Text

Submittal of Proposed License Amendment Request to Revise Technical Specification Limits for Primary and Secondary Coolant Activity
	ML19218A177
Person / Time
Site:	Millstone
Issue date:	07/30/2019
From:	Mark D. Sartain; Dominion Energy Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	19-243
	Download: ML19218A177 (61)
	v • d • e

Dominion Energy Nuclear Connecticut, Inc. 5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Energy.com July 30, 2019 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555 DOMINION ENERGY NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 Dominion :;iiiii" Energy Serial No NRA/SS Docket No. License No.19-243 RO 50-336 DPR-65 PROPOSED LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION LIMITS FOR PRIMARY AND SECONDARY COOLANT ACTIVITY In accordance with the provisions of 1 O CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a request for an amendment to th.e Technical Specifications (TS) for Millstone Power Station Unit 2 (MPS.2). DENG proposes to reduce the TS Reactor Coolant System (RCS) and secondary side specific activity by 50%. This change impacts the primary and secondary liquid source terms. The proposed TS 3.4.8 limits for RCS concentration are 0.5 µCi/gm Dose Equivalent (DE) 1-131 and 550 µCi/gm DE Xe-133, and the maximum iodine concentration allowed by the TS as the result of an iodine spike is 30 µCi/gm DE 1-131. Secondary side radionuclide concentrations are based on the proposed TS 3.7.1.4 lirnitof0.05

µCi/gm DE 1-131. These TS changes are based on evaluations that were conducted to assess the radiological consequences following postulated design basis Main Stearn Line Break and Stearn Generator Tube Rupture accidents to address analysis deficiencies documented in the corrective action program. A reduction in the TS RCS and secondary side specific activity is required to meet the control room dose regulatory limits. Additionally, the proposed reduction in TS RCS and secondary side specific activity will provide inherent source term margin. The analysis results from the reanalyzed events meet the acceptance criteria as specified in 10 CFR 50.67, 10 CFR 50 *Appendix A Generic Design Criterion 19, Standard Review Plan (SRP) 15.0.1, and Regulatory Guide 1.183. Attachment 1 provides the description and assessment of the proposed changes. Attachment 3 provides marked-up MPS2 TS pages showing the proposed change. Attachment 4 provides marked-up MPS2 TS Bases pages showing the proposed change. The TS Bases mark-ups are provided for information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program after this LAR is approved.

Serial No: 19-243 Docket Nos. 50-336 Page 2 of 3 The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in Attachment

2. DENC has determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be release.d offsite or any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. The Facility Safety Review Committee has reviewed and concurred with the determinations herein. DENC requests approval of this license amendment request by July 31, 2020, with implementation within 60 days of issuance.

In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of Connecticut.

If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely, Mark D. Sartain Vice President

-Nuclear Engineering and Fleet Support COMMONWEALTH OF VIRGINIA ) ) . COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President

-Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief. ""' Acknowledged before me this 3o day of _jy3 2019 *

MyCommissionExpires:

"'° :!.\ zo1, a~.-,.,_+

GARY DON MILLER Notary Public Commonwealth of Virginia Reg. # 7629412 *My Commission Expires August 31, 20.L!

Attachments: 1 . Description and Assessment of Proposed Changes Serial No: 19-243 Docket Nos. 50-336 Page 3 of 3 2. No Significant Hazards Consideration Determination and Environmental Consideration

3. Marked-Up Technical Specification Pages for MPS2 4. Marked-Up Technical Specification Bases Pages for MPS2 for Information Only Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 ATTACHMENT 1 Serial No.19-243 Docket No. 50-336 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES DOMINION ENERGY NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNITS 2 Table of Contents Seria.1 No.19-243 Docket No. 50-336 Attachment 1, Page 1 of 45

1.0 INTRODUCTION

AND BACKGROUND

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

3

1.1 INTRODUCTION

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

3 1.2 CURRENT LICENSING BASIS

SUMMARY

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

4 1.3 SELECTION OF EVENTS REQUIRING REANALYS[S

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

4 1.4 ANALYS[S ASSUMPTIONS AND KEY PARAMETER VALUES ............................................................

7 2.0 PROPOSED LICENSING BASIS CHANGES ..........................................................................

10 2.1 DECREASE OF THE TS RCS AND SECONDARY SIDE SPECIF[C ACT[VITY ....................................

10 2.2

SUMMARY

OF DESIGN AND LICENS[NG BASIS CHANGES ............................................................

10 3.0 RADIOLOGICAL EVENT RE-ANALYSES AND EVALUATION

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

17 3.1 MAIN STEAM LINE BREAK ACCIDENT .......................................................................................

18 3. 1. 1 MSLB Scenario Description

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

.' .... 18 3.1.2 MSLB Source Term Definition

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

19 3.1.3 MSLB Release Transport

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

23 3. 1. 4 Determination of Atmospheric Dispersion Factors ............................................................

29 ' 3.1.5 MSLB Key Analysis Assumptions and Inputs ...................................................................

30 3.1.6 MSLB Analysis Results ....................................................................................................

33 3.2 STEAM GENERATOR TUBE RUPTURE ACCIDENT..

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

34 3.2.1 SGTR Scenario Description

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

34 3. 2. 2 SG TR Source Term Definition

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

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

35 3.2.3 SGTR Release Transport

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

36 3. 2. 4 Determination of Atmospheric Dispersion Factors ............................................................

40 3.2.5 SGTR Key Analysis Assumptions and Inputs ...................................................................

40 3.2.6 SGTR Analysis Results ....................................................................................................

43

4.0 CONCLUSION

S

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

44

5.0 REFERENCES

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

45 Index of Tables Serial No.19-243 Docket No. 50-336 Attachment 1, Page 2 of 45 Table 1.4-1: Common Assumptions and Key Parameters

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

8 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses ..............................................................................................................................................

11 Table 3. 0-1: Accident Dose Acceptance Criteria ................................

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

17 Table 3.1-1: RCS and Secondary Side Liquid Source Term ................................................................

20 Table 3.1-2: Pre-Accident Iodine Spike ..............

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

21 Table 3.1-3: Concurrent Iodine Spike ..................................................................................................

21 Table 3.1-4: MP2 Core Inventory

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

22 Table 3.1-5: Faulted SG Steam Flow Rates .........................................................................................

24 Table 3.1-6: Before 20 Seconds -Intact SG Steam Flow Rates ...........................................................

27 Table 3.1-7: After 20 Seconds -Intact SG Steam Flow Rates ..............................................................

28 Table 3.1-8: MP2 Control Room X/Qs for MSLB ..................................................................................

29 Table 3.1-9: Basic Data and Assumptions for MSLB ............................................................................

30 Table 3.1-10: Dose Summary for the MSLB Accident Outside Containment..

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

33 Table 3.1-11: Dose Summary for the MSLB Accident Inside Containment..

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

33 Table 3.2-1: Concurrent Iodine Spike ..................................................................................................

36 Table 3.2-2: Ruptured SG Break Flow Rates and ADV/MSSV Steam Flow Rates ...............................

38 Table 3.2-3: Intact SG Primary to Secondary Leakage and ADV/MSSV Steam Flow Rates ................

40 Table 3.2-4: Basic Data and Assumptions for SGTR ...........................................................................

40 Table 3.2-5: Dose Summary for the SGTR Accident. ...........................................................................

43 Index of Figures Figure 3.1-1: Faulted SG Primary to Secondary Leakage to Environment

{Outside Containment)

....... 23 Figure 3.1-2: Faulted SG Secondary Side Liquid Leakage to Environment (Outside Containment)

..... 24 Figure 3.1-3: Faulted SG Primary to Secondary Leakage to Environment (Inside Containment)

.........

25 Figure 3.1-4: Faulted SG Secondary Leakage to Environment (Inside Containment)

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

25 Figure 3.1-5: Intact SG Primary to Secondary Leakage to Environment

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

26 Figure 3.1-6: Intact SG Secondary Side Liquid Leakage to Environment.

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

27 Figure 3.2-1: Ruptured SG Break Flow to Environment

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

37 Figure 3.2-2: Ruptured SG Secondary Side Liquid to Environment.

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

37 Figure 3.2-3: Intact SG Primary to Secondary Leakage to Environment

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

39 Figure 3.2-4: Intact SG Secondary Side Liquid Leakage to Environment..

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

39 1.0 Introduction and Background 1.1 Introduction Serial No.19-243 Docket No. 50-336 Attachment 1, Page 3 of 45 This report describes the evaluations conducted to assess the radiological consequences at Millstone Unit 2 following postulated design basis Main Steam Line Break (MSLB) and Steam Generator Tube Rupture (SGTR) accidents per Regulatory Guide (RG) 1-.183 [Reference 1] that were required to address analysis deficiencies documented in the corrective action program. This application reduces the Technical Specification (TS) Reactor Coolant System (RCS) and secondary side specific activity by 50%. This change impacts the primary and secondary liquid source terms. The proposed TS 3.4.8 limits for RCS concentration are 0.5 µCi/gm Dose Equivalent (DE) 1-131 and 550 µCi/gm DE Xe-133, and. the maximum iodine concentration allowed by the TS as the result of an iodine spike is 30 µCi/gm DE 1-131. Secondary side concentrations are based on the proposed TS 3.7.1.4 limit of 0.05 µCi/gm DE 1-131. The accident source term discussed in Reference 1 is herein referred to as the Alternative Source Term (AST). The evaluations have employed the detailed methodology contained in RG 1.183 for use in design basis accident analyses for the AST. The results have been , compared with the acceptance criteria contained in 10 CFR 50.67 [Reference 2], 10 CFR 50 Appendix A General Design Criterion 19 [Reference 14], Standard Review Plan SRP-15.0.1

[Reference 15], and the supplemental guidance in RG *1.183. The proposed radiological dose analyses were performed with a controlled version of the computer code RADTRAD-NAI

[Reference 3]. The RADTRAD-NAI computer code calculates the control room and offsite doses resulting from releases of radioactive isotopes based on user supplied atmospheric dispersion factors, breathing rates, occupancy factors, and dose conversion factor~. Control room and offsite Atmospheric Dispersion Factors (X/Qs) were approved in the Safety Evaluation Report (SER) for the Millstone Unit 2 AST license amendment

[Reference 1 O]. Innovative Technology Solutions (ITS) of Albuquerque, New Mexico developed the RADTRAD code for the NRC. The original version of the NRC RADTRAD code was documented in NUREG/CR-6604

[Reference 4]. The Numerical Applications, Inc. (NAI) version of RADTRAD was originally derived from NRC/ITS RADTRAD, Version 3.01. Subsequently, RADTRAD-NAI was changed to conform to NRC/ITS RADTRAD, Version 3.02 with additional modifications to improve usability.

Numerical Applications is currently a division of Zachry Nuclear Engineering, Inc. The RADTRAD-NAI code is maintained under Zachry Nuclear Engineering, lnc.'s Quality Assurance (QA) program, which conforms to the requirements of 10 CFR 50, Appendix B.

1.2 Current Licensing Basis Summary Serial No.19-243 Docket No. 50-336 Attachment 1, Page 4 of 45 The current design basis radiological analyses that appear in the Millstone Unit 2 Final Safety Analysis Report (FSAR) consist of assessments of the following events: 1) Main Steam Line Break (MSLB) 2) Control Rod Ejection Accident (CREA) 3) Steam Generator Tube Rupture (SGTR) 4) Loss of Coolant Accident (LOCA) 5) Fuel Handling Accident (FHA) 6) Spent Fuel Cask Drop Accident 1 7) Waste Gas System Failure (WGSF) The MSLB, CREA, SGTR, LOCA, FHA and Cask Drop/Tip Accident were reanalyzed and the WGSF was relocated to FSAR Chapter 11 as a part of the AST implementation approved in Reference

10. 1.3 Selection of Events Requiring Reanalysis The proposed licensing and plant operational changes are discussed in Section 2.0. These changes require corresponding changes to the Millstone Unit 2 Operating License and Technical Specifications, which are also described in Section 2.0 of this report. A non-conservatism was identified for the Millstone Unit 2 FSAR Section 14.6.3 SGTR radiological consequence analysis.

