Text

Proposed Technical Specification Changes Implementation of Alternate Source Term
	ML041560464
Person / Time
Site:	Millstone
Issue date:	05/27/2004
From:	Grecheck E; Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	04-285
	Download: ML041560464 (149)
	v • d • e

Dominion Nuclear Connecticut, Inc.

4;iDo *ie Millstone Power Station Rope Ferry Road Waterford, CT 06385 May 27, 2004 U.S. Nuclear Regulatory Commission Serial No.04-285 Attention: Document Control Desk MPS Lic/MAE RO Washington, DC 20555 Docket No.

50-423 License No.

NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests amendments in the form of changes to the Technical Specifications to Facility Operating License Number NPF-49 for Millstone Power Station Unit 3. The proposed changes are being requested based on the radiological dose analysis margins obtained by using an alternate source term consistent with 10 CFR 50.67. A discussion of the proposed Technical Specifications changes is provided in Attachment 1. The marked-up and proposed Technical Specifications pages are provided in Attachments 2 and 3, respectively.

The associated Bases changes are provided in Attachment 6 for information only and will be implemented in accordance with the Technical Specification Bases Control Program and 10 CFR 50.59.

We have evaluated the proposed technical specifications changes and have determined that they do not involve a significant hazards consideration as defined in 10 CFR 50.92.

The basis for our determination that the changes do not involve a significant hazards consideration is provided in Attachment 4. We have also determined that operation with the proposed changes will not result in any significant increase in the amount of effluents that may be released offsite and no significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion 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 changes. The basis for our determination that the changes do not involve any significant increase in effluents or radiation exposure is provided in Attachment 5.

DNC requests approval of the approved change and implementation within 90 days upon issuance of the amendment.

The Site Operations Review Committee and Management Safety Review Committee have reviewed and concurred with the determinations.

Serial No.04-285 Alternative Source Term Page 2 of 4 In accordance with 10 CFR 50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact Mr. Paul R.

Willoughby at (804) 273-3572.

Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services

Serial No.04-285 Alternative Source Term Page 3 of 3 Attachments: (6)

Commitments made in this letter: None.

cc:

U.S. Nuclear Regulatory Commission (w/o Enclosure 1 to Att.

1)

Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider (w/o Enclosure I to Att. 1)

NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring & Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.04-285 Alternative Source Term Page 4 of 4 COMMONWEALTH OF VIRGINIA

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President -

Nuclear Support Services, of Dominion Nuclear Connecticut, Inc.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in 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 '7 Mday qPfoA

, 2004.

My Commission Expires:

3/31 )02 Notary

ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM DISCUSSION OF CHANGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.04-285 Discussion of Changes Page 1 of 98 Index of Contents

1.0 INTRODUCTION

& BACKGROUND...........................................

3

1.1 INTRODUCTION

3 1.2 CURRENT LICENSING BASIS

SUMMARY

4 1.3 ANALYSIS ASSUMPTIONS & KEY PARAMETER VALUES............................................

5 1.3.1 Selection of Events Requiring Reanalysis...........................................

5 1.3.2 Analysis Assumptions & Key Parameter Values...........................................

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

16 2.1 IMPLEMENTATION OF REGULATORY GUIDE 1.183 METHODOLOGYAS DESIGN BASIS SOURCE TERM....16 2.2 RELAXATION OF SURVEILLANCE REQUIREMENTS FOR THE CONTROL ROOM EMERGENCY AIR FILTRATION SYSTEM FILTER EFFICIENCY..................

16 2.3 ELIMINATION OF CREDIT FOR THE CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM 16 2.4 INCREASE IN THE ACCEPTABLE CONTAINMENT LEAKAGE RATES.......................................................

17 2.5 MISCELLANEOUS BASES ONLY CHANGES..................................................................

17 2.6

SUMMARY

OF DESIGN AND LICENSING BASIS CHANGES..................................................................

18 3.0 RADIOLOGICAL EVENT RE-ANALYSES & EVALUATION......................................................... 23 3.1 DESIGN BASIS LOSS OF COOLANT ACCIDENT (LOCA) REANALYSIS.................................................... 25 3.1.1 LOCA Scenario Description.................................................................. 25 3.1.2 LOCA Source Term Definition...........................................................................................26 3.1.3 Determination of Atmospheric Dispersion Factors (X/Q).................................................... 30 3.1.4 Determination of Containment Airborne Activity........................................

32 3.1.5 LOCA Analysis Assumptions & Key Parameter Values............................................

33 3.1.6 LOCA Results.......................

43 3.2 FUEL HANDLING ACCIDENT (FHA)............................

44 3.2.1 FHA Scenario Description...........................

44 3.2.2 FHA Source Term Definition...........................

45 3.2.3 FHA Release Transport...........................

45 3.2.4 Determination of Atmospheric Dispersion Factors (XAQ)...............................................

47 3.2.5 FHA Analysis Assumptions & Key Parameter Values.......................................

........ 47 3.2.6 FHA Analysis Results.......................................................................................................49 3.3 STEAM GENERATOR TUBE RUPTURE ACCIDENT................................................

50 3.3.1 SGTR Scenario Description...............................................

50 3.3.2 SGTR Source Term Definition...............................................

51 3.3.3 Release Transport...............................................

54 3.3.4 Determination of Atmospheric Dispersion Factors........................

....................... 56 3.3.5 SGTR Key Analysis Assumptions and Inputs................

............................... 56 3.3.6 SGTR Analysis Results...............................................

62 3.4 MAIN STEAM LINE BREAK ANALYSIS....................

63 3.4.1 MSLB Scenario Description...............................................

63 3.4.2 MSLB Source Term Definition...............................................

64 3.4.3 Release Transport...............................................

64 3.4.4 Determination of Atmospheric Dispersion Factors............................................................. 66 3.4.4.1 Control Room Atmospheric Dispersion Factors

............................................... 66 3.4.4.2 Offsite Atmospheric Dispersion Factors................................................

66 3.4.5 MSLB Key Analysis Assumptions and Inputs...............

................................ 67 3.4.5.2 MSLB Analysis Results................................................

68 3.5 LOCKED ROTOR ANALYSIS................................................

70 3.5.1 Locked Rotor Scenario Description.70

Serial No.04-285 Discussion of Changes Page 2 of 98 3.5.2 Locked Rotor Source Term Definition................................................................

71 3.5.3 Release Transport................................................................

71 3.5.4 Determination of Atmospheric Dispersion Factors (X'Q).................................................... 71 3.5.5 Locked Rotor Analysis Assumptions and Key Parameters................................................. 72 3.5.6 Locked Rotor Analysis Results................................................................

74 3.6 RCCA EJECTION ACCIDENT ANALYSIS................................................................ 75 3.6.1 RCCA Ejection Accident Scenario Description................................................................ 75 3.6.2 RCCA Ejection Accident Source Term Definition............................................................... 75 3.6.3 Release Transport................................................................

75 3.6.4 Determination of Atmospheric Dispersion Factors (X'Q).................................................... 78 3.6.5 RCCA Ejection Accident Analysis Assumptions and Key Parameters.............

................... 78 3.6.6 RCCA Ejection Accident Analysis Results......................................................................... 79 4.0 ADDITIONAL DESIGN BASIS CONSIDERATIONS................................................................

81 4.1 IMPACT UPON EQUIPMENT ENVIRONMENTALQUALIFICATION............................................................ 81 4.2 RISK IMPACT OF PROPOSED CHANGES ASSOCIATED WITH AST IMPLEMENTATION............

.................. 82 4.3 IMPACT UPON EMERGENCY PLANNING RADIOLOGICAL ASSESSMENT METHODOLOGY...........

.............. 84

5.0 CONCLUSION

S................................................................

85

6.0 REFERENCES

86 7.0 TECHNICAL SPECIFICATION AND BASES CHANGE................................................................ 90 7.1 SPECIFIC TECHNICAL SPECIFICATION CHANGES................................................................

94 7.1.1 Definitions.94 7.1.2 Technical Specification 3/4.7.7, "Control Room Emergency Air Filtration System.95 7.1.3 Bases 3/4.7.7, "Control Room Emergency Air Filtration System".96 7.1.4 Technical Specification 3/4.7.8, "Control Room Envelope Pressurization System.97 7.1.5 Bases 3/4.7.8, "Control Room Envelope Pressurization System.97 7.1.6 Section 6.8.4.f, "Containment Leakage Testing Program.97 7.1.7 Additional Bases Changes........................

98

Serial No.04-285 Discussion of Changes Page 3 of 98 1.0 Introduction & Background 1.1 Introduction This report describes the evaluations conducted to assess the radiological consequences of fully implementing the Regulatory Guide 1.183 (RG 1.183) (Reference 1) accident methodology for Millstone Unit 3. The accident source term discussed in Reference 1 is herein referred to as the Altemative Source Term (AST).

The evaluations documented herein have employed the detailed methodology contained in RG 1.183 for use in design basis accident analyses for alternative source terms. The results have been compared with the acceptance criteria contained either in 10 CFR 50.67 (Reference 2) or the supplemental guidance in RG 1.183.

This application, if granted, would:

Implement RG 1.183 as the design basis source term for Millstone Unit 3,

Allow Millstone Unit 3 to achieve a consistent design basis for all accident dose assessments, Increase operational flexibility by allowing increased Refueling Water Storage Tank (RWST) leakage, Increase operational flexibility by allowing increased unfiltered control room inleakage Increase operational flexibility by allowing increased unfiltered containment leakage and

Remove from the Technical Specifications the Control Room Envelope Pressurization System.

All the radiological dose analyses for the above accidents were performed with a controlled version of the computer code RADTRAD-NAI 1.1 (QA) (Reference 3). The RADTRAD 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 factors. Innovative Technology

Serial No.04-285 Discussion of Changes Page 4 of 98 Solutions 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. The RADTRAD-NAI code is maintained under NAI's QA program, which conforms to the requirements of 1 OCFR50, Appendix B.

Control Room Atmospheric Dispersion Factors were evaluated using the ARCON96 computer code (Reference 5). The ORIGEN and QADS computer codes from the SCALE code package (Reference 6) were used to evaluate the containment and filter shine doses.

1.2 Current Licensing Basis Summary The current design basis radiological analyses that appear in the Millstone Unit 3 Updated Final Safety Analysis Report (UFSAR) consist of assessments of the following events:

1) Main Steam Line Break

2) Locked Rotor Accident

3) Rod Control Cluster Assembly (RCCA) Ejection Accident

4) Small Line LOCA Outside Containment

5) Steam Generator Tube Rupture

6) Loss of Coolant Accident

7) Fuel Handling Accident

8) Waste Gas System Failure

9) Radioactive Liquid Waste System Leak or Failure (Atmospheric Release)

Serial No.04-285 Discussion of Changes Page 5 of 98 1.3 Analysis Assumptions & Key Parameter Values 1.3.1 Selection of Events Requiring Reanalysis In accordance with Standard Review Plan (SRP) 15.0.1, Section 1, Item Number 4, the following radiological analyses have been superceded:

a) Steam System Piping Failures Inside and Outside of Containment b) Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break c) Spectrum of Rod Ejection Accidents d) Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment e) Radiological Consequences of Steam Generator Tube Failure f) Loss-of-Coolant Accidents g) Radiological Consequences of Fuel Handling Accidents A full implementation of the AST (as defined in Section 1.2.1 of Reference 1) is proposed for Millstone Unit 3. To support the licensing and plant operation changes discussed in Section 2.0 of this application, the following accidents were reanalyzed employing the RG 1.183 source term:

Loss of Coolant Accident (LOCA),

Fuel Handling Accident (FHA),

Main Steam Line Break (MSLB) Accident,

Locked Rotor Accident (LRA),

Rod Control Cluster Assembly (RCCA) Ejection Accident (REA) and

Steam Generator Tube Rupture (SGTR) Accident.

