Document", December 2013.

21. MP-26-EPA-REF02 Rev. 022, "Millstone Unit 2 Emergency Action Level (EAL)

Technical Basis Document"

22. MP-26-EPA-REF03 Rev. 018, "Millstone Unit 3 Emergency Action Level (EAL)

Technical Basis Document"

23. NEI 99-01 Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors", November 2012.

Computer Codes Used:

MICROSHIELD Version 7.02 WATPROP Version 4 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Tech Spec RCS coolant activities and limits

3. Response from iodine concentration in degassed RCS sample

4. Release fraction from the fuel gap: 2% iodine (Ref. 5)

5. Detector placed 12 inches from outer edge of sample volume

6. Relative insensitivity of detector response demonstrated in the calculation to minor differences in distance to the sample volume

7. Relative insensitivity demonstrated in the calculation to sample container geometry

8. Relative proportional dose rate response demonstrated in the calculation to sample volume Method of Analysis:

The Technical Specification DE 1-131 spike iodine concentrations and core inventory iodine activity along with the RCS mass will be used to determine the sources for the various sample volumes. This source is then modeled in the code Microshield 7.02 with a dose point one foot from the source.

Results and

Conclusions:

In summary, dose rate responses have been determined for SPS, NAPS, MPS3 and MPS2 for TS coolant activity spikes and 5% cladding failure, which are used to develop EALs for Fuel Clad Degradation and Fuel Clad Barrier Loss, respectively, in various sample sizes and at several times after shutdown. A summary of conservative values representative of the expected detector response for Fuel Clad Degradation and Fuel Clad Barrier Loss in terms of mR/hr/ml vs. decay post-shutdown are presented Tables 1 and 2 below.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 40 of 54 Table 1: Summary of Unpressurized RCS Sample Dose Rates taken at 1 foot for Fuel Clad Degradation vs. Decay Post-Shutdown Station/Unit 1 hr 2 hr 4 hr 8 hr 12 hr 24 hr (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml)

SPS 0.15 0.13 0.10 0.07 0.06 0.04 NAPS 0.76 0.66 0.54 0.40 0.33 0.21 MPS2 0.76 0.66 0.54 0.40 0.33 0.21 MPS3 0.76 0.66 0.54 0.40 0.33 0.21 Table 2: Summary of Unpressurized RCS Sample Dose Rates taken at 1 foot for Fuel Barrier Loss vs. Decay Post-Shutdown Station/LI nit 1 hr 2 hr 4 hr 8 hr 12 hr 24 hr (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml)

SPS 17 12.7 8.5 5.4 4.0 2.2 NAPS 17 12.7 8.5 5.4 4.0 2.2 MPS2 17 12.7 8.5 5.4 4.0 2.2 MPS3 17 12.7 8.5 5.4 4.0 2.2 The scale of the response factors in the tables above are normalized to 1 ml. Samples when obtained in the plant will likely consist of greater volume (e.g., 120 or 250 ml).

Plant sampling procedures will direct to take a reading from 1 foot and divide by the actual collection volume and report the reading as (mR/hr per ml).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 41 of 54 Calculation Summary Post-Accident Radiation Response for Primary Sample Line The following pertinent information has been extracted from Fleet Calculation RA-0079, Post- Accident Radiation Response Curves for Primary Hot Leg Sample Lines. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrie*r matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to document the generation of radiation response curves for primary hot leg sample lines assuming an intact RCS. The accident scenario analyzed modeled 5% failed fuel (gap release).

References:

1. ORNL/TM-2005/39, Version 6.2.2, "SCALE Code System".

2. PA-0219, Rev. 0, "Post-Accident Radiation Response Curves for North Anna Primary Hot Leg Sample Lines."

3. ET-NAF-05-0013, Rev 0, "Post-Accident Radiation Measurement and Accident Classification Based on PA-0219."

4. Memorandum MP-CHEM-13-01 dated March 19, 2013, "Bases for Proposed M2 and M3 EALs."

5. MGP Instruments Document# 15-00031, Revision 5, "Area Monitor Probes AMP 50-100-200 Operations and Maintenance Manual."