It was identified that the RADTRAD-NAI code inputs for the ruptured steam generator (SG) iodine spike cases incorrectly used a partitioning coefficient 2 of 250 (0.4% release) instead of 100 (1 % release) for the primary to secondary leak that does not flash to steam. The text of the SGTR analysis of record and the AST license amendment request [Reference 9] indicate a PC of 100. The modeling of the iodine released from the ruptured SG that flashes to steam was modeled correctly with a PC of 1. The implication of this latent error is that iodine releases from the primary to secondary leak that spills into the ruptured SG would be modeled as 2.5 times larger for the portion of the SGTR iodine spike case releases that used the incorrect PC. Correcting the PC produces dose consequence results that were within the regulatory limits; however, the increase in dose consequence was more than a minimal increase when compared to the current SGTR analysis of record approved in Reference

9. An administrative limit of 85% for TS RCS specific activity 1 The Unit 2 Spent Fuel Cask Drop or Cask Tip Accident is analyzed as a subset of a FHA It is an event in the spent fuel pool where a shielded cask tips over and damages all fuel within the potential impact area. 2 Partitioning coefficient (PC) per Reference 1 is defined as: PC (mass of h per unit mass of liquid)/ (mass of Ii per unit mass of gas). The FSAR uses the term partition factor which would be 0.01 (or 1 %) for a partitioning coefficient of 100.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 5 of 45 DE 1-131 was applied to maintain the control room dose consequences less than the analysis of record. The MSLB and SGTR dose calculations previously utilized a value of 0.4% for SG moisture carryover (MCO), which was two times the maximum full-power MCO value of 0.2% specified by the SG manufacturer.

The results of a recent MCO test conducted at Millstone Unit 2 concluded that the average measured MCO value at hot full-power conditions is greater than the 0.4% value used in the MSLB and SGTR dose calculations.

The MCO at full-power does not represent the MCO expected following a reactor trip where doses are accumulated following a MSLB or SGTR accident.

As such, the full power measured MCO value is not an input to the dose calculations.

Following the reactor trip, steam flows via the atmospheric dump valves (ADVs) or main steam safety valves (MSSVs) are much less than those generated at full power. This would result in a lower MCO value post trip than the full power value; therefore, the basis for the MCO value used in the MSLB and SGTR dose calculations was revisited.

Assuming the SGs are flooded to the secondary separator deck, a post trip MCO value is calculated three feet above the pool surface using the correlation for the momentum controlled region in the low gas flux regime contained in NUREG/CR-3304 "Mechanistic Modeling and Correlations for Pool Entrainment Phenomenon" [Reference 13]. Using this correlation, the 0.4% value used in the previous dose calculations was justified as conservative.

The MCO was increased to 1 % for added analysis margin in the proposed MSLB and SGTR dose calculations supporting the license amendment request. A non-conservatism was also identified in the Millstone Unit 2 FSAR Section 14.1.5 MSLB radiological consequence analysis.

It was identified that the first 20 seconds of the MSLB event (i.e., prior to closure of the main steam isolation valves (MS!Vs)) involves intact SG steam release rates that are large enough to entrain a significant fraction of water (i.e., water

entrainment well beyond the range of available correlations).

During the first 20 seconds of the event prior to MSIV closure, the main steam flow from the in.tact SG to the break location via main steam header briefly approaches 300% of the full power steam flow value. The MSLB analysis of recprd models the intact SG iodine release path with a PC of 100, in accordance with RG 1.183. However, because there is no available basis for estimating water entrainment at these higher initial steam velocities (greater than velocities at 100% power), releases for this time period are conservatively modeled with a MCO of 100% and a PC of 1 (no partitioning).

Additionally, during this period of time, the X/Q value in the limiting analysis case is conservatively modeled as a break release instead of an ADV release. After the first 20 seconds, the MCO is appropriately modeled as entrainment from a pool as discussed above. Correcting for fhe PC and break release X/Q produces dose consequence results that are within the regulatory limits when the aforementioned administrative reduction in TS RCS specific activity is assumed. Conservatism inherent in the modeling of the intact SG initial Serial No.19-243 Docket No. 50-336 Attachment 1, Page 6 of 45 activity was credited to maintain the corrected MSLB dose consequences less than the current analysis of record. The SGTR and MSLB analyses deficiencies are documented in the corrective action program. As the changes to the MSLB and SGTR analyses result in more than a minimal increase in dose consequences, a license amendment request is required.

A reduction in the TS RCS and secondary side specific activity is required to meet the control room dose analysis of record. In order to provide inherent source term margin, it is proposed that TS RCS and secondary side specific activity be reduced to 50% of the Current Licensing Basis (CLB). The unfiltered inleakage to the Millstone Unit 2 control room was increased to allow for additional operational flexibility.

Sections 3.1 and 3.2 provide the detailed description of the re-analyses for these events.

1.4 Analysis Assumptions and Key Parameter Values Serial No.19-243 Docket No. 50-336 Attachment 1 , Page 7 of 45 This section describes the general analysis approach and presents analysis assumptions and key parameter values that are common to the accident analyses performed.

Sections 3.1 and 3.2 of this Attachment provide specific assumptions that were employed for the MSLB and SGTR, respectively.

The dose analyses documented in this application employ the Total Effective Dose Equivalent (TEDE) calculation method as specified in RG 1.183 for AST applications.

TEDE is determined at the Exclusion Area Boundary (EAB) for the worst 2-hour interval.

TEDE for individuals at the Low Population Zone (LPZ) and for the Millstone Unit 2 control room personnel are calculated for the assumed duration of the event. The TEDE concept is defined to be the Deep Dose Equivalent, DOE (from external exposure) plus the Committed Effective Dose Equivalent, CEDE (from internal exposure).

In this manner, TEDE assesses the impact of a!! relevant nuclides upon al! body organs. The ODE is nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure if the whole body is irradiated uniformly.

Since this is a reasonable assumption for submergence exposure situations, EDE is used in lieu of DOE in determining the contribution of external dose to the TEDE. EDE dose conversion factors .were taken from Table 111.1 of Federal Guidance Report (FGR) 12 [Reference 6] per Section 4.1.4 of Reference

1. CEDE dose conversion factors were taken from Table 2.1 of FGR 11 [Reference 5] per Section 4.1.2 of Reference

1. There are a number of common analysis assumptions and plant features that are used in the analysis of all of the events. These items are unchanged from the AST license amendment request submittal

[Reference 9] and are presented in Table 1.4-1.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 8 of 45 Table 1.4-1: Common Assumptions and Key Parameters Assumption/

Parameter Value Loss of Offsite Power Timing

Assumed to occur at accident initiation Rated Thermal Power (2700 MWt) plus 2% calorimetric uncertainty 2754 MWt (used for steam releases and coolant activity source terms) Dose Conversion Factors

Table 2.1 of Reference 5

Table 111.1 of Reference 6 Control Room Control Room Normal Intake Flow Rate prior to Isolation 800 cfm Control Room Emergency Ventilation System Recirculation Flow 2,250 cfm Rate Response Time to Isolate upon Receipt of Control Room 20 seconds Ventilation Radiation Monitor Alarm or Safety Injection Signal [delay for damper operation]

Time Credited for Operator Action to Align Control Room 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Emergency Ventilation after Isolation to Filtered Recirculation Mode Control Room Emergency Ventilation Filter Efficiencies 3 Elemental:

90% Organic: 70% Particulate:

90% Iodine Chemical of Primary to Secondary Leakage Elemental:

97% Organic: 3% Particulate:

0% Wall and Ceiling Concrete Thickness

.=: 2 feet Control Room Occupancy Factors 0 -24 hours 1.0 24-96 hours 0.6 96 -720 hours 0.4 Breathing Rates Control Room 0 -720 hours 3.SOE-04 m 3/sec 3 90% efficiency for elemental/particulate and 70% for organic are conservatively used in the analysis.

TS 4.7.6.1.c.2 states that the charcoal filter has to have a removal efficiency of~ 95% and TS 4.7.6.1.f states that HEPA filter efficiency has to have a removal efficiency of~ 99%.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 9 of 45 Table 1.4-1: Common Assumptions and Key Parameters Assumption I Parameter Value Low Population Zone (LPZ) O -8 hours 3.50E-04 m 3/sec 8 -24 hours 1.80E-04 m 3/sec 24 -720 hours 2.30E-04 m 3/sec Atmospheric Dispersion Factors EAB X/Q Enclosure Building -ground level 3.66E-04 sec/m 3 LPZ X/Qs Enclosure Building -ground level 0-4 hours 4.80E-05 sec/m 3 4 -8 hours 2. 31 E-05 sec/m 3 8-24 hours 1.60E-05 sec/m 3 24-96 hours 7 .25E-06 sec/m 3 96 -720 hours 2.32E-06 sec/m 3 MP2 Control Room X/Qs Atmospheric Dump Valves (ADVs) 0 -2 hours 7.40E-03 sec/m 3 2-8 hours

5.71E-03 sec/m 3 8 -24 hours 2.13E-03 sec/m 3 24-96 hours 1.74E-03 sec/m 3 96 -720 hours 1.43E-03 sec/m 3 Reactor Coolant System (RCS) Leakage Limits Total Accident Induced Primary to Secondary Leak Rate 150 gal/day 0.87 lbm/min 2.0 Proposed Licensing Basis Changes Serial No.19-243 Docket No. 50-336 Attachment 1, Page 1 O of 45 This section provides a summary description of the key proposed licensing basis changes based on the Millstone Unit 2 AST analyses accompanying this license amendment request. 2.1 Decrease of the TS RCS and Secondary Side Specific Activity This application reduces the TS RCS and secondary side specific activity by 50%. This change impacts the primary and secondary liquid source terms. Table 2.2-1 provides the CLB values and the proposed values for TS specific activity.

RCS radionuclide concentrations are based on the proposed TS 3.4.8 limits of 0.5 µCi/gm DE 1-131 and 550 µCi/gm DE Xe-133. The maximum iodine concentration allowed by the proposed Technical Specification as the result of an iodine spike is 30 µCi/gm DE 1-131. This value is treated as the pre-accident iodine spike. The concurrent iodine spike appearance rates are based on either 500 times (MSLB) or 335 times (SGTR) the appearance rates required to maintain the coolant activity at the proposed 0.5 µCi/gm DE 1-131 concentration.

Secondary side concentrations are based on the proposed TS 3.7.1.4 limit of 0.05 µCi/gm DE 1-131. The proposed limits are reflected throughout the Limiting Condition for Operation and Actions for TS 3.4.8 and TS 3.7.1.4, as well as the Surveillance Requirements for TS 3.4.8. Marked up pages for the proposed TS changes are provided in Attachment

3. Marked up TS Basis pages are provided in Attachment 4 for information.

2.2 Summary of Design and Licensing Basis Changes This section provides a summary of the proposed changes to the CLB values for the MSLB and SGTR analyses.

A summary is provided in Table 2.2-1 and is broken into segments detailed in Section 3.0. The existing analyses for these radiological events were performed as part of the Millstone Unit 2 AST license amendment request [Reference 9] and approved in Reference

10. The proposed amendment reduces the TS specific activity to provide margin. In addition, some margin was added to accommodate thermal-hydraulic changes. This accounts for differences in some of the parameters list_ed in Table 2.2-1.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 11 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Parameter CLB Value Proposed Value Reason for Change . i.* Code Version ... . RADTRAD Version 1.1a (QA) 1.3 (QA) RADTRAD Version 1.1 a is retired and no longer supported by Zachry Nuclear EngineerinQ, Inc. ' :: .. ,, :: . ' .. Source Terhi *. / .. . . . ' <-Primary Coolant Iodine TS Change Specific Activity TS Limits DE 1-131 1.0 µCi/gm 0.5 µCi/gm Pre-Accident Spike 60 µCi/gm 30 µCi/gm Primary Coolant Noble Gas 1100 µCi/gm DE Xe-133 4 550 µCi/gm DE Xe-133 TS Change Activity Primary Coolant Iodine Consistent with the TS Change Concentrations at TS Limit Isotope b!Ci/gm b!Cilgm !-131 5.62E-01 2.81E-01 . !-132 3.10E+OO 1.55E+OO 1-133 1.85E+OO 9.25E-01 1-134 5.14E+OO 2.57E+OO 1-135 3.66E+OO 1.83E+OO Pre-Accident Iodine Spike 60 µCi/gm DE 1-131 30 µCi/gm DE 1-131 The pre-accident iodine source Isotope 1:JCi/gm b!Ci/gm term was reduced by half to 1-131 3.37E+01 1.69E+01 remain consistent with the 1-132 1.86E+02 9.30E+01 proposed new TS DE 1-131 limits 1-133 1.11E+02 5.55E+01 for the primary coolant specific 1-134 3.09E+02 1.55E+02 activity.