The analysis methodology applied the guidance of RG 1.183, in conjunction with the Total Effective Dose Equivalent (TEDE) methodology. If this request is granted, the implementation of RG-1.183 in this plant-specific application will become the

Serial No.04-285 Discussion of Changes Page 6 of 98 bases for the source term employed in design basis radiological analyses for Millstone Unit 3.

The discussion of radiological consequences for "Failure of Small Lines Carrying Primary Coolant Outside Containment" are not required and will be deleted from the FSAR.

The 'Waste Gas System Failure" and "Radioactive Liquid Waste System Leak or Failure (Atmospheric Release)" radiological analyses are being retained for FSAR Chapter 11, Radioactive Waste Management. These events are unaffected by the conversion to AST. The FSAR thyroid and whole body results for these events as shown in Table 15.0-8 will be converted to TEDE so as to employ consistent methodology. Both events result in only an EAB dose.

The TEDE result is:

1) Waste Gas System Failure: 0.22 rem

2) Radioactive Liquid Waste System Leak or Failure (Atmospheric Release): 0.013 rem The proposed licensing and plant operational changes are discussed in Section 2.0. These changes require appropriate changes to the Millstone Unit 3 Technical Specifications, which are also described in Section 2.0 of this report.

The key changes considered are listed below:

a. revise definition of Dose Equivalent 1-131 in Section 1.10 of the Technical Specifications Definitions to reference Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, " 1989, as the source of thyroid dose conversion factors (Reference 8).

b. change Technical Specification 3/4.7.7, "Control Room Emergency Air Filtration System," Surveillance Requirements c.2 and d to reflect a methyl penetration less than or equal to 5% for the Control Room Emergency Air Filtration System filters instead of 2.5%.

Serial No.04-285 Discussion of Changes Page 7 of 98

c. delete Technical Specification 3/4.7.8, "Control Room Envelope Pressurization System."

The Control Room Envelope Pressurization System is no longer credited in the accident analyses described in the evaluation.

d. change the leakage rate acceptance criteria for all penetrations that are Secondary Containment bypass leakage paths in Technical Specification Section 6.8.4.f, "Containment Leakage Testing Program," from s 0.042 La to s 0.06 La.

e. revise the bases Sections 3/4.7.7 and 3/4.7.8 to reflect the above listed changes in accordance with the Millstone Unit 3 Bases Control Program as described in Section 6.18 of the Technical Specifications.

It can be concluded from the evaluation summarized above that implementing the AST, in conjunction with the proposed plant operational changes, will require reanalysis of the LOCA, FHA, SGTR, MSLB, LRA, and REA. Sections 3.1 through 3.6, respectively, provide the detailed description of the re-analyses for these events.

1.3.2 Analysis Assumptions & Key Parameter Values This section describes the general analysis approach and presents analysis assumptions and key parameter values that are common to the accident analyses performed to implement the RG 1.183 source term. Sections 3.1 through 3.6 of this Attachment provide specific assumptions that were employed for the LOCA, FHA, SGTR, MSLB, LRA and REA, 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. The Total Effective Dose Equivalent (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 3 Control Room personnel are calculated for the assumed 30-day duration of the event.

Serial No.04-285 Discussion of Changes Page 8 of 98 The TEDE concept is defined to be the Deep Dose Equivalent, DDE, (from external exposure) plus the Committed Effective Dose Equivalent, CEDE, (from internal exposure). In this manner, TEDE assesses the impact of all relevant nuclides upon all body organs, in contrast with the previous single, critical organ (thyroid) concept for assessing internal exposure.

The DDE 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 DDE in determining the contribution of external dose to the TEDE.

EDE dose conversion factors were taken from Table 111.1 of Federal Guidance Report 12 (Reference 9) per Section 4.1.4 of Reference 1.

There are a number of analysis assumptions and plant features that are used in the analysis of all of the events. These items are presented in Table 1.3-1 through 1.3-5.

Serial No.04-285 Discussion of Changes Page 9 of 98 Table 1.3-1 Control Room Assumptions & Key Parameters Employed in the AST Analyses Assumption / Parameter Value Control Room Effective Volume 2.38E+05 ft3 Normal Control Room Intake Flow Rate prior to Isolation 1595 cfm Unfiltered Inleakage during Periods Of Neutral Pressure 350 cfm Unfiltered Inleakage during Periods Of Positive Pressure 100 cfm Emergency Ventilation System Recirculation Flow Rate 666 cfm Emergency Ventilation System Pressurization Flow Rate 230 cfm Response Time for Control Room Inlet Radiation Monitor 5 seconds to generate the Control Building Isolation (CBI) Signal (Note: this value is validated for each accident analysis in the referenced calculations)

Response Time for Control Room to Isolate upon Receipt 5 seconds of CBI Time credited for delay of Control Room Envelope 1 minute Pressurization System Time credited for Control Room Envelope Pressurization 60 minutes System Discharge to the Control Room Time credited for operator action to align Control Room 40 minutes Emergency Ventilation System after the Control Room Envelope Pressurization System stops Time to place Emergency Ventilation System in service 101 minutes after (summation of the 3 preceding time credits)

CBI signal Filter Efficiencies 90%

elemental aerosol 70% organic Millstone Unit 3 Control Building Wall Thickness:

2 feet concrete Millstone Unit 3 Control Room Ceiling Thickness:

8 inches concrete Millstone Unit 3 Control Building Roof Thickness:

1ft-10in concrete

Serial No.04-285 Discussion of Changes Page 10 of 98 Table 1.3-1 Control Room Assumptions & Key Parameters Employed in the AST Analyses Assumption / Parameter Value Millstone Unit 3 Control Room Occupancy Factors 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 0.6 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 0.4

Serial No.04-285 Discussion of Changes Page 11 of 98 Table 1.3-2 NSS Assumptions & Key Parameters Commonly Employed in the AST Analyses Assumption / Parameter Value Containment Free Volume 2.35E6 ft3 Millstone Unit 3 Containment Wall Thickness:

4.5ft concrete Millstone Unit 3 Containment Dome Thickness:

2.5ft concrete Distance from Millstone Unit 3 Containment to the MP3 228ft Control Room:

Millstone Unit 3 Containment Inner Radius:

70ft

Serial No.04-285 Discussion of Changes Page 12 of 98 Table 1.3-3 Offsite Atmospheric Dispersion Factors (seclm3)

Receptor/ Source Location I Duration X/Q (sec/rn3 )

Exclusion Area Boundary (EAB) (0 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months )

Containment 5.42E-04 Millstone Stack (includes fumigation) 1.OOE-04 Other Release Points 4.30E-04 Low Population Zone (LPZ)

Non-Millstone Stack Release Points 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.91 E-05 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.99E-05 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 8.66E-06 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.63E-06 Millstone Stack (includes fumigation) 0 - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 2.69E-05 4 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.07E-05 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.72E-06 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 2.46E-06 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 5.83E-07

Serial No.04-285 Discussion of Changes Page 13 of 98 Table 1.3-4 Control Room Atmospheric Dispersion Factors Source Location / Duration X/Q (sec/M 3)

Turbine Building Ventilation Vent 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2.82E-03 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.65E-03 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.67E-04 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.83E-04 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.80E-04 Main Steam Valve Building Ventilation Exhaust 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.46E-03 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.76E-04 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.42E-04 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 2.71 E-04 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 1.96E-04 Containment Enclosure Building 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.34E-04 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.23E-04 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.38E-04 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 8.78E-05 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 7.42E-05 Engineering Safety Features Building Ventilation Exhaust 0-2 hour 3.18E-04 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.26E-04 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 9.06E-05 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 6.42E-05 96-720 hour 4.59E-05 Refueling Water Storage Tank Vent 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2.61 E-04 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.59E-04 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.45E-05 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.83E-05 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.63E-05 Millstone Stack 0 -4 hour 1.39E-04 4 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.23E-05 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.56E-05 24 - 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 3.20E-06 96 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.30E-07

Serial No.04-285 Discussion of Changes Page 14 of 98 Table 1.3-4 Control Room Atmospheric Dispersion Factors

Serial No.04-285 Discussion of Changes Page 15 of 98 Table 1.3-5 Breathing Rates Source Location / Duration XIQ (m3lsec)

Offsite (EAB & LPZ) 0 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.50E-04 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.80E-04 24 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.30E-04 Control Room 0 - 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.50E-04

Serial No.04-285 Discussion of Changes Page 16 of 98 2.0 Proposed Licensing Basis Changes This section provides a summary description of the key proposed licensing basis changes that are justified with the Millstone Unit 3 AST analyses accompanying this license amendment request.

2.1 Implementation of Regulatory Guide 1.183 Methodology as Design Basis Source Term This report supports a request to revise the design basis accident source term for Millstone Unit 3. Subsequent to approval of this license amendment, the design basis source term for use in evaluating the consequences of design basis accidents will become the source term documented in RG 1.183 (Reference 1),

including any deviations approved by the NRC staff. This license amendment application is made pursuant to the requirements of CFR § 50.67(b)(1), which specifies that any licensee seeking to revise its current accident source term used in design basis radiological consequences analysis shall apply for a license amendment.

2.2 Relaxation of Surveillance Requirements for the Control Room Emergency Air Filtration System Filter Efficiency The Control Room Emergency Ventilation System is considered operable based on limiting the radiation exposure to personnel occupying the Control Room to the applicable dose. The proposed changes have been analyzed for radiological events with acceptable consequences. The proposed changes meet the criteria as specified in 10CFR50.67 and RG 1.183.

2.3 Elimination of Credit for the Control Room Envelope Pressurization System The Control Room Envelope Pressurization System ensures that a positive pressure is maintained in the control room envelope for any event with the

Serial No.04-285 Discussion of Changes Page 17 of 98 potential for radioactive releases.

The positive pressure limits control room inleakage and consequently dose to the control room occupants so that regulatory dose limits are not exceeded. The proposed change and associated analysis does not credit the system in the calculation of the dose to the control room occupants resulting from the radiological event.

Since the system is not credited and the acceptance criterion is met for the radiological event, the elimination of the Technical Specification is proposed.

Additionally, removal of the Technical Specification is further warranted since it does not meet the criteria in 10CFR50.36 for inclusion into the Technical Specifications.

2.4 Increase in the Acceptable Containment Leakage Rates Containment leakage testing is performed to ensure that the leakage rate shall not exceed the leakage rate values as specified in the Technical Specifications.

The Technical Specification acceptance criteria were developed to result in acceptable dose consequences following a radiological event. The proposed change in the acceptable leakage rate coupled with the analysis of the consequences from radiological events using the core and coolant source term specified by RG 1.183 continue to meet the acceptance criteria of 10CFR50.67 and RG 1.183.

2.5 Miscellaneous Bases Only Changes Bases-only changes will be made primarily to change "10 CFR 100" to "10 CFR 50.67" or "Regulatory Guide 1.183", and to delete phrases such as "well within" and usmall fraction of" that will not have regulatory significance with the AST design basis.

The Bases changes are provided for information only and will be implemented in accordance with 1 OCFR50.59.

Serial No.04-285 Discussion of Changes Page 18 of 98 2.6 Summary of Design and Licensing Basis Changes This section provides a summary of the current design and licensing basis and the proposed changes. The summary is listed in Table 2.6-1.

The existing analyses for the radiological events, as listed in Section 1.2, were performed at various times using different codes and/or hand calculations.

The common element for these events is the assumption of the radiological source term documented in TID-14844 (Reference 7). The proposed amendment utilizes the approach in Regulatory Guide 1.183 and its supporting documents. Additionally, Westinghouse recently updated the thermal-hydraulic analyses supporting the SGTR, MSLB, LRA, and REA radiological calculations.