Computer Codes Used:

SCALE 6.2 (Reference 1)

Microshield 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Response from gap release of noble gas, halogens, and cesiums

3. Release fraction from the fuel gap: 3% (Ref. 2)

4. Detector placed 2 inches from outside of various diameter sample line tubing

5. Sample line tubing length of 240 cm Methodology:

Source Term The initial source term for 5% failed fuel (gap release) was determined by multiplying the 1% failed fuel source term by a factor of 5. The results were then scaled down by the ratio of the water volume in the tubing to the RCS liquid inventory. The resulting source term in the tubing was decayed and converted to a photon spectrum using the ORIGEN code from the SCALE 6.2.2 code package.

Shielding Model - Scenario 1:

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 42 of 54 The sample line tubing was modeled as straight length of tubing running vertically up a concrete wall. The shielding model included the wall, tubing, and water within the tubing. The SCALE module MAVRIC was used to perform the photon transport. This is an improvement over the QADS model in that scattering/reflection from the concrete surfaces will be included and a dose rate measurement volume can be used in lieu of a point location. A tally volume comprised of air will be used to calculate the dose rate based on SCALE-supplied ANSl-77 flux-to-dose conversion factors.

~

I Detector Shielding Model - Scenario 2:

The sample line tubing was modeled as straight length of tubing. The shielding model included the tubing and water within the tubing. Microshield was used to determine the dose rate at a point 2 inches from the outside of the tubing. This scenario is designed to support a broad spectrum of possible sample line locations and geometries.

Results &

Conclusions:

Dose rates from a primary hot leg sample line assuming an intact RCS and 5% failed fuel were calculated. Tabular results and an associated radiation response curve are provided below. An adjustment factor to correct for a range of tubing sizes is provided.

These results are valid for North Anna, Surry and Millstone Power Stations.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 43 of 54

...............................................................................................-****-********-*-***--****-*****--*-*****-*-----*******"**************"**..************..** .... **********************************************-**************-*1 w I 18 *--*---- -*---*--*---r----*-***-**---*-*- *------***--,*---*****************************r******-***-****-***************- i 16 I l 14 --4.---1----+------l----;-----!-----I

... 12 +-----+------..1.-****-**-*--*-* ,

1 111 10 + - - - - - ' - . . + - - - + - - - - - - - - - - - ; i_ __

~ 8 ,

6 +----+--~&::-----+----i-----+-----1 l ____ _ I 4 - -*---***--**--** ***---*---*-+**-**-----*-*-*,~~F:::::***-**+:*****-==**::1 I i _ _ _.___

i i !

2 - - - - + - - - + - - - - - - - - - - . . . . ; !- - - - - 1 j 0 ** *--,---.,---*r-**- -*-,---,--,-1-...,--,--,-- *--.--.-*-.*--l.*--,--..,.-....,..-...1. . . . -r-....-r*-*-,-*-

Ij 0 4 8 12 16 20 24 Decay (hrs)

In some instances the tubing diameter may be larger than 3/8" tubing. In those instances the results can be scaled up by the ratio of tubing cross sectional area. The following scaling factors should be used:

Tubing size Li~uid Area Scaling (Nominal OD) (in ) Factor 3/8" 0.110 1.0 1/2" 0.196 1.8 3/4" 0.442 4.0 Note that the tubing size scaling factors are first-order approximations. Tubing wall thickness can vary somewhat, and the changing solid-angle between source and detector will also vary. However, the table is considered a reasonable indicator of the dose rate changing with tubing size.

Shielding - Scenario 2 The source term determined was used as a source term in MicroShield to determine the dose rate at a location 2" from the outside edge of a sample line. The sample line was modelled as a cylinder of water. The MicroShield output shows following results.