1-135 2 19E+02 1.10E+02 Secondary Iodine Activity 0.1 µCi/gm 0.05 µCi/gm DE 1-131 TS Change Concentration 4 Reference 9 approved primary side concentration based on the TS limits of 100/Ebar*

The coolant activity associated with the 100/Ebar value is equivalent to the DE Xe-133 limit of 1100 µCi/gm. The change to DE Xe-133 was part of the adoption of TSTF-490 [Reference 11] and approved by the NRC in Amendment No. 307 [Reference 12].

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 12 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Control Room Unfiltered lnleakage (cfm) Control Room Effective Volume EAB Breathing Rate 0-8 hours 8 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 -720 hours CLB Value Proposed Value 200 35,656 ft 3 Varies with time 3.50E-04 m 3/sec 1.80E-04 m 3/sec 2.30E-04 m 3/sec 250 35,650 ft 3 Constant at maximum rate 3.50E-04 m 3/sec 3.SOE-04 m 3/sec 3.SOE-04 m 3/sec 5 The maximum control room inleakage results for the past three successful tests are provided below. 2008: 133 cfm 2012: 171 cfm 2018: 157 cfm Unfiltered inleakage to the Millstone Unit 2 control room was increased to provide more o erational flexibilit 5* Net affected volume = (gross volume minus suspended ceiling) (volume of control cabinets, panels & solid items) + (volume of communicating ductwork & equipment (fans)) 41,726 ft3 -8,281 ft 3 + 2,205 ft 3 = 35,650 ft 3 Use the maximum breathing rate (at 0-8 hours) for all times.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 13 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Parameter CLB Value Proposed Value Reason for Change ' Main Steam Lin.e Break Accident (Section 3.1) .. * ., RCS Mass 428,400 lbm 423,000 lbm Consistent with the SGTR analysis and results in higher specific activity with failed fuel. Faulted SG Steam Release The faulted SG is assumed to steam The faulted SG is assumed to It is conservative for the majority dry within 750 seconds. No steam steam dry within 750 seconds. of the initial iodine inventory to be flow rates were provided in Reference This release continues for 5 removed; therefore, extending

9. hours to ensure initial iodine the release out to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is inventory is removed. conservative and allows for a greater amount of iodine Refer to Table 3.1-5 inventory to be released.

Intact SG Steam Release Time (sec) Integral Mass Release Refer to Tables 3.1-6 and 3.1-7 Increase mass releases by (lbm) approximately 12% for increase 20 54,334 in margin. 800.1 54,334 10,000 534,459.42 60,000 1,979,171.92 Peaking Factor 1.69 1.79 COLR value (1.69) plus 6% uncertainty.

Enclosure Building Filtration Credit at 250 seconds No Credit No credit is taken for the EBFS, System (EBFS) this is conservative.

Containment Leak Rate (La) Time Leak Rate Time (hours) Leak Rate No credit is taken for the EBFS; 0 sec La 0 La therefore, there is no bypass or 250 sec 0.014*La {bypass) filtered releases starting at 250 0.986*La (filtered) seconds. Leak Rate Reduction at 24 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.014*La*0.5 (bypass) 24 La*0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 0. 986*La *O. 5 (filtered)

La is unchanged from 0.5% containment air weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (TS 3.6.1.2).

Leak rate reduction to 50% after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is unchanged.

..

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 14 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Parameter SG Iodine PC in Intact SG MCO in Intact SG Concurrent Iodine Spike Isotope 1-131 1-132 1-133 1-134 1-135 CLB Value 100 for all times after release 0.4% for all times after release 500 x Appearance Rate µCi/sec 1.43E+06 3.48E+07 6.31E+06 1.31E+08 2.01E+07 Proposed Value 1 for time < 20 seconds 100 for time >= 20 seconds 100% for time < 20 seconds

1 % for time >= 20 seconds 500 x Appearance Rate µCi/sec 7.15E+05 1.74E+07 3.16E+06 6.55E+07 1.01 E+07 Reason for Change A PC of 100 (1%) is assumed for the intact SG after the first 20 seconds due to the assumption that the tube bundles will remain covered after the MSIV is closed. Prior to the MSIV closing, the intact SG is assumed to release

through the break with a PC = 1.0 (100%). No reduction of releases from the faulted SG for partitioning are taken because the faulted SG is assumed to dry out. MCO was increased to a more conservative value and to align with the iodine PC. Prior to the MSIV closing, the intact SG is assumed to release through the break with a PC= 1.0 (100%). No reduction of releases from the faulted SG for MCO are taken -because the faulted SG is assumed to dry out. The concurrent iodine source term was reduced by half to remain consistent with the proposed new TS DE 1-131 limits for the primary coolant specific activity.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 15 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Parameter CLB Value Proposed Value Reason for Change Steam GeneratorTube Rupture.Accident Section 3.2) . ' *, ; . Ruptured SG Break Flow Time Total Flashed Liquid Time Total Flashed Liquid For times from 0 to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Rates period Break Break Break period Break Break Break increase mass releases by 10% Flow Flow Flow Flow Flow Flow for increase in margin. For time FromJTo FromjTo after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , increase mass (hour) lbm lbm lbm (hour) lbm lbm lbm releases by approximately 12% 0 11 150,000 5,000 145,000 0 11 165,000 5,500 159,500 for increase in margin. After 1 51,600 1,200 50,400 After 1 57,895 1,346 56,549 hour0.00635 days 0.153 hours 9.077381e-4 weeks 2.088945e-4 months hour Refer to Table 3.2-2 for rates Ruptured SG ADV/MSSV Time (hours) ADV/MSSV Mass Time (hours) ADV/MSSV Mass For times from 0 to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Steam Flow Rates (lbm) (lbm) increase mass releases by 10% 0 to 1 1.700E+05 Oto 1 1.870E+05 for increase in margin. For time After 1 9.200E+04 After 1 1.032E+05 after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , increase mass releases by approximately 12% for increase in margin. Refer to Table 3.2-2 for rates Intact SG ADV/MSSV Steam Time (hours) ADV/MSSV Mass Refer to Table 3.2-3 For times from 0 to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Flow Rates (lbm/min) increase mass releases by 10% Oto 1 2.000E+03 for increase in margin. For time 1 to 1.11 7.330E+03 after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , increase mass 1.11to1.71 5.147E+03 releases by approximately 12% 1.71 to 2.33 4.200E+03 for increase in margin. 2.33 to 2.74 3.840E+03 2.74 to 3.18 3.810E+03 3.18 to 3.72 3.780E+03 3.72 to 6.50 2.743E+03 6.50 to 17.61 2.151 E+03 Serial No.19-243 Docket No. 50-336 Attachment 1, Page 16 of 45 Table 2.2-1: Comparative Summary of Design and Licensing Basis Changes to the MSLB and SGTR Analyses Parameter CLB Value Proposed Value Reason for Change Initial SG Liquid 280,000 80,000 Determining steam generator Volume/Mass (lbm) (used to maximize curies in the SGs) (used to minimize holdup in SGs activity using the CLB value of and maintains curies consistent 280,000 lbm is over-conservative 80,000 with proposed TS limit) and resulted in modeling (used to minimize holdup in SGs} secondary coolant activity greater than that stated in the TS. MCO 0.4% 1% MCO was increased to a more conservative value and to align with the iodine PC. The value is bounding of applicable values for the duration of the release. Concurrent Iodine Spike 335 x Appearance Rate 335 x Appearance Rate The concurrent iodine source Isotope µCi/sec µCi/sec term was reduced by half to 1-131 9.56E+05 4.78E+05 remain consistent with the 1-132 2.33E+07 1.17E+07 proposed new TS DE 1-131 limits 1-133 4.23E+06 2.12E+06 for the primary coolant specific 1-134 8.78E+07 4.39E+07 activity.

1-135 / 1.35E+07 6.75E+06 3.0 Radiological Event Re-Analyses and Evaluation Serial No.19-243 Docket No. 50-336 Attachment 1, Page 17 of 45 As documented in Section 1.3, this application involves the reanalysis of the design basis radiological analyses for the following accidents:

Main Steam Line Break (MSLB) Accident

Steam Generator Tube Rupture (SGTR) Accident These analyses utilize the AST guidance documented in RG 1.183 [Reference 1] including modeling the released activities at the maximum allowed by the Technical Specifications.

The Technical Specifications for primary coolant specific activity and secondary activity are reduced by half in this analysis.

The source term is updated to reflect this change. The proposed radiological consequences are compared with the limits provided in 10 CFR 50.67(b)(2), and as clarified per the additional guidance in RG 1.183 for events with a higher probability of occurrence.

Dose calculations are performed at the EAB for the worst 2-hour period, and for the LPZ and Millstone Unit 2 control room for the duration of the accident.

The radiological consequence calculations were performed using the RADTRAD-NAI computer code [Reference 3] as discussed above. The applicable dose acceptance criteria are provided in Table 3.0-1. Table 3.0-1: Accident Dose Acceptance Criteria I Control Room Offsite Accident or Case (EAB and LPZ) Main Steam Line Break Fuel Damage or Pre-Accident Spike 5 rem TEDE 25 rem TEDE Concurrent Iodine Spike 5 rem TEDE 2.5 rem TEDE Steam Generator Tube Rupture Pre-Accident Spike 5 rem TEDE 25 rem TEDE Concurrent Iodine Spike 5 rem TEDE 2.5 rem TEDE

3.1 Main Steam Line Break Accident Serial No.19-243 Docket No. 50-336 Attachment 1, Page 18 of 45 This section describes the methods employed and results of the MSLB design basis radiological analysis.

No fuel failure is expected for the MSLB outside containment case. This analysis includes doses associated with the releases of radioactive material initially present in primary and secondary liquids at maximum aflowable Technical Specification concentrations plus iodine spiking scenarios.

Fuel failure is assumed to be 3.7% for the MSLB inside containment case. Doses were calculated at the EAB for the worst-case two-hour period, LPZ boundary, and Millstone Unit 2 control room. The methodology used to evaluate the control room and offsite doses resulting from the MSLB accident is consistent with RG 1.183 [Reference 1]. 3.1.1 MSLB Scenario Description The MSLB accident begins with a break in one of the main steam lines leading from a SG (denoted as the affected or faulted generator) to the turbine coincident with a loss of offsite power. As a result, the condenser is unavailable and cool down of the primary system is through the release of steam to the environment from the intact SG. In order to maximize doses, break scenarios are assessed in the following structures:

1) Turbine Building (outside containment 6), and 2) Containment.

For the MSLB outside containment, the break occurs downstream of the MSIV and the release continues from the faulted SG for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months until the RCS temperature reaches 212 °F, after which the release stops. The intact SG is used for cooldown via ADVs. The single active failure in this scenario is loss of the MSIV on the faulted SG. The release from the faulted SG is assumed to go directly to the environment because the turbine building blowout panels will have lifted. The primary to secondary leakage of 150 GPO (per TS 3/4.4.5 Bases and TS 6.26), equivalent to 0.87 lbm/min 7 , is assumed to continue from the intact SG for approximately 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months after which decay heat is removed by the shutdown cooling system. Noble gases and iodine progeny are released without holdup in the SGs. No fuel damage is postulated for this scenario.