This accounts for differences in some of the parameters listed in Table 2.6-1.

Table 2.6-1 Summary of Changes to the Design and Licensing Basis For the Radiological Event Analyses Parameter Current Basis I

Proposed Basis Alternate Source Term (Section 3.1)

EAB Dose First 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of accident Worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of accident Offsite Breathing Rates 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.47E-04 3.50E-04 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.75E-04 1.80E-04 24 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.32E-04 2.30E-04 Dose Conversion Factors ICRP30 FGR 11 and 12 RCS and Secondary Side FSAR Table 15.0-10 Table 3.3-1 Technical Specification Activity Pre-accident Iodine Spike FSAR Table 15.0-12 Table 3.3-2 Activities Coincident Spike I FSAR Table 15.0-13 Table 3.3-3 and 3.4-1

Serial No.04-285 Discussion of Changes Page 19 of 98 Table 2.6-1 Summary of Changes to the Design and Licensing Basis For the Radiological Event Analyses Parameter I

Current Basis I

Proposed Basis Millstone Unit 3 Control Room (Section 3.1.5.4)

Unfiltered Inleakage 115 cfm 350 cfm during Periods of Neutral Pressure Unfiltered Inleakage 10 cfm 100 cfm during Periods of Positive Pressure Intake Flow prior to 1450 cfm 1595 cfm Isolation Time to isolate the 5 seconds 10 seconds Control Room Control Room Envelope Credited Not credited (neutral Pressurization System pressure during operation)

X/Q's (unchanged for Murphy & Campe ARCON96 (listed in Millstone stack)

Table 1.3-4)

Control Room Filter Elemental, Aerosol and Elemental & Aerosol:

Efficiencies Organic: 95%

90%

Organic: 70%

Loss-of Coolant Accident (Section 3.1)

Containment Volume 2.32E+06 ft3 2.35E+06 ft3 Sprays Quench spray only at Quench Spray at 72.5 70.2 seconds seconds Recirculation Spray at 14 minutes Spray Coverage Based on quench spray Increases when only Recirculation Spray becomes effective (section 3.1.4.1)

Serial No.04-285 Discussion of Changes Page 20 of 98 Table 2.6-1 Summary of Changes to the Design and Licensing Basis For the Radiological Event Analyses Parameter Current Basis Proposed Basis Mixing Rate Based on Quench Spray Decreases when only Recirculation Spray becomes effective (section 3.1.4.1)

Particulate Iodine Based on Quench Spray Increases when Removal Coefficient only Recirculation Spray becomes effective (section 3.1.5)

Natural Deposition 50% of the lodines

'Powers" model used for plateout aerosol in the unsprayed region Containment Bypass 0.042 La 0.06 La Leak Rate Iodine Chemical Form in 5% Cesium Iodide 95% Cesium Iodide Containment Atmosphere 91% Elemental Iodine 4.85% Elemental Iodine 4% Organic Iodine 0.15% Organic Iodine Iodine Chemical Form in 5% Cesium Iodide 97% Elemental Iodine Containment Sump &

91% Elemental Iodine 3% Organic Iodine RWST 4% Organic Iodine ECCS Leakage Initiation 220 seconds 640 seconds RWST Backleakage Leak rates and RWST Section 3.1.5.3 airflow rate per Amendment 176 Fuel Handling Accident (Section 3.2)

Unfiltered Inleakage 300 cfm 350 cfm during Periods of Neutral Pressure Unfiltered Inleakage 300 cfm 100 cfm during Periods of Positive Pressure

Serial No.04-285 Discussion of Changes Page 21 of 98 Table 2.6-1 Summary of Changes to the Design and Licensing Basis For the Radiological Event Analyses Parameter I

Current Basis I

Proposed Basis Steam Generator Tube Rupture Accident (Section 3.3)

Steam Generator Mass FSAR Table 15.6.3-3 Table 3.3-4 Releases Initial Steam Generator 4.316E+07 grams 4.414E+07 grams Mass RCS Mass 2.359E+08 grams 2.358E+08 grams Release Timing FSAR Table 15.6.3-2 Table 3.3-4 Duration of Primary to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 18 hours Secondary Leakage for Intact Steam Generators Main Steam Line Break Accident (Section 3.4)

Steam Generator Mass FSAR Table 15.1-3 Table 3.4-2 Releases Steam Generator 167,000 Ibm 164,200 Ibm Releases Fuel Defects 0.29%

Technical Specification Limits on RCS Activity Duration of Primary to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 55.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Secondary Leakage for Affected Steam Generator Duration of Primary to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 18 hours Secondary Leakage for Intact Steam Generators Locked Rotor Accident (Section 3.5)

Failed Fuel 16%

7%

Steam Generator Liquid 103,000 Ibm 97,222 Ibm Mass l

l

Serial No.04-285 Discussion of Changes Page 22 of 98 Table 2.6-1 Summary of Changes to the Design and Licensing Basis For the Radiological Event Analyses Parameter Current Basis Proposed Basis RCCA Ejection Accident (Section 3.6)

Containment Volume 2.32E+06 ft3 2.35E+06 ft3 Containment Bypass 0.042 La 0.06 La Leak Rate Iodine Chemical Form in 5% Cesium Iodide 95% Cesium Iodide Containment Atmosphere 91 % Elemental Iodine 4.85% Elemental Iodine 4% Organic Iodine 0.15% Organic Iodine Iodine Chemical Form 5% Cesium Iodide 97% Elemental Iodine released from Steam 91% Elemental Iodine 3% Organic Iodine Generator 4% Organic Iodine Steam Dump 40,604 Ibm 200,000 Ibm Duration of Steam Dump 125 seconds 1,200 seconds Duration of Steam Not considered 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (2 - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months )

Release for Cooldown Time to initiate Safety 1 minute 2 minutes Injection RCS Mass 520,000 Ibm 5.194 E+05 Ibm Steam Generator Liquid 103,000 Ibm 97,222 Ibm Mass Time for Primary System 140 seconds 1,200 seconds Pressure to fall below Secondary System Pressure

Serial No.04-285 Discussion of Changes Page 23 of 98 3.0 Radiological Event Re-analyses & Evaluation As documented in Section 1.3.1, this application involves the reanalysis of the design basis radiological analyses for the following accidents:

Loss-of-Coolant Accident (LOCA)

Fuel Handling Accident (FHA)

Steam Generator Tube Rupture (SGTR) Accident

Main Steam Line Break (MSLB) accident

Locked Rotor Accident (LRA)

Rod Control Cluster Assembly (RCCA) Ejection Accident (REA).

These analyses have incorporated the features of the Alternate Source Term (AST), including the TEDE analysis methodology and modeling of plant systems and equipment operation that influence the events. The calculated radiological consequences are compared with the revised 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 Exclusion Area Boundary (EAB) for the worst 2-hour period, and for the Low Population Zone (LPZ) and Millstone Unit 3 Control Room for the duration of the accident (30 days). Dominion performed all the radiological consequence calculations for the AST with the RADTRAD-NAI and SCALE computer code systems (References 4 and 6) as discussed above.

The dose acceptance criteria that apply for implementing the AST are provided in Table 3.0-1.

Serial No.04-285 Discussion of Changes Page 24 of 98 Table 3.0-1 Accident Dose Acceptance Criteria Accident or Case Control EAB & LPZ Room Design Basis LOCA 5 rem TEDE 25 rem TEDE Steam Generator Tube Rupture Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE Main Steam Line Break Fuel Damage or Pre-accident Spike 5 rem TEDE 25 rem TEDE Coincident Iodine Spike 5 rem TEDE 2.5 rem TEDE Locked Rotor Accident 5 rem TEDE 2.5 rem TEDE RCCA Ejection Accident 5 rem TEDE 6.3 rem TEDE Fuel Handling Accident 5 rem TEDE 6.3 rem TEDE

Serial No.04-285 Discussion of Changes Page 25 of 98 3.1 Design Basis Loss of Coolant Accident (LOCA) Reanalysis This section describes the methods employed and results obtained from the LOCA design basis radiological analysis. The analysis includes dose from several sources. They are:

Containment Leakage Plume,

Emergency Core Cooling System (ECCS) Component Leakage

Refueling Water Storage Tank Vent

Shine from the plume,

Shine from containment and

Shine from the control room filter loading.

Doses are calculated at the Exclusion Area Boundary (EAB) for the worst-case two-hour period, at the Low Population Zone Boundary (LPZ), and in the Millstone Unit 3 Control Room. The methodology used to evaluate the Control Room and offsite doses resulting from a LOCA was consistent with RG 1.183 (Reference 1).

3.1.1 LOCA Scenario Description The design basis LOCA scenario for radiological calculations is initiated assuming a major rupture of the primary reactor coolant system piping. In order to yield radioactive releases of the magnitude specified in RG 1.183, it is also assumed that the ECCS does not provide adequate core cooling, such that significant core melting occurs. This general scenario does not represent any specific accident sequence, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 source term characteristics. Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the severity of incidents evaluated for design basis transient analysis.

Serial No.04-285 Discussion of Changes Page 26 of 98 3.1.2 LOCA Source Term Definition RG 1.183 (Reference

1) provides explicit description of the key AST characteristics recommended for use in design basis radiological analyses.

There are significant differences between the source term in RG 1.183 and the existing design basis source term documented in TID-14844 (Reference 7). The primary differences between the key characteristics of the two source terms are shown in Table 3.1-1 below.

Table 3.1-1 Comparison of TID-14844 and Regulatory Guide 1.183 Source Terms Characteristic TID Source Term RG 1.183 Source Term Noble Gases 100%

Noble Gases 100%

Iodine 50%

Iodine 40%

Core Fractions Released (half of this plates out)

Cesium 30%

To Containment Solids 1%

Tellurium 5%

Barium 2%

Others - 0.02% to 0.25%

Timing of Release Instantaneous Released in Two Phases Over 1.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Interval 91% Inorganic Vapor 4.85% Inorganic Vapor Iodine Chemical and 4% Organic Vapor 0.15% Organic Vapor 5% Aerosol 95% Aerosol Solids Ignored in Analysis Treated as an Aerosol RG-1.1 83 divides the releases from the core into two phases:

1) The Fuel Gap Release Phase during the first 30 minutes and

2) The Early In-vessel Release Phase in the subsequent 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Table 3.1-2 shows the fractions of the total core inventory of various isotope groups that are assumed released in each of the two phases of the LOCA analysis.

Serial No.04-285 Discussion of Changes Page 27 of 98 Table 3.1-2 RG 1.183 Release Phases I Core Release Fractionsa I Isotope Group Gap Early Noble Gasesb 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium 0

0.05 Barium, Strontium 0

0.02 Noble Metals 0

0.0025 Cerium 0

0.0005 Lanthanides 0

0.0002 Duration (hours) 0.5 1.3

a. Release duration apply only to the Containment release. The ECCS leakage portion of the analysis conservatively assumes that the entire core release fraction is in the containment sump from the start of the LOCA.

b. Noble Gases are not scrubbed from the containment atmosphere and therefore are not found in either the sump or ECCS fluid.

The core radionuclide inventory for use in determining source term releases was generated using the ORIGEN code. ORIGEN is part of the SCALE computer code system (Reference 6). Table 3.1-3 lists the 66 isotopes and the associated curies at the end of a fuel cycle that was input to RADTRAD-NAI.

The core inventory used in the LOCA analysis is the identical source term that is used in the selective AST submittal for the Millstone Unit 3 FHA, which was approved by Reference 35. Table 3.1-3 also provides the CEDE and EDE dose conversion factors for each of the isotopes. These dose conversion factors were taken from Federal Guidance Reports 11 and 12 (References 8 and 9, respectively).