Decay Step Dose Rate (hrs) (R/hr) 1 4.5 2 3.45 4 2.35 8 1.5 16 0.95 24 0.7

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 44 of 54

~~~** ...................................................................................................................................................................................................................................................]

~

2.000

\

\ "- I.,

I 1.000 0.000 0 4

' 8

~ -*-

12 16 20 24 Decay (hrs)

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 45 of 54 Calculation Summary Post-Accident Radiation Monitor Response for Core Uncovery The following pertinent information has been extracted from Calculation RA-0078, Verification of Rad Monitor Response to Core Uncovery. It is provided to assist technical reviewers that will be evaluating this license amendment request for MPS3, NAPS & SPS. (Note: MPS2 is analyzed in M2EP-04164R2).

Purpose:

The purpose of this calculation is to determine a single, common value for radiation monitor responses to core uncover for Millstone Unit 3, North Anna Units 1 & 2, and Surry Units 1 & 2. This analysis is only applicable when the reactor head is off of the vessel.

References:

1. ORNL/TM-2005/39, Version 6.2.2, "SCALE Code System".

2. LA-CP-14-00745, Rev. 0, "MCNP6 User's Manual, Code Version 6.1.1 beta."

3. Calculation M3EP-04140R3 Rev. 0, "MP3 Rad Monitor Response to Core Uncovery."

4. MP-DWG-000-25212-27012 SH-00000000 Rev. 16, "Machine Location Cntmt Structure Plan El 51 Feet 4 Inches."

5. MP-DWG-000-25212-27013 SH-00000000 Rev. 13, "Machine Location Cntmt Structure Section 1-1 and 4-4."

6. MP-DWG-000-25212-27014 SH-00000000 Rev. 14, "Machine Location Containment Structure Section 2-2."

7. MP-DWG-000-25212-27015 SH-00000000 Rev. 12, "Machine Location Containment Structure Sect 3-3."

8. MP-DWG-000-25212-11060 SH-00000000 Rev. 9, "Plan Elevation 51 Feet 4 Inch Outline Containment Structure."

9. ETE-NAF-2014-0098 Rev. 0, "Millstone Unit 3 cycle 17 Nuclear Design Report."

10. M3AST-01942R3 Rev. 1, "Millstone 3 Alternate Source Term."

11. PNNL-15870 Rev. 1, "Compendium of Material Composition Data for Radiation Transport Modeling."

12. ET-NAF-06-0114 Rev. 0, "Dose Rate at the Containment Manipulator Crane Radiation Monitor Due to a Draindown Event Including Scatter from the air and Containment Dome."

13. PA-0227 Rev. 0, "Dose Rate at the Containment Manipulator Crane Radiation Monitor Due to a Draindown Event at North Anna or Surry."

14. MP-DWG-000-25212-11075 SH-00000000 Rev. 3, "Containment Structure Section 1-1."

Computer Codes Used:

SCALE 6.2 (Reference 1)

MCNP 6.1 (Reference 2)

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 46 of 54 Key Inputs and Assumptions Used:

1. Water level at the top of the active fuel

2. Reactor head removed *

3. Millstone 3 Core Inventory Methodology:

Source Term The LOCA core inventory source term will be decayed and converted to a photon spectrum using the ORIGEN code from the SCALE 6.2.2 code package.

Shielding Model The core, refueling cavity, and containment will be modeled using MCNP 6.1. The fuel region of the core will be modeled as a single homogeneous zone containing fuel and water. The water level is modeled as being level with the top of the active fuel. The gamma source in the fuel is distributed axially to model a nominal axial fuel burnup.

This distribution is intended to capture the self-shielding in the fuel zone. The gamma source has a uniform radial distribution in the fuel region. Structures inside the containment dome are limited to the crane wall and concrete structures close to the radiation monitors. These structures should reasonably model the photon backscatter to the radiation monitors. The radiation monitors will be modeled as finite air volumes used exclusively to tally the dose rate.

Results and/or

Conclusions:

Results from the MCNP output file provides an estimate of the average photon Mean Free Path (MFP) in each cell. An examination of these results indicates that the MFP in air varied between approximately 60 and 90 meters (197 and 295 feet). Given the distance between the operating floor and the containment dome, few photons interactions in air would occur during a photon's transit from the core to the containment dome and back to a radiation monitor.

Reference 3 also assumed that, other than air scatter, the primary contributor to radiation monitor dose rates was from photons that traveled vertically from the fuel through the air and scattered on the containment dome. This assumption was evaluated by modifying the MCNP input files to terminate any photon track that enters the refueling deck and refueling cavity concrete surfaces. The change was implemented by setting the importance of cell 10 to O (i.e. imp:p=O). The dose rates at the radiation monitors were reduced by -69% at RE-05A and -42% at RE-04A, bringing the results into reasonable agreement with the containment dome contributions calculated in Reference 3. analysis shows much of the dose rates at the radiation monitors in this calculation is due to photons scattering on or transiting through the refueling cavity concrete. Thus, while photon scattering on the containment dome is a contributor to the dose rates, the primary contributor is attributed to other concrete surfaces.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 47 of 54 It was found that a drain-down event earlier in a refueling outage could increase the count rates by a factor of three. This is consistent with the change in photon intensity.

In summary, containment radiation monitor dose rates at Millstone, North Anna, and Surry are expected to be between 3 R/hr and 40 R/hr, depending on the unit, time after refueling, and radiation monitor location in containment. A value of 3 R/hr would be a conservative dose rate for use in identifying a potential drain-down event.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 48 of 54 Calculation Summary Post-Accident Radiation Monitor Response for Core Uncovery The following pertinent information has been extracted from Calculation M2EP-04164R2, Verification of Rad Monitor Response to Core Uncovery. It is provided to assist technical reviewers that will be evaluating this license amendment request for MPS2.

Purpose:

The purpose of this calculation is calculate the radiological response of containment high range radiation monitors to core uncovery during refueling.

References:

1. M3EP-0414R3, Rev. 0, MP3 Rad Monitoring Response To Core Uncovery

2. ANS/SD-76/14, "Handbook of Radiation Shielding Data", July 1976

3. Cale. SFPGAMMA-04011 R2, Rev. 0, "MP2 Gamma Heating Analysis", dated 4/4/2003

4. MP2 Technical Specifications through Change# 318-01

5. Not Used

6. Not Used

7. Dwg 25203-29531, Rev.1, Millstone 2 Fuel Assembly

8. Dwg 25203-28014, Rev. 8, Instrument Location - Containment Plan El 14'-6" &38'-

6"

9. Dwg 25203-27022, Rev. 6, General Arrgt- Containment & Aux Bldg Section B-B

10. Dwg 25203-29141, Sh. 100, Rev. 1, Unit 2 - Reactor Arrangement Sectional Elevation Layout

11. SCALE 4.4a Code File Computer Code Used:

SCALE Package / QADS Key Inputs and Assumptions Used:

1. Water level at the top of the active fuel *

2. Reactor head removed

3. Millstone 2 Core Inventory Method of Analysis:

EALs require the ability to measure gamma radiation from the reactor core while under conditions where the water level in the reactor vessel is at the top of the active core.

Considering the location of the core relative to RM-8240 and 8241 , there does not appear to be any direct, line-of-sight communication of radiation. This requires crediting of air and concrete scattered radiations from the core to the rad monitors. Based upon a review of detector location (14'6" elevation on SG shield walls), and the need to credit multiple scattering there does not appear to sufficient radiation to register on these