For the MSLB in containment, the intact SG is used for cooldown via ADVs. The faulted SG releases activity directly to containment for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months until the RCS has reached a temperature of 212 °F. Core damage occurs and the radioactivity released in containment leaks to the environment.

The single active failure credited in this scenario is loss of an emergency diesel 6 Based on turbine building and enclosure building X/Qs (submitted in Reference 9), the turbine building X/Qs are at least four times that of the enclosure building (4.1 times for 0-2 hours). As the steam releases for each scenario will be the same, it is reasonable to model only the turbine building break scenario because the X/Qs are greater and therefore the dose consequences of the turbine building will be greater than the enclosure building.

7 150 x x 3785.41 cm 3 x 1 gm x 0.002205 lbm =O. 87 lbm day 1440 min 1 gal cm3 1 gm mm Serial No.19-243 Docket No. 50-336 Attachment 1, Page 19 of 45 generator.

The primary to secondary leak is assumed to continue from the intact SG for approximately 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months until *shutdown cooling is initiated, after which the ADVs are no longer

required for cooldown and can be closed to terminate the release. Containment leakage is credited for 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (30 days) and no credit for containment sprays is taken. This scenario assumes 3.7% fuel failure due to departure from nucleate boiling (DNB). The RADTRAD-NAI code [Reference 3] is used to calculate the radiological dose consequences from a MSLB at MP2 to the EAB, LPZ, and control room. 3.1.2 MSLB Source Term Definition MSLB Outside Containment The analysis of the MSLB outside containment accident indicates that no fuel failures occur as a result of the transient.

Thus, radioactive material releases during the event are determined by assuming the radionuclide concentrations initially present in primary and secondary liquid are at maximum TS limits plus iodine spiking. In accordance with RG 1.183, Appendix E, two .independent cases are evaluated.

The first case assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike. The MSLB outside containment analysis uses the primary and secondary liquid source term provided in Table 3.1-1. RCS concentrations are based on the proposed TS 3.4.8 limits of 0.5 µCi/gm DE 1-131 and 550 µCi/gm DE Xe-133. Secondary side concentrations are based on the proposed TS 3.7.1.4 limit of 0.05 µCi/gm DE 1-131. For Millstone Unit 2, the maximum iodine concentration allowed by the proposed TS as the . result of an iodine spike is 30 µCi/gm DE 1-131. This value is treated as the pre-accident iodine spike and is listed in Table 3.1-2. RG 1.183 defines a concurrent iodine spike as an accident initiated value 500 times the appearance rate corresponding to the TS limit for normal operation (proposed TS limit: 0.5 µCi/gm DE 1-131) for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The MSLB concurrent iodine spike appearance rates and inventory are listed in Table 3.1-3. The appearance rates developed address the issues raised by NSAL-00-004

[Reference 8].

' Serial No.19-243 Docket No. 50-336 Attachment 1, Page 20 of 45 Table 3.1-1: RCS and Secondary Side Liquid Source Term Primary Secondary Nuclide Concentration Concentration

(µCi/gm) (µCi/gm) 1-131 2.81 E-01 2.96E-02 1-132 1.55E+OO 9.SOE-02 1-133 9.25E-01 9.15E-02 1-134 2.57E+OO 9.15E-02 1-135 I 1.83E+OO 1.56E-01 Kr-85m 2.45E+OO O.OOE+OO Kr-85 5.50E+OO O.OOE+OO Kr-87 2.30E+OO O.OOE+OO Kr-88 4.29E+OO O.OOE+OO Xe-131m 1.06E+01 O.OOE+OO Xe-133m 1.06E+OO O.OOE+OO Xe-133 3.88E+01 O.OOE+OO Xe-135m 2.00E+OO O.OOE+OO Xe-135 1.30E+01 O.OOE+OO Xe-138 1.84E+OO O.OOE+OO Br-84 2.43E-01 1.66E-06 Co-58 5.55E-02 3.22E-06 Co-60 6.35E-03 3.72E-07 Rb-88 2.86E+OO 1.38E-05 Sr-89 1.69E-03 9.65E-08 Sr-90 1.44E-04 8.30E-09 Sr-91 1.30E-02 5.65E-07 Y-91m 6.90E-03 7.90E-08 Y-91 6.25E-05 3.56E-09 Y-93 5.65E-02 2.40E-06 Zr-95 4.69E-03 2.71 E-07 Nb-95 3.37E-03 1.87E-07 Mo-99 7.90E-02 4.40E-06 Tc-99m 6.55E-02 2.32E-06 Ru-103 9.05E-02 5.25E-06 Ru-106 1.08E+OO 6.25E-05 Te-129m '2.29E-03 1.32E-07 Serial No.19-243 Docket No. 50-336 Attachment 1, Page.21 of 45 Table 3.1-1: RCS and Secondary Side Liquid Source Term Primary Secondary Nuclide Concentration Concentration

(µCi/gm) (µCi/gm) Te-129 3.58E-01 5.35E-06 Te-131m 1.91E-02 9.85E-07 Te-131 1.17E-01 7.35E-07 Te-132 2.10E-02 1.16E-06 Cs-134 4.99E-02 5.20E-06 Cs-136 6.30E-03 6.45E-07 Cs-137 6.60E-02 6.90E-06 Ba-140 1.57E-01 8.85E-06 La-140 3.15E-01 1.67E-05 Ce-141 1.81E-03 1.04E-07 Ce-143 3.55E-02 1.82E-06 Ce-144 4.81 E-02 2.71 E-06 Np-239 2.74E-02 1.49E-06 Table 3.1-2: Pre-Accident Iodine Spike Nuclide Iodine Activity in RCS at Iodine Activity in RCS at 0.5 µCi/gm DE 1-131 30 µCi/gm DE 1-131 1-131 2.81 E-01 1.69E+01 1-132 1.55E+OO 9.30E+01 1-133 9.25E-01 5.55E+01 1-134 2.57E+OO 1.55E+02 1-135 1.83E+OO 1.10E+02 Table 3.1-3: Concurrent Iodine Spike Spike= 500, MSLB Concurrent Nuclide Appearance Rate Inventory (Ci)8 (µCi/sec) 1-131 7.15E+05 2.06E+04 1-132 1.74E+07 5.01 E+05 1-133 3.16E+06 9.09E+04 1-134 6.55E+07 1.89E+06 8 No correction made for decay of iodine isotopes.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 22 of 45 Table 3.1-3: Concurrent Iodine Spike Spike = 500, MSLB Concurrent Nuclide Appearance Rate Inventory (Ci)8 (µCi/sec) 1-135 1.01E+07 2.89E+05 MSLB Inside Containment The analysis for the MSLB inside containment assumes 3.7% fuel failure. This source term is used in conjunction with a 1.79 peaking factor and gap fractions as determined in Table 3 of RG 1.183. The core inventory is listed in Table 3.1-4. Table 3.1-4: MP2 Core Inventory Nuclide Curies Nuclide Curies Nuclide Curies Xe-133 1.569E+08 Te-127M 1.287E+06 La-142 1.251 E+08 Xe-135 5.658E+07 Te-129 2.653E+07 Zr-95 1.389E+08 Xe-138 1.316E+08 Te-129M 3.943E+06 Zr-97 1.273E+08 Kr-85 1.194E+06 Te-131M 1.161E+07 Nd-147 4.995E+07 Kr-85M 2.451E+07 Te-132 1.084E+08 Nb-95 1.400E+08 Kr-87 4.860E+07 Sb-127 9.663E+06 Pr-143 1.233E+08 Kr-88 6.865E+07 Sb-129 2.694E+07 Y-90 1.004E+07 1-131 7.719E+07 Ba-139 1.388E+08 Y-91 1.171E+08 1-132 1.105E+08 Ba-140 1.352E+08 Y-92 1.176E+08 1-133 1.504E+08

/ Sr-89 9.426E+07 Y-93 1.300E+08 1-134 1.666E+08 Sr-90 9.627E+06 Cm-242 6.566E+06 1-135 1.407E+08 Sr-91 1.122E+08 Cm-244 1.102E+06 Cs-134 2.821 E+O? Sr-92 1.173E+08 Am-241 1.916E+04 Cs-136 7.545E+06 Ru-103 1.323E+08 Ce-141 1.302E+08 Cs-137 ! 1.319E+07 Ru-105 1.011 E+08 Ce-143 1.232E+08 Cs-138 1.437E+08 Ru-106 6.438E+07 Ce-144 1.046E+08 Rb-86 2.693E+05 Rh-105 9.246E+07 Pu-238 6.162E+05 Rb-88 6.939E+07 Mo-99 1.427E+08 Pu-239 3.656E+04 Rb-89 9.005E+07 Tc-99M 1.250E+08 Pu-240 6.590E+04 Te-127 9.588E+06 La-140 1.359E+08 Pu-241 1.496E+07 La-141 1.275E+08 . Np-239 1.980E+09 3.1.3 MSLB Release Transport Faulted Steam Generator Serial No.19-243 Docket No. 50-336 Attachment 1, Page 23 of 45 Two scenarios are considered for break release pathways (Turbine Building and Containment) as discussed in Section 3.1.1. In both cases, the release of secondary side liquid (SSL) is based on maximum SG volume to maximize isotopic inventory.

The faulted SG models are: 1) Primary to secondary leakage to environment

a. Particulates

& Noble gases + progeny 1 b. Iodine spike activity 9 + progeny 2) Secondary side liquid steaming to environment For the MSLB outside containment, releases are assumed to pass directly from the building to the environment with no credit taken for holdup, partitioning, or scrubbing by the SG liquid. The faulted SG is assumed to steam dry as a result of the MSLB, releasing all of the nuclides in the secondary coolant that were initially contained in the SG. Coincident with the release of SSL, a 150 GPO primary to secondary leakage occurs with the assumption that 100% of the flow flashes and is released to the environment without mitigation.

After 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the RCS will have cooled to below 212 °F and the release via this pathway terminates.

The transport model utilized for noble gases, iodines, and particulates is consistent with Appendix E of RG 1.183. Figures 3.1-1 through 3.1-2 illustrate the release pathways for the outside containment faulted SG cases. Figure 3.1-1: Faulted SG Primary to Secondary Leakage to Environment (Outside Containment)

Primary to Intake T=O hr, 800 cfm Post-isolation:

250 cfm . Secondary Leakage " EN.VIRqNMENJ

..-------'!>!<..

0.87 lbm/min *: * *

Exhaust Flow= Intake MP2 Control 9 Iodine spike activity refers to two cases: 1) pre-accident iodine spike at TS limits and 2) concurrent iodine spike at TS limits. These two cases are modeled separately for the outside containment case.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 24 of 45 Figure 3.1-2: Faulted SG Secondary Side Liquid Leakage to Environment (Outside Containment)

Flow Rates Intake T=O hr, 800 cfm ~-----~ Post-isolation:

250 cfm ENVIRONMENT Exhaust Flow= Intake MP2 Control Room ,35,650 ft: 3 The flow rates for the faulted SG are provided in Table 3.1-5. In order to release the majority of the initial iodine inventory from the faulted SG (as it is assumed to flash directly to the environment), the steam flow at 600 seconds is extended until 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Table 3.1-5: Faulted SG Steam Flow Rates Time (sec) Time (hours) Steam Flow Rate (lbm/min) 0 0.0000 267941.01 10 0.0028 192185.81 20 0.0056 105082.93 50 0.0139 50846.02 100 0.0278 27877.83 200 0.0556 21918.76 400 0.1111 14324.12 600 0.1667 2597.43 18000 5 0.00 2592000 720 0.00 For the MSLB inside containment, faulted SG liquid releases to containment are assumed at the rates in Table 3.1-5 and primary to secondary leakage from the RCS to containment is assumed at 150 GPO. After 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the RCS will have cooled to below 212 °F and the primary to secondary release to containment terminates.