Serial No.04-285 Discussion of Changes Page 28 of 98 Table 3.1-3 Core Inventory and Dose Conversion Factors by Isotope Isotope Isotope Group Curies EDE CEDE Sv-m3/Bq-sec Sv/Bq Kr-85 Noble gas 1.075E+06 1.190E-16 O.OOOE+00 Kr-85m Noble gas 2.590E+07 7.480E-15 O.OOOE+00 Kr-87 Noble gas 4.755E+07 4.120E-14 O.OOOE+00 Kr-88 Noble gas 7.060E+07 1.020E-13 O.OOOE+00 Xe-133 Noble gas 1.980E+08 1.560E-15 O.OOOE+00 Xe-135 Noble gas 6.440E+07 1.190E-14 O.OOOE+00 Xe-135m Noble gas 3.589E+07 2.040E-14 O.OOOE+00 Xe-138 Noble gas 8.61 OE+07 5.770E-14 O.OOOE+0O Br-84 Halogen 1.904E+07 9.410E-14 2.270E-11 1-131 Halogen 9.71 OE+07 1.820E-14 8.890E-09 1-132 Halogen 1.416E+08 1.120E-13 1.030E-10 1-133 Halogen 2.008E+08 2.940E-14 1.580E-09 1-134 Halogen 2.146E+08 1.300E-13 3.550E-11 1-135 Halogen 1.864E+08 7.980E-14 3.320E-10 Rb-86 Alkali Metal 2.170E+05 4.810E-15 1.790E-09 Rb-88 Alkali Metal 7.500E+07 3.360E-14 2.260E-1 1 Rb-89 Alkali Metal 6.400E+07 1.060E-1 3 1.160E-11 Cs-134 Alkali Metal 2.037E+07 7.570E-14 1.250E-08 Cs-1 36 Alkali Metal 6.270E+06 1.060E-13 1.980E-09 Cs-137 Alkali Metal 1.256E+07 7.740E-18 8.630E-09 Cs-138 Alkali Metal 1.711 E+08 1.210E-13 2.740E-11 Sb-127 Tellurium 8.81 OE+06 3.330E-14 1.630E-09 Sb-129 Tellurium 3.080E+07 7.140E-14 1.740E-10 Te-127 Tellurium 8.700E+06 2.420E-16 8.600E-11 Te-127m Tellurium 1.463E+06 1.470E-16 5.81 OE-09

Serial No.04-285 Discussion of Changes Page 29 of 98 Table 3.1-3 Core Inventory and Dose Conversion Factors by Isotope Isotope Isotope Group Curies EDE CEDE Sv-m 3/Bq-sec Sv/Bq Te-129 Tellurium 3.013E+07 2.750E-15 2.090E-11 Te-129m Tellurium 6.140E+06 1.550E-15 6.470E-09 Te-131m Tellurium 1.969E+07 7.01OE-14 1.730E-09 Te-132 Tellurium 1.391 E+08 1.030E-14 2.550E-09 Te-1 33m Tellurium 7.620E+07 1.140E-13 1.170E-10 Te-134 Tellurium 1.438E+08 4.240E-14 3.440E-11 Barium-Sr-89 Strontium 1.056E+08 7.730E-17 1.1 20E-08 Barium-Sr-90 Strontium 9.330E+06 7.530E-18 3.51 OE-07 Barium-Sr-91 Strontium 1.276E+08 3.450E-14 4.490E-10 Barium-Sr-92 Strontium 1.278E+08 6.790E-14 2.180E-10 Barium-Ba-139 Strontium 1.722E+08 2.170E-15 4.640E-1 1 Barium-Ba-140 Strontium 1.800E+08 8.580E-15 1.01OE-09 Mo-99 Noble Metal 1.826E+08 7.280E-15 1.070E-09 Rh-1 05 Noble Metal 1.052E+08 3.720E-15 2.580E-10 Ru-1 03 Noble Metal 1.606E+08 2.250E-14 2.420E-09 Ru-1 05 Noble Metal 1.137E+08 3.810E-14 1.230E-10 Ru-1 06 Noble Metal 6.120E+07 O.OOOE+00 1.290E-07 Tc-99m Noble Metal 1.618E+08 5.890E-15 8.800E-12 Ce-141 Cerium 1.657E+08 3.430E-15 2.420E-09 Ce-143 Cerium 1.558E+08 1.290E-14 9.160E-10 Ce-144 Cerium 1.290E+08 8.530E-16 1.01OE-07 Np-239 Cerium 2.080E+09 7.690E-15 6.780E-10

Serial No.04-285 Discussion of Changes Page 30 of 98 Table 3.1-3 Core Inventory and Dose Conversion Factors by Isotope Isotope Isotope Group Curies EDE CEDE Sv-m3/Bq-sec Sv/Bq Pu-238 Cerium 4.083E+05 4.880E-18 7.790E-05 Pu-239 Cerium 3.404E+04 4.240E-18 8.330E-05 Pu-240 Cerium 4.81 OE+04 4.750E-18 8.330E-05 Pu-241 Cerium 1.511 E+07 7.250E-20 1.340E-06 Am-241 Lanthanides 4.520E+03 8.180E-16 1.200E-04 Cm-242 Lanthanides 5.129E+06 5.690E-1 8 4.670E-06 Cm-244 Lanthanides 6.289E+05 4.910E-18 6.700E-05 La-140 Lanthanides 1.864E+08 1.170E-13 1.31 OE-09 La-141 Lanthanides 1.628E+08 2.390E-15 1.570E-10 La-142 Lanthanides 1.551 E+08 1.440E-13 6.840E-11 Nb-95 Lanthanides 1.738E+08 3.740E-14 1.570E-09 Nd-147 Lanthanides 6.590E+07 6.190E-15 1.850E-09 Pr-143 Lanthanides 1.519E+08 2.100E-17 2.190E-09 Y-90 Lanthanides 9.700E+06 1.900E-16 2.280E-09 Y-91 Lanthanides 1.347E+08 2.600E-16 1.320E-08 Y-92 Lanthanides 1.366E+08 1.300E-14 2.11 OE-10 Y-93 Lanthanides 1.01 8E+08 4.800E-15 5.820E-1 0 Zr-95 Lanthanides 1.728E+08 3.600E-14 6.390E-09 Zr-97 Lanthanides 1.587E+08 9.020E-1 5 1.170E-09 3.1.3 Determination of Atmospheric Dispersion Factors (X/Q) 3.1.3.1 Millstone Unit 3 Control Room X/Q The onsite atmospheric dispersion factors were calculated by Dominion using the ARCON96 code (Reference 17) and guidance from Regulatory Guide 1.194 (RG 1.194) (Reference 10). of this Attachment includes an electronic format of site meteorological data taken over the years 1997-2001 and the

Serial No.04-285 Discussion of Changes Page 31 of 98 calculation of onsite X/Q values to the Millstone Unit 3 control room using ARCON96.

The Control Room X/Qs were calculated for the LOCA for the following Millstone Unit 3 source points:

Turbine Building Ventilation Vent

Main Steam Valve Building (MSVB) Ventilation Exhaust

Containment Enclosure Building

Engineered Safeguards (ESF) Building Ventilation Exhaust

Refueling Water Storage Tank (RWST) vent The control room X/Q's from the Millstone stack were not recalculated using ARCON96. The values are consistent with current licensing basis and are based on Regulatory Guide 1.145 (Reference 36) methodology using fumigation conditions.

These values are conservative when compared to the options recommended in RG 1.194 for determination of X/Q values from 'Elevated (Stack) Releases."

The control room XIQ's used in the LOCA analysis are listed in Table 1.3-4.

3.1.3.2 Offsite (EAB & LPZ) XIQ The Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors are part of the existing design basis offsite dose calculations. The X/Q values, which were not revised for the AST analysis, are listed in Table 1.3-3.

These offsite atmospheric dispersion factors were approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002, (References 13 and 14) to Facility Operating License No. NPF-49 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700). The application was dated June 6, 1998 (Reference 19) and supplemented by References 20 through 28.

Serial No.04-285 Discussion of Changes Page 32 of 98 3.1.4 Determination of Containment Airborne Activity 3.1.4.1 Containment Sprays The current licensing basis for the LOCA uses containment sprays to remove elemental and particulate iodine from the containment atmosphere. The use of containment sprays and approval of elemental and particulate iodine removal rates for quench spray were approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002, (References 13 and 14) to Facility Operating License No. NPF-49 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700).

The application was dated June 6, 1998 (Reference 19) and supplemented by References 20 through 28.

The percentage of containment that is covered by quench spray is 49.63%. The Quench Spray system becomes effective at 72.5 seconds. At 14 minutes post-LOCA the Recirculation Spray system becomes effective and the sprayed coverage of containment increases to 64.5% during the time when both spray systems are operating. The mixing rate during spray operation is 2 turnovers of the unsprayed volume per hour.

The elemental and particulate iodine removal rates due to sprays are listed in Table 3.1-6. These spray removal rates are used until the Quench Spray system is secured at 7,480 seconds. At that time further iodine removal is ignored due to sprays even though the Recirculation Spray system remains operating.

An elemental iodine DF of 79 was calculated during the period that sprays are assumed operating. A particulate iodine DF of 49.5 was calculated up to 6,840 seconds, at which time it was reduced per Table 3.1-6.

Serial No.04-285 Discussion of Changes Page 33 of 98 3.4.1.2 Natural Deposition A reduction in airborne radioactivity in the containment by natural deposition within containment was credited. The model used is described in NUREG/CR-6189 (Reference 12) and is incorporated into the RADTRAD computer code.

This model is called the Powers model and it's used for aerosols in the unsprayed region and set for the 10th percentile.

3.1.5 LOCA Analysis Assumptions & Key Parameter Values 3.1.5.1 Method of Analysis The RADTRAD-NAI code (Reference 3) is used to calculate the radiological consequences from airborne releases resulting from a LOCA at Millstone Unit 3 to the EAB, LPZ, and Millstone Unit 3 Control Room. The ORIGEN code is used to determine the grams of iodine in the core for calculating RWST backleakage.

The QADS code was used to calculate the shine dose to the Control Room from containment shine and control room filter shine.

This analysis addresses a plant specific issue of unfiltered post-LOCA releases due to damper bypass and duct leakage from the plant ventilation system that was described and approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002 (References 13 & 14), to Facility Operating License No.

NPF-49 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700). The application was dated June 6,1998 (Reference 19) and supplemented by References 20 through 28.

Amendment 211 identified potential release pathways from the secondary containment to the environment that could bypass the SLCRS filter following a design-basis accident due to non-nuclear safety grade (NNS) exhaust fan operation after the accident. Amendment 211 also approved an operator action

Serial No.04-285 Discussion of Changes Page 34 of 98 that would manually trip the breakers on selected fans at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 20 minutes post-LOCA. This operator action is only credited in the control room habitability analysis.

This licensing basis is further described in section 15.6.5.4, Radiological Consequences of a LOCA, in the Millstone Unit 3 FSAR. The AST analysis does not change the licensing basis for the post-accident operation of SLCRS as described and approved in Amendment 211.