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 49 of 54 detectors which have a low range limit of 1 R/hr. The insufficiency of radiation will be validated using RM-7891 which only requires a single scatter determination. RM-7891, the refuel floor rad monitor, is located adjacent to the refuel pool. No credit is taken for any radiation beyond single scattering.

The following scatter methods were validated in Reference 1. The methods are described in Reference 2 and applicable excerpts are retrievable as attachments to Reference 1 .

Scattering by air is addressed by a method using a line beam response factor (LBRF) to evaluate the single scatter from direct radiation to a receptor located out of the direct beam. In the case of a reactor core, the gamma flux directly over the core at the elevation of the radiation monitors is determined using the QADS shielding code. This flux is converted to an equivalent line beam source by multiplying the flux by the horizontal surface area of the core. The line beam source is multiplied by the LBRF to get the dose rate at the rad monitor from air scattered radiation.

Scattering by the containment dome is addressed using a method that determines an appropriate albedo factor and appropriate dose rate response that reflects a single scatter from the containment structure above the core ..

In this assessment, a lesser source term is more conservative to use (because it results in lower dose rates) and the twice burn, 5% enrichment provides a lesser source term than lower enrichments or single burns. The lesser source term results in a lower dose rate response ensuring that control room or SERO response to rad monitors used would be timely. In addition, since the core may be in a reduced inventory condition with containment closure not established, an assumption of 25 days decay will be assumed (this is roughly when draindown has occurred in order to put the reactor head back on the vessel).

This source term represents 1 fuel assembly decayed for 25 days using the ORIGENS code from the SCALE package (SQA Level 2). ORIGENS is a neutron depletion and decay code that will generate a source term that can be used in the QADS code with a multiplier to reach the equivalent of 217 assemblies. QADS is also from the SCALE package and is a point-kernel gamma shielding code. All of the fuel assemblies in a core will be modeled as a large cylindrical source. Receptor points will be located at the elevation of the rad monitors (for air scatter determination) and at the top of the containment dome (so that concrete/ steel scatter can be assessed).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 50 of 54 Summary:

Dose rate at R7891 from a reactor vessel draindown condition to top of active core, with no reactor head in place, crediting scattered radiation and a 25 day decay time for fuel as determined by this calculation are:

R7891, R/hr Air Scatter 1.4 Dome Scatter 3.0 Total 4.4

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 51 of 54 Calculation Summary Post-Accident Letdown Radiation Monitor Response The following pertinent information has been extracted from North Anna Calculation PA-0234, Rev. 1, Post-Accident Letdown Radiation Monitor Response. It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

This calculation provides the setpoint as a function of decay of the letdown line radiation monitor during accident condition for a 1% and 5% failed fuel (gap release) and iodine spike cases of 60 µCi/gm and 300 µCi/gm Dose Equivalent 1-131 released coolant activity in the RCS. During accident conditions, letdown radiation monitors support in determining the fuel failures that correspond to specific radiological criteria in the Emergency Action Levels (EALs).

References:

1. RF- DCP-000, 59-DCP-07-006, "Letdown Radiation Monitor Replacement/ North Anna/ Unit 1 & 2"

2. *Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