Initially, all of the releases from containment bypass the secondary containment at the TS limit of 0.5% per day. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the leak rates are reduced by 50%, consistent with RG 1.183. Figures 3.1-3 through 3.1-4 illustrate the release pathways for the inside containment faulted SG cases.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 25 of 45 Figure 3.1-3: Faulted SG Primary to Secondary Leakage to Environment (Inside Containment)

Containment Leakage Intake T = O hr, 0.5 % vol/day T = 0 hr, 800 dm Primary to Secondary Leakage __.----T-....=

24 hr, 0.25 % vol/day;,._

~Post-Isolation:

250c _fm ____ 0.87 lbm/min

Containment 1.899E+6 ft3 Exhaust Flow = Intake Room 35,650 ft' Figure 3.1-4: Faulted SG Secondary Side Leakage to Environment (Inside Containment)

Containment Leakage Intake T = 0 hr, 0.5 % vol/day l"-0 hr, 800 cfm Faulted SG Steam Flow Rates ___ I __ = 24 hr, 0.25 % vol/day1-_____ P__,ost-isolation:

250 cfm __ ___ MP2Contro!

Containment

l.';899H6 ti. Intact Steam Generator Two scenarios are considered for break release pathways (Turbine Building and Containment) as discussed in Section 3.1.1. The intact SG models are: 1) Primary to secondary leakage to environment

a. Particulates

& Noble gases + progeny b. Iodine spike activity 10 + progeny 2) Secondary side liquid steaming to environment For the MSLB outside containment, primary to secondary leakage enters the intact SG at the rate of 150 GPO. The activity for the first 20 seconds is released to the environment via the Turbine Building Blowout Panels, resulting in a bounding pathway when the MSIV is open. The activity after 20 seconds is released to the environment via the ADVs, as the X/Qs for ADVs are greater than the MSSVs, resulting in a bounding pathway when the MSIV is closed. There are several nuclide transport models associated with the intact SGs to ensure proper accounting of particulate, iodine and noble gas releases.

Releases of radionuclides initially in the SG liquid and those entering the SG from the primary to secondary leakage flow are 10 Iodine spike activity refers to two cases: 1) pre-accident iodine spike at TS limits and 2) concurrent iodine spike at TS limits. These two cases are modeled separately for the outside containment case.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 26 of 45 released as a result of SSL boiling. The minimum SG mass is used to minimize holdup and retention of activity.

Radionuclides initially in the steam space do not provide any significant dose contribution and are not considered.

Noble gases that are released from the primary system to the intact SG are released to the environment without reduction or mitigation.

The noble gas release pathway after 20 seconds includes a filter for particulates, elemental iodine, and organic iodine, which only allows noble gas isotopes to pass through the environment.

Before 20 seconds, no filter is applied in order for particulates, iodine, and noble gases to be released without any mitigation.

Iodine and particulates after 20 seconds are released via steam flow rates divided by 100 (Table 3.1-7). The PC and MCO are relevant to the dose assessment of the MSLB event in that it indicates how much radioactive inventory is exiting the intact SG. The PC is modeled according to the guidance of RG 1.183 [Reference 1, Section 5.5] in that the iodine PC is 100 (1/100 = 0.01 for entrainment or 1 %). The retention of particulate radionuclides in the intact SG is limited by the MCO from the SGs at 1 %. For the first 20 seconds of a MSLB (prior to the closure of the MSIVs), the intact SG steam release rates are large enough to entrain a significant fraction of water (i.e., steam velocities greater than the 100% power value and beyond the range of available correlations).

Because there is no available basis for estimating water entrainment at these higher initial steam velocities, releases for this time period are conservatively modeled with a MCO of 100% and a PC of 1.0 (no partitioning).

The liquid RCS break flow through the intact SG assumes a reduction factor of 100 for iodine and non-iodine particulate releases due to PC and MCO after 20 seconds. For the MSLB inside containment, the release pathway from the intact SG is identical to that of the MSLB outside containment; however, the source term is different (dependent on 3.7% fuel damage). Figures 3.1-5 through 3.1-6 illustrate the release pathways for the intact SG outside and inside containment cases. Figure 3.1-5: Intact SG Primary to Secondary Leakage to Environment After 20 sec -Noble Gas Pathway 0.87 lbm/mln, 100% filter Before 20 sec -Noble Gas, Iodine, and Particulate Pathway 0.87 lbm/mln, 0% filter, Turbine Building X/Qs ENVIRONMENT Intake T= Ohr, 800 cfm Post-Isolation:

250 cfm Flow Intake Serial No.19-243 Docket No. 50-336 Attachment 1, Page 27 of 45 Figure 3.1-6: Intact SG Secondary Side Liquid Leakage to Environment Before 20 sec Intact SG Steam Flow Rates, -~-Turbine Building X/Qs Intake ~-----~ T=O hr. 800 cfm Post-isolation:

2SO cfm Exhaust Flow = Intake Table 3.1-6 provides the intact SG ADV/MSSV steam flow rates for time between O and 20 seconds using a PC of 1 and a MCO of 100%. Table 3.1-7 provides the intact SG ADV/MSSV steam flow rates for 20 seconds through 16.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months (erid of release) using a PC of 100 and a MCO of 1%. Table 3.1-6: Before 20 Seconds -Intact SG Steam Flow Rates ADV/MSSV Steam Time (sec) Time (hours) Flow Rate (lbm/min) 0 0 214657 2 0.00056 302597 4 0.00111 252861 6 0.00167 214717 8 0.00222 205895 10 0.00278 195773 12 0.00333 183319 14 0.00389 170306 16 0.00444 84640 18 0.005 4140 20 0.00556 0.00 Serial No.19-243 Docket No. 50-336 Attachment 1, Page 28 of 45 Table 3.1-7: After 20 Seconds -Intact SG Steam Flow Rates Time (hours) ADV/MSSV ADV/MSSV Steam Time (sec) Steam Flow Rate Flow Rate +100 averaged (lbm/min) (lbm/min) 20 0.00556 0 O* 800.1 0.222 4375.80 43.76 1650 0.458 4072.86 40.73 2750 0.764 3601.62 36.02 4980* 1.38 3366.00 33.66 6450 1.79 3258.29 32.58 9050 2.51 2812.29 28.12 10000 2.78 2352.50 23.52 12500 3.47 2257.24 22.57 15000 4.17 2157.27 21.57 20000 5.56 2056.29 20.56 25000 6.94 1981.90 19.82 30000 8.33 1923.33 19.23 35000 9.72 1875.20 18.75 40000 11.11 1818.65 18. 19 50000 13.89 1757.73 17.58 60000 16.667 2220.63 0.0

After 20 seconds the MSIV on the intact SG is closed, isolating that portion of the break flow. At 800.1 seconds, releases from the intact SG initiate to allow cooldown and to maximize dose consequences.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 29 of 45 3.1.4 Determination of Atmospheric Dispersion Factors Control room and offsite X/Q values for the MSLB were previously submitted in Reference 9 and approved in Reference

10. Offsite X/Qs for ground level releases from the enclosure building are listed in Table 1.4-1. The enclosure building offsite X/Q values are applied to all MSLB offsite releases due to near proximity of release locations.

Faulted SG releases for the MSLB outside containment could involve a break location in either the turbine building or enclosure building.

The turbine building X/Qs for the control room were used to model releases from the faulted SG and the first 20 seconds of releases from the intact SG as the turbine building X/Qs are larger than corresponding enclosure building X/Qs and are therefore, conservative.

Table 3.1-8 provides the turbine building blowout panels X/Q values. Faulted SG releases for the MSLB inside containment involves a break in containment.

Enclosure building X/Qs for the control room were used to model releases from the faulted SG and the first seconds of releases from the intact SG as the release occurs from the enclosure building and the EBFS is not credited.

The enclosure building X/Q values are listed in Table 3.1-8. Intact SG releases after 20 seconds could occur from either the ADVs or MSSVs. The ADV X/Qs for the control room were used to model releases as the ADV X/Qs are larger than the corresponding MSSV X/Qs and are therefore, conservative.

Table 1.4-1 p*rovides the ADV X/Q values for the control room. Table 3.1-8: MP2 Control Room X/Qs for MSLB Time Turbine Building Enclosure Building (hours) Blowout Panels -ground level (sec/m 3) (sec/m 3) 0-2 1.22E-02 3.00E-03 2-8 8.67E-03 1.87E-03 8-24 3.77E-03 6.64E-04 24-96 2.92E-03 5.83E-04 96-720 2.23E-03 4.97E-04 3.1.5 MSLB Key Analysis Assumptions and Inputs Serial No.19-243 Docket No. 50-336 Attachment 1, Page 30 of 45 The basic data and assumptions are listed below in Table 3.1-9. All numeric values specific to this evaluation are listed in this section. Generic data such as control room information, breathing rates, and X/Q values for offsite locations and ADV releases to the MP2 control room are listed in Table 1.4-1. Table 3.1-8 provides additional MP2 control room X/Q values. Table 3.1-9: Basic Data and Assumptions for MSLB Parameter Value Used Changed from prior license basis? Source Term : ... Primary Coolant Iodine Specific Refer to Table 2.2-1 Yes Activity TS Limits Primary Coolant Noble Gas Refer to Tables 2.2-1 and 3.1-1 Yes Activity Primary Coolant Iodine Refer to Tables 2.2-1 and 3.1-1 Yes Concentrations at TS Limit Pre-Accident Iodine Spike Refer to Tables 2.2-1 and 3.1-2 Yes Concurrent Iodine Spike 500 x Appearance Rate No Concurrent Spike Duration 8 No (hours) Concurrent Iodine Spike (Ci) Refer to Tables 2.2-1 and 3.1-3 Yes Total Accident Induced Primary Refer to Table 1.4-1 No to Secondary Leak Rate Core Inventory for Inside Refer to Table 3.1-4 No Containment Percent Failed Fuel Inside 3.7% No Containment Secondary Iodine Activity Refer to Tables 2.2-1 and 3.1-1 Yes Concentration

MSLB.,PaJc1meters

.... .: . .* . ' . . . .. Loss of Offsite Power Timing Refer to Table 1.4-1 No Iodine Chemical Form of Refer to Table 1.4-1 No Primary to Secondary Leakage MCO in Intact SG Refer to Table 2.2-1 Yes SG Iodine PC in Intact SG Refer to Table 2.2-1 Yes Release Points No Faulted SG Turbine Building / Containment Intact SG ADVs

. Serial No.19-243 Docket No. 50-336 Attachment 1, Page 31 of 45 Table 3.1-9: Basic Data and Assumptions for MSLB Parameter Value Used Changed from prior license basis? Release Duration No Faulteo SG 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Intact SG Approximately 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months Faulted SG Steam Release Refer to Table 2.2-1 Yes Intact SG Steam Release Refer to Table 2.2-1 Yes RCS Volume / Mass Refer to Table 2.2-1 Yes SG Liquid Volume / Mass No Faulted SG 248,891 lbm \ Intact SG 91,092 lbm EAB X/Q Refer to Table 1.4-1 No LPZ X/Qs Refer to Table 1.4-1 No .* . . .' ' Control Room '* *. Control Room X/Qs Refer to Tables 1.4-1 and 3.1-5 No Control Room Ventilation Timing No Outside Containment T= O: Start of MSLB T= 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months : Control room isolation on operator action T = 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months : Control room on filtered recirculation Inside Containment T= 0 : Start of MSLB T = 140 seconds : Control room isolation on Safety Injection Actuation Signal (containment high pressure)

T = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 140 seconds : Control room on filtered recirculation Control Room Emergency 2,250 cfm No Ventilation System Recirculation Flow Rate Refer to Table 1.4-1 Control Room Unfiltered Refer to Table 2.2-1 Yes lnleakage

{cfm) Applicable Time (hours) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (outside containment)

No 0.0389 hours0.0045 days 0.108 hours 6.431878e-4 weeks 1.480145e-4 months (140 seconds, inside containment)

Control Room Normal Intake Refer to Table 1.4-1 No Prior to Isolation Serial No.19-243 Docket No. 50-336 Attachment 1, Page 32 of 45 Table 3.1-9: Basic Data and Assumptions for MSLB Parameter Value Used Changed from prior license basis? Control Room Effective Volume Refer to Table 2.2-1 Yes (ft3) Control Room Emergency Refer to Table 1.4-1 No Ventilation Filter Efficiencies

' ',; . ' ,: . Containment, ', ' ' Containment Volume (ft 3) 1.899E+6 No Containment Leak Rate Refer to Table 2.2-1 Yes Peaking Factor 1.79 Yes Refer to Table 2.2-1 Gap Fractions Group Fraction No 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 3.1.6 MSLB Analysis Results Serial No.19-243 Docket No. 50-336 Attachment 1, Page 33 of 45 The total TEDE to the EAB, LPZ, and control room from a MSLB is summarized below in Tables 3.1-10 and 3.1-11 for the outside and inside containment cases, respectively.