3.1.5.2 Basic Data & Assumptions for LOCA Table 3.1-4 Basic Data and Assumptions for LOCA Parameter or Assumption /

Value (Reference)

Containment Leak Rate: (Technical 0.3% by weight of the containment air Specifications) per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (La)

Containment Bypass Leak Rate:

0.06La (Technical Specifications &

Conservative Assumption)

Containment leak rate Reduction:

50% after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (offsite)

(Reference 1) 50% after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (control room)

Secondary Containment Drawdown 2 minutes Time: (Technical Specifications)

Iodine Chemical Form in Containment 95% Cesium Iodide Atmosphere: (Reference 1) 4.85% Elemental Iodine 0.15% Organic Iodine Iodine Chemical Form in the Sump and 97% Elemental RWST: (Reference 1) 3% Organic Containment Sump pH:

at least 7 Dose Conversion Factors:

References 8 and 9 SLRCS Filter Efficiency: (Technical 95% all lodines and Particulates Specifications & Conservative Assumption)

Serial No.04-285 Discussion of Changes Page 35 of 98 Table 3.1-4 Basic Data and Assumptions for LOCA Parameter or Assumption /

Value (Reference)

Auxiliary Building Filter Efficiency:

95% all lodines and Particulates (Technical Specifications &

Conservative Assumption)

Quench Spray System Effective Period 72.5 - 7,480 seconds of Operation:

Recirculation Spray System Start Time:

660 +/- 20 seconds (Technical Specifications)

Recirculation Spray System 14 minutes Effectiveness Time:

Elemental Iodine Removal Coefficient:

20 per hour Particulate Iodine Removal Coefficient

DF < 50: 12.73 for Quench Spray:

DF > 50: 1.27 Particulate Iodine Removal Coefficient

DF < 50: 16.14 for Quench and Recirculation Spray:

DF > 50: 1.61 Quench Spray Volume of Containment:

1,166,200 ft3 Quench and Recirculation Spray 1,515,858 ft3 Volume:

ECCS System Leakage Outside 4,730 cc/hr Containment of Containment:

Minimum Available RWST Volume:

1,072,886 gallons Minimum Quench Spray System Auto 47,652 gallons Trip Value:

RWST Maximum Fill Volume:

1,206,644 gallons 3.1.5.1 Containment Leakage Model The containment leakage normally consists of filtered and bypass leakage. As stated in the data and assumptions, the total containment leak rate (La) is 0.3%

per day. The bypass leak rate is assumed to be 0.06

La or 0.018% per day

Serial No.04-285 Discussion of Changes Page 36 of 98 after SLCRS drawdown time. The bypass leak rate bypasses the secondary containment and is released unfiltered at ground level directly from containment.

The entire containment leak rate bypasses the secondary containment until the SLCRS drawdown time of 2 minutes (Reference 1). The leak rate is reduced by one-half (0.009% per day) at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for offsite calculations and at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for control room calculations.

The reduction in the containment leak rate by 50% at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the control room analysis was approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002 (References 13 and 14), to Facility Operating License No.

NPF-49 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700). The application was dated June 6, 1998 (Reference 19) and supplemented by References 20 through 28.

This reduction in containment leakage is based on the fact that the Millstone Unit 3 containment pressure is rapidly reduced compared to typical PWR's because of its original design as a negative pressure containment.

The collection, processing, and release of containment leakage vary depending on the location of the leak. Ventilation characteristics and release paths are different for each building comprising the secondary containment. Tables 15.6-9 and 15.6-12 of the Millstone Unit 3 FSAR describe the ventilation characteristics and release paths.

3.1.5.2 Model of ECCS Leakage The Emergency Core Cooling System (ECCS) fluid consists of the contaminated water in the sump of the containment. This water contains 40% of the core inventory of iodine, 5% released to the sump water during the gap release phase (30 minutes) and 35% released to the sump water during the early in-vessel phase during the next 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . During a LOCA the highly radioactive ECCS fluid is pumped from the containment sump to the recirculation spray headers

Serial No.04-285 Discussion of Changes Page 37 of 98 and sprayed back into the containment sump.

Also, following a design basis LOCA, valve realignment occurs to switch the suction water source for the ECCS from RWST to the containment sump.

ECCS leakage develops when ESF systems circulate sump water outside containment and leaks develop through packing glands, pump shaft seals and flanged connections.

The Technical Specification 6.8.4a, Primary Coolant Sources Outside Containment Program Manual calculates this leakage at 4,780 cc/hr. In accordance with Reference 1, the ECCS analysis makes use of 10,000 cc/hr for ECCS leakage. The leakage of recirculating sump fluids commences at 640 seconds, which is the earliest time of recirculation.

The temperature of the containment sump is conservatively assumed to reach a maximum of 240 degrees F. At this maximum temperature, a flash fraction of 0.03 is calculated. However, per the guidance of RG-1.183, a conservative flash fraction of 0.1 was used for the ECCS leakage during the entire event. The water volume of the sump at 640 seconds is 1.068E+05 gallons and increases to 4.074E+05 gallons at 0.8217 hours0.0951 days 2.283 hours 0.0136 weeks 0.00313 months , where it is assumed to remain constant.

3.1.5.3 Model of ECCS Back Leakage to Refueling Water Storage Tank Following a design basis LOCA, valve realignment occurs to switch the suction water source for the ECCS from the Refueling Water Storage Tank (RWST) to the containment sump.

In this configuration, Motor Operated Valves (MOVs) and check valves in the normal suction line from the RWST and MOVs in the recirculation line provide isolation between this contaminated flow stream and the RWST.

RADTRAD-NAI is used to model leakage of ECCS fluid through these valves back into the RWST with subsequent leakage of the evolved iodine through the vent at the top of the RWST to the environment.

Serial No.04-285 Discussion of Changes Page 38 of 98 The RADTRAD-NAI source term used to model the ECCS leakage into the RWST contains only the iodine isotopes. Forty percent of the core inventory of iodine isotopes was modeled as being instantaneously transported from the core to the containment sump. The iodine form is 97% elemental and 3% organic in accordance with RG-1.183.

Serial No.04-285 Discussion of Changes Page 39 of 98 New times, new flow rates and new contaminated volumes discharged to the RWST from assuming higher leak rates in the RWST backleakage paths have been calculated. The leak paths back to the RWST are:

CHS Alternate Recirculation Leakage

RHR Leakage through V*43

SIH Pump Recirculation

RHRS A and B suction

CHS Suction

SIH Suction The leakage paths and methodology to calculate times, flow rates and volumes were approved in Amendment 176 (Reference 29) to Facility Operating License No. NPF-49 for Millstone Unit 3, in response to the application dated May 7, 1198, (Reference 30) as supplemented January 22, 1999 (Reference 31) regarding RWST backleakage (TAC No. MA1749).

Table 3.1-5 summarizes the results of the above leakages for the 7 sources of backleakage to the RWST.

Table 3.1-5 Contaminated Inflow to RWST Source Time, hours Flow Rate, gpm CHS Suction 137.66 0.20 RHRS A Suction 126.25 0.20 RHRS B Suction 144.43 0.20 SIH Recirculation 8.50 0.20 RHRS Recirculation 29.83 0.60 CHS Recirculation 36.93 0.20 SIH Suction 67.48 0.20

Serial No.04-285 Discussion of Changes Page 40 of 98 Using the methodology approved in Amendment 176, the time for contaminated sump water to reach the RWST is based on the calculated flow rates and the volume of clean water in the associated piping. The time required to displace the clean volume is reduced by 50% to account for mixing in the lines. This is considered a reasonable assumption since the sump fluid is relatively cool and thermal mixing will be minimal. In addition, the lines are isolated and stagnant except for minor leakage rates and the mixing due to flow is negligible. Table 3.1-6 reduces the times in Table 3.1-5 by 50%, integrates the flow rates over time and calculates the total contaminated volume discharged to the RWST over the 30 day LOCA period.

Table 3.1-6 Summary of Times, Integrated Flow Rates & Volumes for RWST Backleakage Time (hrs)

Flow Rate Volume (ft3)

(cfm) 4.25 0.03 0.00 SIS R 14.91 0.11 17.11 RHS R 18.46 0.13 39.88 CHS R 33.74 0.16 162.41 SIH S 63.13 0.19 445.21 RHRS A 68.83 0.21 509.27 CHS S 72.21 0.24 552.68 RHR B 720.0 0.24 9,904.49

@ 30 days For the analysis of the partition coefficient, the amount of water remaining in the RWST at the end of the injection phase is conservatively taken to correspond to the lowest possible value; the minimum QSS auto trip value or 47,652 gallons.

The RWST airflow rate of 8.7 cfm was determined by making use of the ideal gas law and expected volumetric change. The latter was based on a conservative rise in air temperature within the RWST as a result of solar heating. The air

Serial No.04-285 Discussion of Changes Page 41 of 98 released from the RWST will be free of radioactivity until the backleakage reaches the RWST at 4.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months post-LOCA.

The partition coefficient (PC) applicable to the iodines in the RWST water is based upon information in Reference 15. For this application, the RWST was assumed to behave like a closed system for the establishment of equilibrium conditions between the water and air.

ORIGEN was used to calculate the quantity of grams of iodine in the core at 20,000 grams. The fraction of iodine released during the LOCA is 0.4, resulting in the grams of iodine in the sump at 8.OE+03. The volume of liquid in the sump is the sum of 1,160,776 gallons, resulting in an iodine concentration in the sump of 7 mgrams / gallon. Total volume transferred to the RWST over the 30 days as a result of backleakage is 7.41 E+04 gallons resulting in a total of 5.187E+05 mgrams of iodine transferred to the RWST.

The maximum concentration of iodines in the RWST is 4.3 mgrams/gallon or 1.2 mgrams/liter. The PC of 4,000 corresponds to an iodine concentration of 1.2 mgrams/liter, taken from Reference 15.

A PC of 4,000 results in a DF of 450. A DF of 100 was used for conservatism.

3.1.5.4 Millstone Unit 3 Control Room The control room volume is 2.38E5 ft3. The LOCA causes a Control Building Isolation (CBI) signal to isolate the control room (current Technical Specifications). The control building is isolated within 5 seconds after a CBI signal. According to Reference 1, the onset of the gap release does not start until 30 seconds post-LOCA. Therefore the control room will be isolated prior to the arrival of the radioactive release.

The following is taken from the current Technical Specifications.

"After a 60 second time delay the control room envelope pressurizes to greater than or equal to 1/8-inch water gauge relative to adjacent areas and the outside

Serial No.04-285 Discussion of Changes Page 42 of 98 atmosphere. The positive pressure is maintained for greater than or equal to 60 minutes."

The timing and operation of the Control Room Envelope Pressurization System and Control Room Emergency Ventilation System is described in Section 9.4 of the Millstone Unit 3 FSAR. The timing and operation of these systems were approved in Amendment No. 211, dated September 16, 2002 and November 25, 2002, to Facility Operating License No. NPF-49 for Millstone Unit 3 regarding the revised Final Safety Analysis Report licensing basis for post-accident operation of the Supplementary Leakage Collection and Release System (TAC No.

MB3700). The application was dated June 6, 1998 and supplemented by letters dated April 5, 1999; April 7, April 19, July 31, and September 28, 2000; March 19, June 11, September21, and December 20, 2001.

In the LOCA analyses the Control Room Envelope Pressurization System is not credited with operating and providing a positive pressure in the control room.

Therefore, during the one-hour period that the Control Room Envelope Pressurization System should be operating, the control room is assumed to be at a neutral pressure. During periods of neutral pressure in the Millstone Unit 3 control room, unfiltered inleakage is assumed to be at the analysis limit of 350 cfm.

During periods of positive pressure in the Millstone Unit 3 control room unfiltered inleakage is assumed to be at the analysis limit of 100 cfm.

The Control Room Emergency Ventilation System filter efficiencies are conservatively assumed at 90% for both elemental and aerosol and 70% for organic iodines.

Serial No.04-285 Discussion of Changes Page 43 of 98 The post LOCA dose consequences to the Millstone Unit 3 control room are due to the following sources:

1. 1. airborne contribution

containment leakage

ESF leakage

RWST backflow

2. external sources

control room filter shine

cloud shine RWST direct shine

containment direct shine The doses due to external sources were calculated using data from Tables 1.3-1, 1.3-2, and Section 3.1.5 3.1.6 LOCA Results Table 3.1-7 lists TEDE to the EAB and LPZ from a LOCA at Millstone Unit 3.