3. PA-DWG-000, 11715-FM-095A, Rev. 28, "FlowNalve Operating Numbers Diagrc~m Chemical and Volume Control System", North Anna Power Station -

Unit 1 .

4. RF-DCP-000, 59-DCP-94-013, "Letdown Radiation Monitor Replacement - North Anna Unit 1", September 1995

5. RF- DCP-000, 59-DCP-94-014, "Letdown Radiation Monitor Replacement- North Anna Unit 2", March 1995.

6. RF - CALC-NFL, PA-0195, Rev. 0, "Radiological Consequences of Fuel Handling Accident at North Anna Based on the Alternative Source Term", June 2003.

7. RF - CALC-NFL, PA-0219, Rev. 0, "Post Accident Response Curves for North Anna Primary Hot Leg Sample Lines", January 2005

8. RF - CALC-MEC, ME-0438, Rev. 2 "Reactor Coolant Letdown Radiation Monitor Setpoints", June 1995.

9. *PA-REF=OOO, NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U .S. Nuclear Regulatory Commission, Washington, D .C, 1988

10. *PA-REF-0, REG. Gu-1 .109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR PART 50, Appendix I," Rev .. 1, U.S. Nuclear Regulatory Commission, Washington, D .C, 1977.

11. RF-CODE-000, MicroShield, Grove Software, Inc, Verification 7 .02.

12. RF - CALC-RAD, PA-0246, Rev. 1, "Letdown Radiation Monitor Setpoint for North Anna", Apri1 2008.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 52 of 54 Computer Codes Used:

MicroShield Version 7.02 (Ref. 11)

Key Inputs and Assumptions Used:

1. Fuel rod gap fractions : 3% noble gas, 2% iodine, 5% cesium (Ref. 9)

2. Detector placed -0.5 inches from 2-inch pipe shielded with lead

3. North Anna Core Inventory Method of Analysis:

The 1% failed fuel is modeled in this calculation by calculating the dose rates resulting from 1% of the failed fuel gap inventory being released into the primary coolant that resulted following an accident. The 5% failed fuel was modeled by scaling up by a factor of 5 the results for the 1% failed fuel.

Dose rates to the monitor are calculated using MicroShield computer code. (Detector assembly is placed approximately 2 .5 inch (including O .5 inch of insulation of the pipe) from the center of a 2-inch diameter pipe source. The use of the MicroShield code is reasonable since the source is surrounded by lead shield and scattering is considered to be minimal. The MicroShield modeling assumes liquid source in a 1-inch radius pipe and 1O inches long with a thickness of O .154 inches. The dose point is assumed to be located at 2 .5 inch from the center (radial direction) of the source. The pipe and the detector are placed inside lead shield wall. It is assumed that the detector is located in the middle of the length of the pipe.

==

Conclusion:==

Tables 4 below provides summary of results of the letdown radiation monitor dose rates that can be used for the EAL radiological criteria.

The dose rate are calculated for the letdown line.

Table4 Nonh Anna Letdown Radiation Monitor Dose Dose Dose Dose Dose

'time(hr.) (Rlhr.) (R/hr,l ffiihr.) lMrr,)

!%FF S%FF 60 uCill!IIl 300uCi/gm 0.0

46.85 234.2S 24.SS

124.26 1.0 28.65 143.25 IS.20 7S.99 2.0 20.94 104.70 11.11 SS.S4 4.0 14.20 71.00 7.53 37,66 8.0 9.30 46.'iO 4.93 24.67 16.0 6.03 30.15 3.20 15.99 24.0 4.76 23.80 2.52 12.62 36.0 3.S6 19.30 2.05 10.24 48.0 3.41 17.0S 1.81 9.04 72.0 2.97 14.8S 1.58 7.88

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 53 of 54 Calculation Summary Post-Accident Letdown Radiation Monitor Response The following pertinent information has been extracted from Surry Calculation PA-0236, Rev. 0., Add. A, Post-Accident Letdown Radiation Monitor Response for Surry. It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

The purpose of this addendum is to determine the letdown radiation monitor (RM) response to source terms representative of gross reactor coolant activity and fuel failure that correspond to specific radiological criteria in the Emergency Action Levels (EALs) based on guidance from the Nuclear Energy Institute (NEI) for primary coolant activity level.