The concurrent iodine spike results in the highest dose consequences for the control room. The MLSB inside containment results in the highest dose consequences for the EAB. The LPZ dose consequences are the same, within round off, for all cases. The doses are within the limits specified in RG 1.183 and 10 CFR 50.67. Table 3.1-10: Dose Summary for the MSLB Accident Outside Containment Location Result Dose Criteria (rem-TEDE) (rem -TEDE) > Concl!rrent fodioe Spike. J ,. *' . / Control Room 3.0 5 EAB 0.1 2.5 LPZ 0. 1 2.5 ' .. 0 : : * Pre-Acdtlent Iodine Spike: ... < ':."* *'. ,,r Control Room 2.2 5 EAB 0.1 25 LPZ 0.1 25 Table 3.1-11: Dose Summary for the MSLB Accident Inside Containment Location Result Dose Criteria (rem-TEDE} (rem -TEDE) ' ,,'. . '". : .. ' Inside. Containment . ,: .. :,, ' .. ' -, -A., ., ' Control Room 1.7 5 EAB 0.2 25 LPZ 0.1 25 3.2 Steam Generator Tube Rupture Accident Serial No.19-243 Docket No. 50-336 Attachment 1, Page 34 of 45 This section describes the methods employed and the results of the SGTR design basis radiological analysis.

This analysis includes dose consequences with the releases of the radioactive material initially present in primary liquid and secondary liquid at the proposed maximum allowable TS concentrations plus iodine spiking (Table 2.2-1). Doses are calculated at the EAB for the worst-case two-hour period, LPZ boundary, and Millstone Unit 2 control room. The methodology used to evaluate the doses resulting from a SGTR is consistent with RG 1.183 [Reference 1 ]. 3.2.1 SGTR Scenario Description A SGTR is a break in a tube carrying primary coolant through the SG. This postulated break allows primary liquid to leak to the secondary side of one of the SGs (denoted as the affected or ruptured SG) with an assumed release to the environment through the ADVs or the MSSVs. The mechanics of the SGTR radiological release consist of releases from the ruptured SG and from*the intact SG. The release from the RCS to the ruptured SG is based on flashed (steam) and unflashed (liquid) RCS break flow. The ADVs/MSSVs on the ruptured SG are assumed to open to control SG pressure at the beginning of the transient.

No credit is taken for release through the condenser due to loss of offsite power. The ruptured SG discharges steam to the environment for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months until the generator is isolated.

Break flow into the ruptured SG continues until the RCS is at a lower pressure.

The MP2 emergency operating procedures recognize that the ruptured SG may need to be steamed following isolation to achieve shutdown cooling conditions.

Depressurization of the SG may be necessary to initiate shutdown cooling. This additional release from the ruptured SG following isolation is modeled early in the event in order to maximize dose consequences.

The intact SG discharges steam for a period of approximately 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months , until the primary system has cooled sufficiently to allow a switchover to the shutdown cooling system. The release from the RCS to the intact SG consists of an assumed primary to secondary leak rate (150 GPO) and is assumed to continue from the intact SG via MSSVs or ADVs until shutdown cooling is initiated.

Consistent with the CLB, no fuel damage is predicted.

Therefore, the SGTR analysis was performed assuming both a pre-accident iodine spike and a concurrent iodine spike. In accordance with RG 1.183, the release of noble gases has been analyzed without reduction or mitigation.

The RADTRAD-NAI code [Reference 3] is used to calculate the radiological dose consequences from airborne release resulting from a SGTR at MP2 to the EAB, LPZ, and control room.

3.2.2 SGTR Source Term Definition Serial No.19-243 Docket No. 50-336 Attachment 1, Page 35 of 45 Initial radionuclide concentrations in the primary and secondary systems for the SGTR accident are determined based on the proposed maximum TS levels of activity.

The SGTR thermal-hydraulic accident analysis indicates that no fuel rod failures occur as a result of this transient.

Thus, radioactive material releases were determined using the radionuclide concentrations initially present in primary liquid, secondary liquid, and iodine spiking .. These values are the starting point for determining the radionuclide inventory input for the NAI code runs. RG 1.183 specifies that the released activities should be modeled as the maximum allowed by the Technical Specification.

Table 3.1-1 lists the primary and secondary coolant radionuclide concentrations that are used in the analysis.

The CLB SGTR source term has been halved to reflect the proposed change to the Millstone Unit 2 Technical Specifications.

RCS concentrations are based on the proposed TS 3.4.8 limits of 0.5 µCi/gm DE 1-131 and 550 µCi/gm DE Xe-133. Secondary side concentrations are based on the proposed TS 3.7.1.4 limit of 0.05 µCi/gm DE 1-131. RG 1.183 also dictates that SGTR accidents consider iodine spiking above the value allowed for normal operations based both on a pre-accident iodine spike and a concurrent iodine spike. For Millstone Unit 2, the maximum iodine concentration allowed by the proposed TS as the result of an iodine spike is 30 µCi/gm DE 1-131. This value is treated as the pre-accident iodine spike and is listed in Table 3.1-2. RG 1.183 defines a concurrent iodine spike as an accident initiated value 335 times the appearance rnte corresponding to the TS limit for normal operation (proposed TS limit: 0.5 µCi/gm DE 1-131) for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The SGTR concurrent iodine spike appearance rates and inventory are listed in Table 3.2-1. Appearance rates developed address the issues raised by NSAL-00-004

[Reference 8].

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 36 of 45 Table 3.2-1: Concurrent Iodine Spike Spike= 335, SGTR Concurrent Nuclide Appearance Rate Inventory (Ci)11 (µCi/sec) 1-131 4.78E+05 1.38E+04 1-132 1.17E+07 3.36E+05 1-133 2.12E+06 6.10E+04 1-134 4.39E+07 1.27E+06 1-135 6.75E+06 1.94E+05 3.2.3 SGTR Release Transport There are several aspects of the SGTR analysis that require multiple RADTRAD-NAI models due to limitations of the code. This is due primarily to treatment of the source terms as noble gases which are released without mitigation or reduction, and iodines and particulates that are released crediting PC and MCO. The different models account for the following scenarios:

1) Break flow and primary to secondary leakage to environment

a. Particulates

& noble gases + progeny at initial RCS TS activity b. Iodine spike activity + progeny i. Pre-accident TS iodine spike + progeny ii. Initial RCS TS iodine plus Concurrent spike+ progeny 2) Secondary side liquid steaming to environment

a. Initial Secondary TS Activity-iodine & particulates+

progeny Ruptured Steam Generator The source term resulting from the radionuclides in the primary system coolant and from the iodine spiking in the primary system is transported to the ruptured SG by the break flow. The ruptured SG is isolated at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and an additional release of 103,200 lbm may be required to meet shutdown cooling system conditions later in the event. Releases from the ruptured SG are terminated when the ruptured SG is isolated and depressurized.

A fraction of the break flow is assumed to flash ,to steam in the ruptured SG and to pass directly into to the environment via the ADVs with no credit taken for scrubbing by the SG liquid. The remainder of the break flow (liquid form) enters the SG liquid. Releases of radionuclides in the SG liquid and those entering the SG from the liquid RCS break flow are released as a result of SSL boiling. 11 No correction made for decay of iodine isotopes.

Serial No.19-243 Docket No. 50-336 Attachment 1, Page 37 of 45 A PC of 100 for iodine isotopes is assumed during boiling and also applied to particulate isotopes.

Thus, 1 % of the elemental iodine, organic iodine, and particulates are released from the SG liquid to the environment along with the steam flow. MCO is not specifically modeled but is bounded by application of the partitioning factor for the duration of the release. Noble gases are released from the primary system to the environment without reduction or mitigation.

The noble gas release pathway for the liquid RCS break flow is modeled with a filter, which only allows noble gas isotopes to pass through to the environment.

The transport model utilized for iodine and particulates is consistent with Appendix E of RG 1.183. The flowcharts in Figures 3.2-1 and 3.2-2 illustrate the release transport of radionuclides from the ruptured SG during a SGTR transient as modeled in RADTRAD-NAI.

Figure 3.2-1: Ruptured SG Break Flow to Environment

*RCS i 423,000 lbrn Liquid RCS Break Flow Rate O%filter RupturedSG liquid Iodine and P~rtieulates . . . 80,000 lbrn, Ruptured SG Steam Flow Rates + 100 Flashed RCS Flow Rate -Noble Gas, Iodine, and Particulate Liquid RCS Break Flow Rate -Noble Gas 100% filter . Environment Intake T = 0 hr, 800 cfm Post-isolation:

250 cfm Flow= Intake MP2Control . Roon, 35,650ft 3 Figure 3.2-2: Ruptured SG Secondary Side Liquid to Environment Ruptured SG liquid Intake T O hr, 800 cfm Post-isolation:

250 cfm Ruptured SG Steam Flow Rates +100 ~----~ 1--------*

Erwironinent Exhaust Flow= Intake MP2 Control *.Room Serial No.19-243 Docket No. 50-336 Attachment 1, Page 38 of 45 Table 3.2-2 below shows the progression of the steam release post-tube rupture for the ruptured SG. Table 3.2-2: Ruptured SG Break Flow Rates and ADV/MSSV Steam Flow Rates Time Event ADV/MSSV Flashed Liquid (hr) Steam Flow Break Flow Break Flow + 100 (lbm/min} (lbm/min) (lbm/min}

Before 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SGTR Starts (0 hr) 31.17 91.67 2,658 Ruptured SG is isolated (1 hr) After 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 12 SG Depressurization 204.8 267.1 11,220 Intact Steam Generator The source term resulting from the radionuclides in the primary system coolant is transported to the intact SG by the accident induced leak rate of 150 GPO (per TS 3/4.4.5 Bases and TS 6.26), equivalent to 0.87 lbm/min 13 as modeled in RADTRAD-NAI.

Radionuclides in the primary coolant leaking into the intact SG are assumed to enter the SG liquid. Releases of radionuclides initially in the SG liquid and those entering the SG from the leakage flow are released as a result of SSL boiling, including an allowance for a PC of 100 for iodine isotopes.

Partitioning is modeled by assuming 1 % of the iodines and particulates pass into the steam space and then directly to the environment.

MCO is not modeled but is bounded by application of the partitioning factor for the duration of the release. Noble gases that are released from the primary system to the intact SG are released to the environment without reduction or mitigation.

The noble gas release pathway includes a filter, which only allows noble gas isotopes to pass through the environment.

Releases were assumed to continue from the intact SG for a period of approximately 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months , after which the shutdown cooling system can remove 100% of decay heat with no requirement for steaming to augment cooldown.

Figure 3.2-3 illustrates the release transport of radionuclides from the RCS during a SGTR transient.