The dose to the EAB and LPZ is less than the 25 rem TEDE limit stated in 10CFR50.67 and Regulatory Guide 1.183. The EAB dose represents the worst 2-hour dose for each release pathway.

The dose to the Millstone Unit 3 control room is less than the 5 rem TEDE limit specified in in 10CFR50.67 and Regulatory Guide 1.183.

Table 3.1-7 TEDE from a Millstone Unit 3 LOCA Location TEDE (rem)

EAB 9.1 E+00 LPZ 4.5E+00 Millstone Unit 3 Control Room 3.4E+00

Serial No.04-285 Discussion of Changes Page 44 of 98 3.2 Fuel Handling Accident (FHA)

This section describes the methods employed and results of the Fuel Handling Accident (FHA) design basis radiological analysis. The analysis includes doses associated with release of gap activity from a fuel assembly either inside containment or in the Fuel Building. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ) boundary, and in the Unit 3 control room. The methodology used to evaluate the control room and offsite doses resulting from the FHA is consistent with RG 1.183 in conjunction with TEDE radiological units and limits, ARCON96 based onsite atmospheric dispersion factors, and Federal Guidance Reports No. 11 and 12 dose conversion factors.

The FHA recently approved in Amendment 219 (Reference 35) for selective implementation of the AST differs from the FHA in this amendment request by the following:

1) Revised Control Room X/Qs (now based on ARCON96),

2) Changes were made to Control Room Inleakage assumptions (larger inleakage rates),

3) Reduced Control Room filtration efficiency (from 95% to 90/90/70%

particulate/ elemental/ organic) and

4) No credit for the Control Room Envelope Pressurization System.

3.2.1 FHA Scenario Description The design basis scenario for the radiological analysis of the FHA assumes that cladding damage has occurred to all of the fuel rods in one fuel assembly plus 50 rods in the struck assembly. This scenario is unchanged from the assumption in the existing UFSAR analysis. The rods are assumed to instantaneously release their fission gas contents to the water surrounding the fuel assemblies. The analyses include the evaluation of FHA cases that occur in both the containment

Serial No.04-285 Discussion of Changes Page 45 of 98 and the Fuel Building. Essentially all radioactivity released from the damaged fuel is assumed to release over a two hour period through an open penetration in the containment or the Fuel Building.

3.2.2 FHA Source Term Definition In accordance with Regulatory Position 3 of RG 1.183 the core source was determined using ORIGEN to evaluate multiple cycle designs (based on the Dominion fuel management scheme for enrichment and burnup).

This core inventory was used and approved in the FHA selective implementation of the AST in Amendment 219 and is described in the LOCA scenario (Section 3.1.2) and is used for the FHA with a 100-hour decay time.

For the FHA analyses, the core inventory was used to calculate the gap activity of one fuel assembly plus 50 rods for input to RADTRAD-NAI. The amount of fuel damage is the same whether the FHA is in the Fuel Building or Containment.

Therefore, the only variable between FHA in the Containment or in the Fuel Building is the release point. Consistent with Amendment 219 for the selective implementation of the AST, the FHA analyses in this amendment request assume the resulting chemical form of the radioiodine in the water is 99.85%

elemental iodine and 0.15% organic iodide.

3.2.3 FHA Release Transport This evaluation does not credit operability or operation of the Containment purge system, Auxiliary Building or Fuel Building ventilation. This evaluation assumes that the personnel hatch, equipment hatch and penetrations are open for the duration of the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months release.

Releases from the Fuel Building or Containment to the environment are at a rate of 3.454 air changes per hour. This assures that greater than 99.9% of the activity in the Fuel Building and Containment analyses were released within 2

Serial No.04-285 Discussion of Changes Page 46 of 98 hours0.00113 days 0.0272 hours 1.62037e-4 weeks 3.7289e-5 months . The release rate is conservative in that it biases the bulk of the release (i.e., > 80%) to occur within the first half hour of the event. No credit is taken for filtration of the release from either the Fuel Building or Containment. Additionally, no credit is taken for dilution or mixing of the activity released to the Fuel Building or Containment air volumes.

If the Containment is not isolated, the release from a FHA inside the Containment could be from:

1) The equipment hatch or penetrations via the Enclosure Building,

2) The personnel hatch or penetrations to the auxiliary building and discharged out the Turbine Building Ventilation Stack,

3) Leakage discharged through purge to the Turbine Building Ventilation Stack.

For a FHA in the Fuel Building, the release is from either the Turbine Building Ventilation Stack or the Fuel Building roll-up doors.

The most conservative release point to the Control Room is an unfiltered release from the Turbine Building Ventilation Stack. For conservatism in calculating off-site doses, the release is discharged unfiltered out the Equipment Hatch via the Containment Enclosure Building.

This is because the EAB Containment Enclosure Building X/Q is higher than for the Ventilation Stack release. This modeling is consistent with Amendment 219, approved, March 17, 2004.

Serial No.04-285 Discussion of Changes Page 47 of 98 3.2.4 Determination of Atmospheric Dispersion Factors (X/Q) 3.2.4.1 Control Room Atmospheric Dispersion Factors The onsite atmospheric dispersion factors were calculated by Dominion using the ARCON96 code and guidance from RG-1.194 [Reference 10] for the Control Rooms. Site meteorological data taken over the years 1997-2001 were used in the evaluations.

Control room X/Q values were calculated at the Turbine Building Ventilation Stack, and Containment Enclosure Building. The control room atmospheric dispersion factors are presented in Table 1.3 4.

3.2.4.2 Offsite Atmospheric Dispersion Factors (X/Q)

The offsite atmospheric dispersion factors (EAB and LPZ) used for the FHA analysis are the same as those used for LOCA. They are reported in Table 1.3-3.

3.2.5 FHA Analysis Assumptions & Key Parameter Values The basic data and assumptions are listed below in Table 3.2-1.

Table 3.2-1 Data and Assumptions for the Fuel Handling Accident Analysis Data / Assumption Value Gap Fractions Noble Gases: 10%

Halogens:

8%

Pool Decontamination Factor:

Noble Gases: 1 lodines: 200 (effective DF)

Release Point:

Turbine Building Ventilation Stack or Enclosure Building / Containment Ground Decay Time:

100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Radial Peaking Factor:

1.7 Duration of Release to the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Environment:

Serial No.04-285 Discussion of Changes Page 48 of 98 Table 3.2-1 Data and Assumptions for the Fuel Handling Accident Analysis Fuel Damage:

l1 assembly plus 50 rods out of a total of 193 lassemblies in a core Plume and Filter Shine are accounted in the Control Room doses.

Control Room Ventilation Timing

T= 0 seconds Accident Initiation Unfiltered Intake Flow -1595 cfm

T= 5 seconds Control Building Isolation (CBI) Signal generated Control Room Envelope Pressurization System receives CBI signal

T= 10 seconds Control Room Isolates on Radiation Monitor Signal Intake Flow- 0 cfm Unfiltered Inleakage Flow -350 cfm

T= 1 minute, 5 seconds Assumes 1 minute delay for Control Room Envelope Pressurization System Bottles response time No credit is taken for operability of Control Room Envelope Pressurization System Unfiltered Inleakage Flow -350 cfm

T= 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 41 minutes, 5 seconds (= 1.685 hours0.00793 days 0.19 hours 0.00113 weeks 2.606425e-4 months )

Assume 40 minute delay for operator action Control Room is pressurized Filtered Intake Flow - 230 cfm Unfiltered Inleakage Flow-100 cfm Filtered Recirculation Flow - 666 cfm

T= 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months

Filtered Recirculation Flow - 666 cfm

Serial No.04-285 Discussion of Changes Page 49 of 98 3.2.6 FHA Analysis Results The offsite and control room doses are listed below. The Millstone Unit 3 Fuel Handling Accident assumes a two-hour release without building integrity for the Containment FHA and without Fuel Building Filtration for the Spent Fuel Pool FHA. The associated worst case TEDE is presented in Table 3.2-2. All doses are less than the limits specified in Regulatory Guide 1.183 and 10 CFR 50.67.

Table 3.2-2 Dose Summary for the Fuel Handling Accident Analysis Location TEDE (rem)

Limits (rem)

EAB 2.4E+00 6.3 LPZ 1.3E-01 6.3 Millstone Unit 3 Control Room 4.9E+00 5

Serial No.04-285 Discussion of Changes Page 50 of 98 3.3 Steam Generator Tube Rupture Accident This section describes the methods employed and the results of the Steam Generator Tube Rupture (SGTR) design basis radiological analysis.

This analysis included doses associated with the releases of the radioactive material initially present in primary liquid, secondary liquid and iodine spiking. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ), and in the control room. The methodology used to evaluate the control room and offsite doses resulting from the SGTR accident was consistent with Regulatory Guide 1.183 in conjunction with TEDE radiological units and limits, used ARCON96 based onsite atmospheric dispersion factors, and utilized Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors.

3.3.1 SGTR Scenario Description A SGTR is a break in a tube carrying primary coolant through the steam generator. This postulated break allows primary liquid to leak to the secondary side of one of the steam generators (denoted as the affected generator) with an assumed release to the environment through the steam generator Atmospheric Dump Valves (ADVs). The ADV on the affected steam generator is assumed to open to control steam generator pressure at the beginning of the event, and then fail fully open after operator action was taken to close the steam generator ADV.

The affected generator discharges steam to the environment for 2946 seconds (0.8183 hours0.0947 days 2.273 hours 0.0135 weeks 0.00311 months ) until the generator is isolated a second time by closure of the steam generator Atmospheric Dump Block Valve (ADBV). Break flow into the affected steam generator continues until 5596 seconds (1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months ), at which time the RCS is at a lower pressure. Additional releases from the affected steam generator are modeled from 2-8 hours to complete depressurization of the steam generator early in the event to maximize the dose consequences.

Depressurization of the steam generator is necessary to initiate Residual Heat Removal System (RHRS) cooling.

Serial No.04-285 Discussion of Changes Page 51 of 98 The intact generator (3 generators modeled as one) discharges steam for a period of 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months until the primary system has cooled sufficiently to allow a switchover to the RHRS, at 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , plus a 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months period of concurrent steaming. The additional 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months of steaming are required to reduce the system heatload to the point where RHRS can remove all the decay heat crediting only safety grade equipment to achieve cold shutdown and steaming is no longer required for cooldown.

No fuel damage is predicted as a result of a SGTR.

Therefore, consistent with the current licensing analysis basis, the SGTR analysis was performed assuming both a pre-accident iodine spike and a concurrent accident iodine spike. In addition, the impact of a coincident loss-of-offsite power (LOOP) at the time of tube rupture was considered.

In accordance with Regulatory Guide 1.183, The release of noble gases has been analyzed. Without credit for holdup in that scenario, the affected generator discharges steam to the environment for 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months after which the break flow stops and the generator block valve is closed. An alternate scenario for the affected steam generator was also evaluated for dose consequences associated with noble gases. Holdup of noble gases in the affected steam generator has been credited in the alternate scenario because of operator action to close the ADBV at 0.8183 hours0.0947 days 2.273 hours 0.0135 weeks 0.00311 months with the break flow continuing to enter the generator until 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months and subsequent release at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

From the period of 0.8183 hours0.0947 days 2.273 hours 0.0135 weeks 0.00311 months to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , noble gases are held up in the generator.

3.3.2 SGTR Source Term Definition Initial radionuclide concentrations in the primary and secondary systems for the SGTR accident are determined based on the maximum Technical Specification levels of activity. The SGTR accident analysis indicates that no fuel rod failures occur as a result of these transients. Thus, radioactive material releases were determined by the radionuclide concentrations initially present in primary liquid,

Serial No.04-285 Discussion of Changes Page 52 of 98 secondary liquid, and iodine spiking. These values are the starting point for determining the curie input for the RADTRAD-NAI code runs.