References:

1. NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels,"

January 2003.

2. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

3. Computer Code MicroShield, "MicroShield User's Manual", Grove Software, Inc.

4. MICROSHIELD-20170323-0-0, "MicroShield V. 7.02 Periodic Effectiveness Review 2017, Code Manager change and Code Owner change."

5. 958.398ABS Rev. A, "Calibration of a 903664 Letdown Monitor and 943-36 Detector", Victoreen, Inc., June 1996.

6. 11448/11548-7.57-148 Sheet 1, "Hi & Lo Range Letdown Monitor, June 1970 (Detector Shield Arrangement- Vendor Drawing 903664).

7. 11448/11548-7.57-26ASheet 1, "Connections of Letdown Monitor, June 1970 (Sample line tubing details - Vendor Drawing 903742).

8. 11448/11548-7.57-268 Sheet 1, "Connections of Letdown Monitor, June 1970 (Sample line tubing details - Vendor Drawing 903742).

9. 903787, "[Aluminum] Spacer Hi-Lo", August 1970.

10. 903751 Rev. X1, "[Lead] Plug", August 1970.

11. RA-0008 Rev. 0 through Addendum C, "Core Isotropic Inventories for Surry Dose Consequences Analyses Based on the Alternate Source Term", July 2010.

12. ETE-NAF-2017-0052 Rev. 1, "Input to a Proposed Surry License Amendment Request Adopting TSTF-490-A Rev. 0 (Dose Equivalent Xe-133) and Updated Alternative Source Term Analyses", January 2018.

13. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, October 1988.

14. RA-0070 Rev. 0 through Addendum A, "Radiological Consequences of a Steam Generator Tube Rupture (SGTR) at Surry Power Station Based on the Alternative Source Term (AST)", May 2017.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/4 7/55/56 Enclosure 6 Page 54 of 54

15. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors For Inhalation, Submersion, and Ingestion", EPA 520/1-88-020, September 1988.

16.SQA-WATPROP-D-20170608, "SQA Documents for the Initial Release of WATPROP-D and Initial Periodic Effectiveness Review", June 2017.

Computer Codes Used:

MicroShield Version 7 .02 [References 3 and 4]

Key Inputs and Assumptions Used:

1. Fuel rod gap fractions : 3% noble gas, 2% iodine, 5% cesium (Ref. 9)

2. Detector response from 0.5 inch sample tubing through lead shield

3. Surry Core Inventory Methodology:

Letdown radiation monitor response in counts per minute (cpm) was determined for accident source terms of 1% failed fuel, 10 µCi/cc DE 1-131, and 300 µCi/cc DE 1-131 at 0, 1, 2, 4, 8, 16, 24, 36, 48, and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown. This was accomplished by using MicroShield to determine dose rate at the detector for the 1% failed source term at O hours. MicroShield was used to decay the 1% failed fuel source and determine the dose rates at the detector for each of the subsequent time steps. MicroShield dose rates were also determined for calibration standards of Co-60 and Mn-54 for which the vendor had determined cpm. Conversion factors for cpm to Dose rate for Co-60 and Mn-54 were derived and applied to the dose rate results for the 1% failed fuel source term at each time step.

==

Conclusion:==

The letdown radiation monitor response in counts per minute (cpm) to accident source terms of 1% failed fuel, 10 µCi/gm DE 1-131, and 300 µCi/gm DE 1-131 at 0, 0.5, 1, 2, 4,

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 6 Page 55 of 54 8, 16, 24, 36, 48, and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown are documented in Table 4.

Table 4: Surry Letdown Radiation Monitor Response

{cpm)

Time (hr.) 1%FF 10 µCi/gm 300 µCi/gm 0.0 l.30E+07 1.18E+06 3.53E+07 0.5 5.98E+06 5.40E+05 1.62E+07 1.0 4.02E+o6 3.62E+05 1.09E+07 2.0 2.33E+06 2.11E+05 6.32E+06 4.0 l.32E+06 l.19E+05 3.57E+06 8.0 7.05E+05 6.37E+04 1.91E+06 16.0 3.08E+05 2.78E+04 8.35E+05 24.0 1.81E+05 1.64E+04 4.91E+05 36.0 l.16E+05 1.04E+04 3.13E+05 48.0 9.34E+04 8.43E+03 2.53E+05 72.0 7.94E+04 7.17E+03 2.15E+051

ML19249B776

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