Figure 3.2-4 illustrates the release transport from the bulk boiling of SSL in the intact SG. 12 Release is assumed to terminate at 1.084 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months , 5 minutes after the ruptured SG is isolated at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . There is an operator time critical action to isolate the rup'tured SG following a RCS cooldown to .less than 515 °F within one hour following the reactor trip. 13 150 gal 1 3785.41 cm 3 1 gm 0.002205 lbm _0 87 lbm -X X X-X --day 1440 min 1 gal cm3 1 gm

min Serial No.19-243 Docket No. 50-336 Attachment 1, Page 39 of 45 Figure 3.2-3: Intact SG Primary to Secondary Leakage to Environment Primary to Secondary leakage RCS ,i23,c1001Rm 0.87 lbm/min Intact SG-Uquld 0% filter Iodine and Liquid RCS Break Flow-Noble Gas 0.87 lbm/min 100% filter Intact SG Steam flow Rates + 100 .-------, Intake T = 0 hr, 800 elm Post-isolation:

250 cfm Figure 3.2-4: Intact SG Secondary Side Liquid Leakage to Environment Intact SG. liquid Iodine and Intact SG Steam Flow Rates + 100 Intake T Ohr, 800dm Post-isolation:

250 cfm Exhaust Flow= Intake Table 3.2-3 shows the progression of the steam release post-tube rupture for the intact SG.

Serial No.19-243 Docket No. 50-336 Attachment 1 , Page 40 of 45 Table 3.2-3: Intact SG Primary to Secondary Leakage and ADV/MSSV Steam Flow Rates Time [hr] Total Intact SG Primary to ADV/MSSV Secondary Steam Flow+ Leakage 100 (lbm/min) (lbm/min)

From To 0.00 1.00 2.200E+01 1.00 1.11 8.224E+01 1.11 1.71 5.775E+01 0.87 1.71 2.33 4.712E+01 2.33 2.74 4.308E+01 2.74 3.18 4.275E+01 3.18 3.72 4.241E+01 3.72 6.50 3.078E+01 6.50 17.61 2.413E+01 3.2.4 Determination of Atmospheric Dispersion Factors The EAB, LPZ, and control room X/Q values are presented in Table 1.4-1. The X/Q values used in the SGTR analysis are unchanged from Reference 9 and are listed in Table 1.4-1. 3.2.5 SGTR Key Analysis Assumptions and Inputs The Basic Data and Assumptions are listed below in Table 3.2-4. All numeric values specific to this evaluation are listed in this section. Generic data such as control room information, breathing rates, and X/Q values are listed in Table 1.4-1. Table 3.2-4: Basic Data and Assumptions for SGTR Parameter Value Used Changed from prior license basis? Source Term * . Primary Coolant Iodine Specific Refer to Table 2.2-1 Yes Activity TS Limits Primary Coolant Noble Gas Refer to Tables 2.2-1 and 3.1-1 Yes Activity Serial No.19-243 Docket No. 50-336 Attachment 1, Page 41 of 45 Table 3.2-4: Basic Data and Assumptions for SGTR Parameter Value Used Changed from prior license basis? Primary Coolant Iodine Refer to Tables 2.2-1 and 3.1-1 Yes Concentrations at TS Limit Pre-Accident Iodine Spike Refer to Tables 2.2-1 and 3.1-2 Yes Concurrent Iodine Spike 335 x Appearance Rate No Concurrent Spike Duration 8 No (hours) Concurrent Iodine Spike (Ci) Refer to Tables 2.2-1 and 3.2-1 Yes Total Accident Induced Primary Ref er to Table 1 .4-1 No to Secondary Leak Rate Secondary Iodine Activity Refer to Tables 2.2-1 and 3.1-1 Yes Concentration

.. .. ., . SGTR Parametl;lrS

' * . .. Loss of Offsite Power Timing Ref er to Table 1.4-1 No Iodine Chemical Form of Refer to Table 1.4.-1 No Primary to Secondary Leakage MCO 1% Yes Refer to Table 2.2-1 SG Iodine PC 100 No Scrubbing of Flashed Break Not Credited No Flow Intact SG Tube Uncovery No tube bundle uncovery No assumed Release Duration (hours) No Ruptured SG 1.084 (Refer to Table 3.2-2) Intact SG 17.61 (Refer to Table 3.2-3) Release Points No . Ruptured SG ADVs Intact SG ADVs Ruptured SG Break Flow Rates Refer to Table 2.2-1 Yes Ruptured SG ADV/MSSV Steam Refer to Table 2.2-1 Yes Flow Rates Intact SG ADV/MSSV Steam Refer to Table 2.2-1 Yes Flow Rates Maximum ADV Flow Rate 1.100E+06 No (lbm/hour)

RCS Volume/Mass 423,000 lbm No Serial No.19-243 Docket No. 50-336 Attachment 1, Page 42 of 45 Table 3.2-4: Basic Data and Assumptions for SGTR Parameter Value Used Changed from prior license basis? Initial SG Liquid Volume/Mass 80,000 No (lbm) Initial SG Liquid Volume/Mass, 80,000 Yes for determining isotopic Refer to Table 2.2-1 inventory (lbm) EAB X/Q Refer to Table 1.4-1 No LPZ X/Qs Refer to Table 1.4-1 No Control 'Room '. Control Room X/Qs Refer to Table 1.4-1 No Control Room Ventilation No Timing 14 Pre-Accident Spike T = 0 : Start of SGTR T = 20 seconds : Control room isolation on radiation monitor alarm T = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 20 seconds : Control room on filtered recirculation Concurrent Spike T = 0 : Start of SGTR T = 10 minutes, 20 seconds : Control room isolation on radiation monitor alarm T = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 1 O minutes, 20 seconds : Control room on filtered recirculation Control Room Emergency 2,250 cfm No Ventilation System Recirculation Flow Rate Refer to Table 1.4-1 Control Room Unfiltered Refer to Table 2.2-1 Yes lnleakage (cfm) Control Room Normal Intake Refer to Table 1.4-1 No Flow Rate Prior to Isolation Control Room Effective Volume Refer to Table 2.2-1 Yes (ft3) l Control Room Emergency Refer to Table 1 .4-1 No Ventilation Filter Efficiencies 14 Values are based on control room inlet radiation monitor alarm signal plus 20 seconds to account for damper operation.

3.2.6 SGTR Analysis Results Serial No.19-243 Docket No. 50-336 Attachment 1, Page 43 of 45 The total TEDE to the EAB, LPZ, and control room from a SGTR is summarized below in Table 3.2-5 for the concurrent iodine spike and pre-accident iodine spike. The pre-accident spike results in the highest dose consequences for the control room and EAB. The LPZ is the same, within round off, for the concurrent and pre-accident spike cases. The doses are within the limits specified in RG 1.183 and 10 CFR 50.67. Table 3.2-5: Dose Summary for the SGTR Accident Location Result Dose Criteria (rem -TEDE) (rem -TEDE) '-'*)' J.,. *d Concuirent

/odihe Sp1Ke .l , ., . Control Room 3.1 5 EAB 0.8 2.5 LPZ 0.2 2.5 Pre.,AccidenUPgine Spikf}. . . . . Control Room 3.4 5 EAB 1.0 25 LPZ 0.2 25 4.0 Conclusions Serial No.19-243 Docket No. 50-336 Attachment 1, Page 44 of 45 The proposed changes in Technical Specifications, assumptions, and design inputs have been incorporated into the reanalysis of radiological effects from a Main Steam Line Break and a Steam Generator Tube Rupture accident.

The Technical Specification RCS and secondary side specific activity has been reduced. The analysis results from the reanalyzed events meet the acceptance criteria as specified in 1 O CFR 50.67, GDC-19, SRP-15.0.1, and RG 1.183.

5.0 References Serial No.19-243 Docket No. 50-336 Attachment 1, Page 45 of 45 1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," USNRC, Office of Nuclear Research Regulatory Research, July 2000. 2. 10 CFR 50.67, "Accident Source Term". 3. Software, RADTRAD-NAl Version 1.3 (QA), Numerical Applications, Inc. 4. NUREG/CR-6604, "RADTRAD:

A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," USNRC, June 1997. 5. 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, Environment Protection Agency, September 1988. 6. Federal Guidance Report No. 12, "External Exposures to Radionuclides in Air, Water and Soil," EPA-420-R-93-081, Environmental Protection Agency, September 1993. 7. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," USNRC, Office of Nuclear Regulatory Research, June 2003 8. NSAL-00-004, Westinghouse Nuclear Safety Advisory Letter, "Non-conservatisms in Iodine Spiking Calculations," March 7, 2000. 9. License Basis Document Change Request (LBDCR), 04-MP2-011, "Proposed Technical Specification Changes, Implementation of Alternate Source Term/ June 13, 2006. ADAMS Accession Number: ML061940105

10. NRC Amendment No. 298, "Millstone Power Station, Unit No. 2 -Issuance of Amendment Regarding Alternate Source Term (TAC No. MD2346)," May 31, 2007. ADAMS Accession Number: ML071450053

11. Technical Specifications Task Force, TSTF-490 Rev. 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec," September 13, 2005. 12. NRC Amendment No. 307, "Millstone Power Station, Unit Nos. 2 and 3 -Issuance of Amendment Re: Deletion of E Bar Definition and Revision to the Reactor Coolant System Specific Activity (TAC Nos. MD8492 and MD8493)," October 27, 2008. ADAMS Accession Number: ML082820615

13. NUREG/CR-3304, "Mechanistic Modeling and Correlations for Pool Entrainment Phenomenon," April 1983.

14. 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 19 -Control Room (GDC 19). 15. -NUREG-0800, Chapter 15, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition -Transient and Accident Analysis," SRP-15.0.1, July 2000.

ATTACHMENT 2 Serial No.19-243 Docket No. 50-336 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION AND ENVIRONMENTAL CONSIDERATION DOMINION ENERGY NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 Serial No.19-243 Docket No. 50-336 Attachment 2, Page 1 of 3 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION AND ENVIRONMENTAL CONSIDERATION

1. No Significant Hazards Consideration Determination Dominion Energy Nuclear Connecticut, Inc. (DENG) is submitting a request for an amendment to the Technical Specification (TS) for Millstone Power Station 2 (MPS2). DENG proposes to reduce the TS Reactor Coolant System (RCS) and secondary side specific activity by 50%. This change impacts the primary and secondary liquid source terms. The proposed TS 3.4.8 limits for RCS concentration are 0.5 µCi/gm Dose Equivalent (DE) 1-131 and 550 µCi/gm DE Xe-133, and the maximum iodine concentration allowed by the TS as the result of an iodine spike is 30 µCi/gm DE 1-131. Secondary side concentrations are based on the proposed TS 3.7.1.4 limit of 0.05 µCi/gm DE 1-131. These TS changes are based on evaluations that were conducted to assess the radiological consequences following postulated design basis Main Steam Line Break (MSLB) and Steam Generator Tube Rupture (SGTR) accidents to address analysis deficiencies documented in the corrective action program. A reduction in the TS RCS and secondary side specific activity is required to meet the control room dose regulatory limits. Additionally, the proposed*

reduction in TS RCS and secondary side specific activity will provide inherent source term margin. The analysis results from the reanalyzed events meet the acceptance criteria as specified in 1 Oc.CFR 50.67, 10 CFR 50 Appendix A Generic Design Criterion 19, Standard Review Plan (SRP) 15.0.1, and Regulatory Guide (RG) 1.183. As required by 10 CFR 50.91 (a), DENG has evaluated whether or not a significant hazards corsideration is involved with the proposed amendment by focusing on the three standards set forth on 10 CFR 50.92, "Issuance of amendment," as discussed below: Criterion 1: Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No RCS and secondary side specific activity are not initiators for any accident previously evaluated.

Reanalyzing the MSLB and SGTR events does not require changes to any plant structures, systems, or components (SSCs) and therefore does not affect accident initiators.

As a result, the proposed changes do not significantly increase the probability of an accident.

The proposed TS change will limit primary coolant activity to concentrations consistent with the accident analyses.

The proposed MSLB and SGTR design basis accident analyses demonstrate that the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room doses are within the limits of 10 CFR 50.67, SRP-15.0.1, and RG 1.183. Therefore, Page 1 of 3 Serial No.19-243 Docket No. 50-336 Attachment 2, Page 2 of 3 the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2: Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No The proposed TS change in specific activity limits and the reanalyzed MSLB and SGTR events do not alter any physical part of the plant, (i.e., no new or different type of equipment will be installed), nor do they affect any plant operating parameter or create new accident precursors.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant reduction in a margin of safety? Response:

No The proposed TS change in specific activity limits is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses.