Regulatory Guide 1.183 specifies that the released activities should be the maximum allowed by the Technical Specifications.

Table 3.3-1 lists all the primary and secondary liquid radionuclide concentrations that are used in the analysis. Primary side concentrations are based on the Technical Specification limits of 100/ Ebar for gross gamma and 1.0 uCi/gm Dose Equivalent (DEQ) 1-131 for iodines.

Secondary side concentrations are based on the Technical Specification limit of 0.1 uCi/gm DEQ 1-131 for iodine. In addition, since there is not a Technical Specification limit for the secondary side gross gamma activity, one was derived from the design basis steam generator liquid activity to ensure that a suitably conservative source term was used.

Regulatory Guide 1.183 stipulates that SGTR accidents consider iodine spiking above the value allowed for normal operations based both on a pre-accident iodine spike and a concurrent accident spike. For Millstone Unit 3, the maximum iodine concentration allowed by Technical Specifications as the result of an iodine spike is 60 uCi/gm dose equivalent 1-131. This value is treated as the pre-accident iodine spike and is listed in Table 3.3-2.

Regulatory Guide 1.183 defines a concurrent iodine spike as an accident initiated value 335 times the appearance rate corresponding to the Technical Specification limit for normal operation (1 uCi/gm DEQ 1-131 RCS TS limit) for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The concurrent iodine spike appearance rates based on 335 times the 1.0 uCi/gm DEQ 1-131 concentration are listed in Table 3.3-3. Appearance rates address the issues raised by NSAL-00-004 [Reference 32].

The dose conversion factors used to calculate the TEDE doses and DEQ 1-131 for the Steam Generator Tube Rupture accident were taken from Table 3.1-3 for the isotopes required by Regulatory Guide 1.183 for the SGTR analysis.

Serial No.04-285 Discussion of Changes Page 53 of 98 Table 3.3-1 Primary Coolant and Secondary Side Liquid Nuclide Concentrations

RCS, Secondary Side Nuclides uCi/gm Water, uCi/gm 1131 7.721 E-01 7.956E-02 1132 2.716E-01 2.650E-02 1133 1.245E+00 1.132E-01 1134 1.700E-01 4.804E-03 1135 6.531 E-01 4.963E-02 Kr85m 4.654E+00 O.OOOE+00 Kr85 9.324E-02 O.OOOE+00 Kr87 3.340E+00 O.OOOE+00 Kr88 9.1 37E+00 O.OOOE+00 Xe133 7.112E+01 O.OOOE+00 Xe135m 3.118E+00 O.OOOE+00 Xe135 1.386E+01 O.OOOE+00 Xe138 1.618E+00 O.OOOE+00 Rb86 7.301 E-04 8.437E-07 Rb88 9.137E+00 5.138E-04 Rb89 2.796E-01 1.358E-05 Cs134 8.927E-01 1.877E-03 Cs136 4.600E-01 9.384E-04 Cs137 4.497E+00 9.411 E-03 Cs138 2.354E+00 2.401 E-04 Co58 4.597E-02 5.408E-05 Co60 5.679E-03 6.760E-06 Br84 1.1 15E-01 9.762E-06 Sr89 1.1 69E-02 1.366E-05 Sr90 4.624E-04 5.408E-07 Sr91 5.398E-03 3.921 E-06 Sr92 2.081 E-03 7.653E-07 Y90 5.627E-04 1.1 55E-06 Y91 1.875E-03 3.948E-06 Y92 2.024E-03 1.739E-06 Y93 9.313E-04 9.627E-07

RCS, Secondary Side Nuclides uCi/gm Water, uCi/gm Zr95 1.929E-03 2.247E-06 Zr97 1.1 27E-03 9.762E-07 Nb95 2.005E-03 2.342E-06 Mo99 9.240E+00 1.679E-02 Tc99m 5.211 E+00 1.239E-02 Ru 103 9.302E-04 1.082E-06 Ru105 8.248E-05 4.164E-08 Ru 106 8.840E-05 1.033E-07 Rh105 2.362E-04 2.374E-07 Sb127 3.905E-05 4.300E-08 Sb129 9.021 E-05 4.489E-08 Te127m 5.706E-03 6.652E-06 Te127 2.880E-03 4.651 E-06 Te1 29m 1.059E-01 1.230E-04 Te129 5.933E-02 1.141 E-04 Te131m 6.349E-02 6.220E-05 Te132 7.334E-01 7.977E-04 Te133m 6.998E-02 1.074E-05 Te134 8.451 E-02 1.079E-05 Ba139 2.096E-01 5.354E-05 Ba140 1.193E-02 1.368E-05 La140 4.086E-03 5.895E-06 Lal41 7.499E-04 3.597E-07 La142 2.282E-06 5.571 E-10 Ce141 1.893E-03 2.199E-06 Ce143 1.408E-03 1.398E-06 Ce144 1.343E-03 1.568E-06 Pr143 1.834E-03 2.131 E-06 Nd147 6.593E-04 7.545E-07 Np239 1.055E-02 1.1 17E-05

Serial No.04-285 Discussion of Changes Page 54 Of 98 Table 3.3-2 Pre-accident Iodine Spike RCS Concentration Iodine Activity in RCS at Iodine Activity in RCS at 60 times 1.0 DEQ 1-131 Nuclide 1.0 DEQ 1-131 uCi/gm uCi/gm 1131 7.72E-01 4.63E+01 1132 2.72E-01 1.63E+01 1133 1.25E+00 7.47E+01 1134 1.70E-01 1.02E+01 1135 6.53E-01 3.92E+01 Table 3.3-3 Concurrent Iodine Spike RCS Concentration Appearance rate for 1 uCi/gm DEQ 1-131, Spike = 335, 1 ui/g DE 1-31,SGTR Appearance Rate, Nuclide uCi/sec uCi/sec 1131 5.24E+03 1.75E+06 1132 7.13E+03 2.39E+06 1133 1.09E+04 3.64E+06 1134 9.91 E+03 3.32E+06 1135 8.77E+03 2.94E+06 3.3.3 Release Transport Affected Steam Generators 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 affected steam generator by the break flow. The break flow is terminated after 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months . A fraction of the break flow is assumed to flash to steam in the affected generator and to pass directly into the steam space of the affected generator with no credit taken for scrubbing by the steam generator liquid. The radionuclides entering the steam space as the result of flashing pass directly to the

Serial No.04-285 Discussion of Changes Page 55 Of 98 environment through the Steam Generator ADVs. The remainder of the break flow enters the steam generator liquid. Releases of radionuclides initially in the steam generator liquid and those entering the steam generator from the break flow are released as a result of secondary liquid boiling including an allowance for a partition factor of 100 for all non-noble gas isotopes.

Thus 1% of the iodines and particulates are released from the steam generator liquid to the environment along with the steam flow (moisture carryover is not actually modeled but is instead bounded by application of the partitioning factor). All noble gases are released from the primary system to the environment without reduction or mitigation. As was mentioned previously, an alternate noble gas release scenario was evaluated which considered isolation of the affected steam generator release to the environment by operator action at 0.8183 hours0.0947 days 2.273 hours 0.0135 weeks 0.00311 months after the tube rupture had occurred, while the break flow continues into the generator until 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months , and subsequent allowance for depressurization of the generator and release of the accumulated contents from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months post-accident. The transport model utilized for iodine and particulates was consistent with Appendix E of Regulatory Guide 1.183.

Intact Steam Generators 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 intact generators by the leak rate limiting condition for operation (1 gpm) specified in the Technical Specifications. All radionuclides in the primary coolant leaking into the intact generator are assumed to enter the steam generator liquid. Releases of radionuclides initially in the steam generator liquid and those entering the steam generator from the leakage flow are released as a result of secondary liquid boiling, including an allowance for a partition factor of 100 for all non-noble gas isotopes. Thus 1% of the iodines and particulates are assumed to pass into the steam space and then directly to the environment. All noble gases that are released from the primary system to the intact generator are released to the

Serial No.04-285 Discussion of Changes Page 56 Of 98 environment without reduction or mitigation.

Releases were assumed to continue from the intact generator for a period of 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months until the primary system cools to below 3500F and the RHRS can remove 100% of decay heat with no requirement for steaming to augment cooldown. The 18-hour steaming period is based on the time necessary to cooldown crediting safety grade equipment only.

3.3.4 Determination of Atmospheric Dispersion Factors 3.3.4.1 Control Room Atmospheric Dispersion Factors The onsite atmospheric dispersion factors were calculated by Dominion using the ARCON96 code and guidance from RG-1.194 [Reference 10] for the control room. Site meteorological data taken over the years 1997-2001 were used in the evaluations. The control room atmospheric dispersion factors are presented in Table 1.3-4 under "Main Steam Valve Building Ventilation Exhaust".

3.3.4.2 Offsite Atmospheric Dispersion Factors The EAB and LPZ X/Q values used in the SGTR analysis are listed in Table 1.3-

3. The EAB values are listed under "Other Release Points" and the LPZ values are listed under 'Non-Millstone Stack Release Points".

3.3.5 SGTR Key Analysis Assumptions and Inputs The Basic Data and Assumptions are listed below in Table 3.3-4. A time line of events is provided in Table 3.3.5. Steam and break flow data are listed in Tables 3.3-6 to 3.3-8.

Generic data such as control room information is available in Table 1.3-1.

Table 3.3-4 Basic Data and Assumptions for the SGTR Accident Data / Assumption Value Primary to Secondary Leak Rate - Technical 1 gpm (intact steam generator)

Serial No.04-285 Discussion of Changes Page 57 Of 98 Table 3.3-4 Basic Data and Assumptions for the SGTR Accident Specification limit Release coincident with loss of off-site power

Serial No.04-285 Discussion of Changes Page 58 Of 98 Table 3.3-4 Basic Data and Assumptions for the SGTR Accident Data / Assumption Value Release points: (On Loss Of Off-site Power, the Steam Generator Atmospheric condenser is not available for cooling, an ADV Dump Valves (ADVs) on the affected steam generator is assumed stuck open until closed by an operator.

Additional cooling is by ADVs on intact steam generators.)

Credited Operator Actions

Affected Steam Generator 0.8183 hrs.

Close stuck open ADV after 1742 seconds (29 minutes)

Close failed open ADV after an additional 1200 seconds (20 minutes)

Credited Operator Actions Intact Steam Generators 18 hrs.

Actions commensurate with cooldown using only safety grade equipment Iodine chemical form (%) released from steam Elemental 97 generators to environment Organic 3

Iodine Partitioning PC for iodine = 100 Moisture Carryover in Intact Steam Generators 1%

Tube Uncovery.

No tube bundle uncovery assumed.