The proposed changes for radiological events related to the computer code used to calculate radiological dose consequences have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10 CFR 50.67, SRP-15.0.1, and RG 1.183. The proposed changes will not result in plant operation in a configuration outside the analyses or design bases and do not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety. Conclusion Based on the above, DENG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Page 2 of 3

2. Environmental Consideration Serial No.19-243 Docket No. 50-336 Attachment 2, Page 3 of 3 Regarding the proposed MSLB and SGTR analyses, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows: (i) The proposed change involves no significant hazards consideration.

As described in Section 1 above, the proposed change involves no significant hazards consideration. (ii) There are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site.

The proposed MSLB and SGTR design basis accident analyses demonstrate that the EAB, LPZ, and Control Room doses are within the limits of 10 CFR 50.67, SRP-15.0.1, and RG 1.183. There are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site. (iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed MSLB and SGTR dose consequences analyses do not implement any plant physical changes, or result in plant operation in a configuration outside the plant safety analyses or design bases. The proposed MSLB and SGTR design basis accident analyses demonstrate that the EAB, LPZ, and Control Room doses are within the limits of 10 CFR 50.67, SRP-15.0.1, and RG 1.183. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure associated with the proposed change. Based on the above, DENC concludes that, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

3. Conclusion DENC concludes, based on the considerations discussed herein, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Page 3 of 3 ATTACHMENT 3 Serial No.19-243 Docket No. 50-336 MARKED-UP TECHNICAL SPECIFICATION PAGES FOR MPS2 DOMINION ENERGY NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 REACTOR COOU.NT SYSTE!\1 SPECIFIC ACffi,1TI' LIMITIN'G CONDIDON FOR OPERATION 3.4_8 The specific activity of the primary coolant shall be limited to: [Q£1----',, a. f-HI 1 1Ciigram DOSE EQD"IVALflH' I-131, and b. .S--+/-100 ft Ci/gram DOSE EQUJVALENT XE-133. 1s501-____-,,~

APPLICABILITY:

MODES 1, 2, 3, 4. ACTION: Serial No.19-243 Docket No. 50-336 Attachment 3, Page 1 of 3 October 27, 2008 a. With the specific activity of the primary coolant> +.-0 ftCi/gram DOSE EQUI\T,UENT I-131, ,.,erify DOSE EQUTu:~ENT I-131 s 60 ft Ci/ gram once per 4hours. 130 I-10.5 I'-), "'---130 I b. With the specific activity of itJe primary coolant > +.-0 fl Ci/ gram DOSE EQUI\l:<U.ENT I-131 but s 60 1 1 Ci/gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months \Vhile efforts are made to restore DOSE EQUIV.<U.ENT I-131 to \,ithin the I0.5 1--3" HI 1 1Ci{gram limit. Specification 3.0.4 is 1 not ar-licable.

0.5 c. With the specific activity of the primary coolant );+,-0 ft Ci/gram DOSE EQUiv:<U.ENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, @QJ-Q;,>.OO ftCi/gramDOSE EQUI\1.:.\1.ENT I-131, be in HOT STAl\'DBYwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0\\1N within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . d. \Vith the specific activity of the prin1ary m-;t.oo r1Ci/gram DOSE EQUI\iAL.E.'IT XE-13 3, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQu"IVA.LTh"T XE-133 to within the -HOO r1Ciigran1 limit. Sp-ecification 3.0.4 is not applicable.

!S; l550f e. With the specific activity of the primary coolant >-HOO ftCi/gram DOSE EQUI\i A.LENT XE-133 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . lvlILLSTONE

-UNTI 2 3/4 4-13 Amendment No. -9, +1+, B-, ill, -!94, m, 3-0-7--

Serial No.19-243 Docket No. 50-336 Attachment 3, Page 2 of 3 Oeteber 29, 2015 REACTOR COOL.\!\'l'f SYSTEM SUR.VEILL.i.NCE REQUIREMENTS 4.4_8.1 4.4.8.2 550 Verify the specific activity of the primary coolant~ -+/-00 µCi/gram DOSE EQUIVALENT XE-13 3 at the frequency specified in the Surveillance Frequency Control Program.* Verify the specific activity of the primary coolant::;* 4-µCi.igram DOSE EQUIVALTh'!'f I-131 at the frequency specified in the Surveillance Frequency Control Program,*

and between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THE!lliii\L POWER change of ;;:: 15% RATED TiiERi.v1AL POWER \\ithin a one hour period.

Surveillance only required to be performed for ~'10DE 1 operation, consistent

\Vith the provisions of Specification 4.0. L -+-~--*** :tv!ILLSTONE

-UNTI 2 3/4 4-14 Amendment No. -B-, w:J., *3***

PL.\.c'IT SYSTEMS ACffi'IIT LIMITING CONDillONFOR OPERATION Serial No.19-243 Docket No. 50-336 Attachment 3, Page 3 of 3 Octobei 29, 2015 [ooJ-**::'!,\

3. 7.1.4 Tire &pecifk activity of the secondary coolant sy:stem shall be~ G:.-+/-G-uCi/gram DOSE EQUrnA..LENTI-131. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: k-0 With the specific activity of the secondary coolant system> {b!.Q uCi/ gram DOSE EQIJiv:>U.Th"T I-131, be in COLD SHUIDO\VN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after detection.

4.7. 1.4 TI1e specific activity of the secondary coolant system shall be det:emi.ined to be within the limit by perl'om1a11ce of the sampling and analysis of Table 4 .Fl. .,,~.,,. }.'fil..LSTONE

-ill\TfI 1 314 7-7 Amendment No.-'B4-*

/

-, Serial No.19-243 Docket No. 50-336 ATTACHMENT 4 MARKED-UP TECHNICAL SPECIFICATION BASES PAGES FOR MPS2 FOR INFORMATION ONLY DOMINION ENERGY NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 REACTOR COOL,\NI' SYSTE.\.f BASES -FOR !NFORMATIOH ONl Y -3i4A.8 SPECIFIC ACTiv1TY (continued)

APPLICABLE SAFETY A.1'\JALYSES Serial No.19-243 Docket No. 50-336 Attachment 4, Page 1 of 3 1'.fan:1!

18, '.'.008 LBDCR O&-MP2-013 The LCO limits on the specific activity of the reactor coolant ensure the resulting offsite and CRE doses meet the appropriate SRP acceptance criteria follo,\ing a SLB or SGTR accident.

The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) rube leakage rate of 150 gpd e..;:ists.

The safety :ui.alyses assume the specific activity of the seconclirycoolant is at its limit ofo+ µCi/ gmDOSEEQUIVALENTI-131 from LCO 3.7.1.4, "Activity." The :ui.alyses for the SLB and SGTR accidents estab]ish the acceptance limits for RCS specific activity.

Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits. / 10.5 f-~ //" The safety analyses cons1d~wo cases of reactor coolant iodine specific activity.

One case ass11m.es specific activity at -1,-0 ~tCi/gm DOSE EQl,'IVALENT I-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cL1dding defects to the prim,,ry coolant immediately after a SLB (by a factor of SOO), or SGTR (by a fac1or of335), respectively.

TI1e second case assumes the initial reactor coolant iodine activity at eQ,G ~tCifgm DOSE EQUIVALENT I-131 due to an iodine spike caused by a reactor or an RCS r, transient prior to the accident.

In both cases, the noble ga5 specific activity is assumed to be-H-00 ftCi/gm DOSE EQliiY:<\LEN'TXE-133.

The SGTR an.-tlysis assumes a rise in pressure in the mpmred SG causes radioactively contantinated steam to discharge to the atmosphere through the atmospheric dump vahres or the 111."lin steam safety valves. TI1e atmospheric disd1arge stops when the turbine bypass to the condenser removes the excess energy to rapidly reduce the RC.S pressure and close the valves. The unaffected SG removes core decay heat by venting steam until !he coo1dmv1l ends and the Slmtdown Cooling (SDC) system is placed u1.service.

The SLB radiological analysis assumes that omite powe.r is lost at the same time as the pipe break occurs outside containment.

The affected SG blows dmvn completely and steam is vented directly to the atmosphere.

The unaffected SG removes core decay he.at by venting steam to the atmosphere until the COO!down end, and the SDI SJ5 1 em is placed in service. _r;-30.0 0. ::, -:-ii_ 'V Operation

with iodine .specific activity levels greater than -l-~1Ci/gm but less than o.r equal to W,G JtCi/gm is permissible for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to r~tore DOSE EQlJiv:4.1.EN'T I-131 to within the-+/-* µCiigmLCO linlit. Operation with iodine specific activity levels greater than~ 11Cilgm is n~permis.sible. '-ill MILLSTOl'U:

-T.JJ\TI 2 B 3/4 4-4a I* Serial No.19-243 Docket No. 50-336 Attachment 4, Page 2 of 3 -FOR !NfORM1-\TlON ONLY-Mareh 18, 2008 -bBDGROS-Ml!J-0.B-REACTOR COOLI\NT SYSTEM BASES 3/4-4.8 SPECmC ACTIVI1Y (coatinued)

APPLICABLE SAFE1Y ANALYSES (CONTINUED)

Toe RCS specific activity limits are also used for establish:ilig standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO };.,-Toe iodine specific activity in the reactor coolant is limited to .hQ. i1Cilgm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to +1-00 ,tCilgm DOSE EQUIVALENT XE-133. Toe limits on specific activity eu.s11re that ofuite ancj\CRE doses will meet the appropriate SRP acceptance criteria (Ref. 2). Toe SLB and SGTR. accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Operation with activities in excess of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2). APPLICAB;ILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQlJlVALEN'T I-131 and DOSE EQUI\Z'liENT XE-133 is necessary to limit the potential consequences of a SLB or SGTR to with:i.11 the SRP acceptance criteria (Ref 2). In J\-10DES 5 and 6, the stean1 generators are not being used for decay heal removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, tl1e monitoring of RCS specific activity is not required.

lvITLLSTONE-UI\TI 2 B 3/44-4b / /

. REACTOR COOL~"T SYSTEi.\*l BASES -FOR lNFORMATlOM otll Y-3/4.4.8 SPECIFIC ACTIVITY (continued)

ACTIONS a. and b. Serial No.19-243 Docket No. 50-336 Attachment 4, Page 3 of 3 }.Jarell.

18, 2008 LBDCR: 08-MP2-013

--With the DOSE EQlJ"I\l.<\I.Th"T 1-131 greater than the LCO limit, Mm.pies at intervals of four hours must be taken to demonstrate that the specific activity is s 60 ~l Ci/gm Four hours is required to obtain and analyze a sample. Sampling is continued every four hours to provide a trend. TI1e DOSE EQUIVA.LE.i'IT I-131 must be restored to within limit \l,ithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there ,vere an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period. A statement in ACTION b. indicates t11e provisions of LCO 3.0.4 are not applicable.

Titls exception to LCO 3.0.4 permits entry into tlle applicable l\*10DE(S), relying on ACTIONS a. and b. wl1ile the DOSE EQUIV.<\LENT 1-131 LCO is not met. This exception is acceptable due to the significant conservatism incorporated into the RCS specific activity limit, the lmv probability of an event w!tlch is limiting due to exceeding thi.;. limi.t, and the ability to re.store transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

C. -,-fil lfthe required action and c~pletion time of ACTION b. is not met, or if the DOSE EQUIV.<\LENT I-131 is >-90 ~tCiigm, the reactor nmst be brought to HOT STAl'{DBY (n,10DE 3) witllin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN (MODE 5) witllin 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed completion times are reasonable, based on operating experience, to reach t11e required plant conditions from full power conditions in an orderly manner and ;,ithout challenging plant S)'5tem5.

MILLSTONE

-UNTI 2 B3/4 4-4c

ML19218A177

Text

1.0 INTRODUCTION

1.1 INTRODUCTION

SUMMARY

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

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