Assumption consistent with conclusions in WCAP-13132 (Reference 33)

Release to Environment Duration Intact steam generators 0 - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Affected steam generator 0 - 0.8183 hours0.0947 days 2.273 hours 0.0135 weeks 0.00311 months & 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

Serial No.04-285 Discussion of Changes Page 59 Of 98 Table 3.3-4 Basic Data and Assumptions for the SGTR Accident Release to Environment Duration Affected steam generator (alternate scenario -

From 0 to 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months (break noble gas only) flow stops after 1.554 hours0.00641 days 0.154 hours 9.160053e-4 weeks 2.10797e-4 months )

Data / Assumption Value Initial Steam Generator Steam Mass 8,870 Ibm / steam generator Initial Steam Generator Liquid Mass 97,222 Ibm / steam generator (updated from Westinghouse thermal-hydraulic analysis)

Control Room Ventilation Timing Same as for the Fuel Handling Accident (Table 3.2.1)

Assumption: Dose consequences from release of initial secondary side steam is not significant Control Room Plume and Filter Shine Dose to the Millstone Unit 3 Control Room

Conservatively set the same as the LOCA analyses

Serial No.04-285 Discussion of Changes Page 60 Of 98 Table 3.3-5 Time Line of Events Time, post Event accident seconds hours 0

0 SGTR - ADV sticks open LOOP 10 0.0028 Control Room Isolates 109 0.03028 Reactor Trip 381 0.1058 Sl Actuated 1742 0.4839 Affected SG Isolated 1744 0.4844 Affected SG ADV Fails Open 2946 0.8183 Affected SGADBV Closed (Release Terminated) 5596 1.554 Break Flow Terminated 6065 1.685 Control Room Emergency Ventilation 7200 2

Affected SG Depressurization Initiated (Release Re-Initiated) 8 Affected SG Depressurized, (Release Terminated) 11 RCS at 3500F; RHRS Placed In Service 18 RHRS capable of 100% of cooldown, Intact SG release stops 55 RCS at 2120F Primary-to-Secondary Leak Stops 720 Event Terminated

Serial No.04-285 Discussion of Changes Page 61 Of 98 Table 3.3-6 RCS Break Flow to Affected Steam Generator Time period Total Flashed Liquid Break Break Break Flow Rate Flow Rate Flow Rate From To (hour)

Ibm/min Ibm/min Ibm/min 0

0.03028 2642 423 2219 0.03028 0.03153 2390 423 1967 0.03153 0.4063 2390 149 2241 0.4063 0.7357 2390 377 2013 0.7357 0.8890 2390 161 2229 0.8890 0.9279 2390 41 2349 0.9279 1.554 2390 0

2390 1.554 2

0 0

0 Table 3.3-7 Affected Steam Generator Steam Release to Environment Time period, sec Time period, hour Release

Rate, From To From To (Ibm/min) 0 109 0

0.03028 64569 109 2946 0.03028 0.8183 3923 2946 7200 0.8183 2

0 7200 28800 2

8 95

Serial No.04-285 Discussion of Changes Page 62 Of 98 Table 3.3-8 Intact Steam Generator Steam Release to the Environment Time period, sec Time period, hour Release Rate From To From To (Ibmlmin) 0 109 0

0.03028 192220 109 1762 0.03028 0.4895 4540 1762 5596 0.4895 1.554 3641 5596 7200 1.554 2

4970 7200 28800 2

8 2614 28800 39600 8

11 2614 39600 64800 11 18 4563 18 720 0

3.3.6 SGTR Analysis Results The results of the analyses are presented below for the Concurrent Spike (Table 3.3-9) and for the Pre-accident Iodine Spike (Table 3.3-10).

Table 3.3-9 Dose Summary for the SGTR Concurrent Iodine Spike Location TEDE (rem)

Limits (rem)

EAB 9.OE-01 2.5 LPZ 9.OE-02 2.5 Millstone Unit 3 Control Room 1.3E+00 5

Table 3.3-10 Dose Summary for the SGTR Pre-accident Iodine Spike Location TEDE (rem)

Limits (rem)

EAB 2.1 E+00 25 LPZ 1.8E-01 25 Millstone Unit 3 Control Room 3.OE+00 5

Serial No.04-285 Discussion of Changes Page 63 of 98 3.4 Main Steam Line Break Analysis This section describes the methods employed and results of the Main Steam Line Break (MSLB) design basis radiological analysis. This analysis included doses associated with the releases of the radioactive material initially present in primary and secondary liquids at Technical Specification concentrations and adjusting for iodine spiking scenarios. No fuel failure is expected. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ), and in the Millstone Unit 3 Control Room.

The methodology used to evaluate the control room and offsite doses resulting from the MSLB accident is consistent with Regulatory Guide 1.183 (Reference 1) in conjunction with TEDE radiological units and limits, used ARCON96 based onsite atmospheric dispersion factors, and utilized Federal Guidance Report No. 11 and 12 (References 7 & 8, respectively) dose conversion factors.

3.4.1 MSLB Scenario Description The Main Steam Line Break (MSLB) accident begins with a break in one of the main steam lines leading from a steam generator (affected generator) to the turbine. In order to maximize control room dose, the break is assumed to occur in the turbine building. The affected steam generator releases steam for 55.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , at which time the RCS has cooled down to 2120F and release via this pathway terminates.

The 55.2-hour steaming period is based on the time necessary to cooldown crediting safety grade equipment only.

Also, it is expected that the affected generator will dry out in 56.3 seconds post-MSLB.

Loss Of Off-site Power is assumed. As a result, the condenser is unavailable and cool down of the primary system is through the release of steam from the intact generators. The release from the intact generators continues for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months through the ADVs until the RHRS can fully remove decay heat. In accordance with RG 1.183, Appendix E, two independent cases are evaluated. Case one assumes a pre-accident iodine spike, while the second case assumes a concurrent iodine spike.

Serial No.04-285 Discussion of Changes Page 64 of 98 3.4.2 MSLB Source Term Definition As with the SGTR accident, the analysis of the MSLB accident indicates that no additional fuel rod failures occur as a result of the transient. Thus, radioactive material releases are determined by the radionuclide concentrations initially present in primary and secondary liquid at maximum Technical Specification limits and iodine spiking.

The Main Steam Line Break analysis uses the primary and secondary liquid source term discussed in Table 3.3-1 and the pre-accident iodine spike source term discussed in Table 3.3-2. The MSLB analysis also assumes a concurrent iodine spike as an accident initiated value 500 times the appearance rate corresponding to the Technical Specification limit for normal operation (1 uCi/gm DEQ 1-131 RCS TS limit) for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and that source term is listed below in Table 3.4-1.

Table 3.4-1 Concurrent Iodine Spike Appearance rate for I uCgm DG 1-31,Spike = 500, gm E MSLB Appearance Rate, Nuclide uCVsec uCl/sec 1131 5.24E+03 2.62E+06 1132 7.13E+03 3.57E+06 1133 1.09E+04 5.44E+06 1134 9.91 E+03 4.96E+06 1135 8.77E+03 4.39E+06 3.4.3 Release Transport 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 steam generators by the leak-rate limiting condition for operation (1 gpm) specified in

Serial No.04-285 Discussion of Changes Page 65 of 98 the Technical Specifications. The maximum amount of primary to secondary leakage allowed by the Technical Specifications to any one steam generator is 500 gallons per day.

This leakage (500 gpd equivalent to 0.35 gpm) was assigned to the affected generator.

For the affected generator, the release pathway is assumed to pass directly into the turbine building with no credit taken for holdup, partitioning or scrubbing by the steam generator liquid. No credit is taken for any holdup or dilution in the Turbine Building. From the Turbine Building it passes to the environment and to the control room.

During the first 56.3 seconds post-trip, the affected steam generator 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 steam generator. During the first 55.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the primary coolant is also assumed to leak into the affected steam generator at the rate of 500 GPD with all activity released unmitigated to the environment. After 55.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the RCS will have cooled to below 2120F and the release via this pathway terminates.

The transport model utilized for noble gases, iodine and particulates was consistent with Appendix E of Regulatory Guide 1.183.

The remainder of the 1 gpm primary side to secondary side leakage, 0.65 gpm, was assigned to 2 intact generators. This leakage continues for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months until shutdown cooling is credited for decay heat removal. The third intact generator is assumed to have a failed closed atmospheric dump valve, which reduces the holdup volume to 2 generators instead of 3, but the steaming rate has not been reduced, which maximizes the release rate.

There are several nuclide transport models associated with the intact steam generators. Together, they ensure proper accounting of gross gamma, iodine and noble gas releases. The first pathway releases gross gamma activity, at the Technical Specification limit of 100/Ew, to the SG liquid volume at 0.65 gpm.

Serial No.04-285 Discussion of Changes Page 66 of 98 Releases of radionuclides initially in the steam generator liquid and those entering the steam generator from the primary to secondary leakage flow are released as a result of secondary liquid boiling. Due to moisture carryover, 1% of the particulates in the steam generator bulk liquid are released to the environment at the steaming rate. Radionuclides initially in the steam space do not provide any significant dose contribution and are not considered. The transport to the environment of noble gases from the primary coolant and from particulate daughters occurs without any mitigation or holdup.

The pre-accident iodine spike is modeled in the same manner as the gross gamma model previously discussed.

The concurrent iodine spike model is modeled in the same manner as the gross gamma model but the iodine spike occurs for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after which the activity remaining in the primary coolant continues to be released for the remainder of the 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

3.4.4 Determination of Atmospheric Dispersion Factors 3.4.4.1 Control Room Atmospheric Dispersion Factors The onsite atmospheric dispersion factors were calculated by Dominion using the ARCON96 code and guidance from RG-1.194 (Reference 10) for the Control Rooms. Site meteorological data taken over the years 1997-2001 were used in the evaluations.

Control room X/Q values were calculated from the closest ventilation point on the Turbine Building to the control room inlet to maximize dose. These and other control room atmospheric dispersion factors are presented in Table 1.3-4.

3.4.4.2 Offsite Atmospheric Dispersion Factors X/Q's from the Turbine Building Ventilation Stack are used for offsite doses since the ventilation stack is located near the Turbine Building roof exhaust vents.

Serial No.04-285 Discussion of Changes Page 67 of 98 3.4.5 MSLB Key Analysis Assumptions and Inputs 3.4.5.1 Basic Data and Assumptions The basic data and assumptions are listed below in Table 3.4-2. All numeric values specific to this evaluation are listed in this section. Generic data such as control room information is available in Tables 1.3-1 and 1.3-2. Steam generator mass releases and timings are a product of updated Westinghouse thermal-hydraulic analyses.

Table 3.4-2 Basic data and Assumptions for the MSLB Accident Data / Assumption Value Loss of Offsite Power:

Assumed to Occur at Accident Initiation Release Points:

Affected Steam Generator:

Turbine Bldg Intact Steam Generator:

ADVs Iodine Partition Coefficients (PC) in Intact Steam 100 Generators:

Moisture Carryover in Intact Steam Generators:

1%

Primary-to-Secondary Leakage:

Affected SG: 500 GPD Total:

1 GPM Steam Generator Liquid Mass:

164,200 Ibm Control Room Ventilation Timing Same as Fuel Handling Accident (Table 3.2.1)

Duration of Steam Generator Release:

Affected Steam Generator: 55.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (to reach RCS temperature of 212 0F)

Intact Steam Generators: 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months (to enter RHRS window of operation)

Steam Release from affected Steam Generator:

Initial Inventory:

0 - 56.3 seconds: 1.75E+05 Ibm/min Primary-to-Secondary Leak 0 - 55.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: 2.918 Ibm/mmin (= 0.35 gpm)

Serial No.04-285 Discussion of Changes Page 68 of 98 Table 3.4-2 Basic data and Assumptions for the MSLB Accident Steam Release from Intact Steam Generators:

0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: 3.41 E+03 Ibm/ min 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:

2.73E+03 Ibm/ min 8 - 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months s: 4.56E+03 Ibm/ min Millstone Unit 3 Auxiliary Feed System is available to maintain water level in intact steam generators.

Control Room Plume and Filter Shine Dose to the Millstone Unit 3 Control Room

Conservatively set the same as the LOCA analyses 3.4.5.2 MSLB Analysis Results The total TEDE to the EAB, LPZ and Millstone Unit 3 Control Room from a Millstone Unit 3 Main Steam Line Break is summarized below for the concurrent (Table 3.4-3) and pre-accident spike (Table 3.4-4). The concurrent spike results in the highest dose consequences for both offsite and the control room.

All doses are within the limits specified in Regulatory Guide 1.183 and 10 CFR 50.67.

Serial No.04-285 Discussion of Changes Page 69 of 98 Table 3.4-3 TEDE - Concurrent Iodine Spike Location TEDE (rem)

Limits (rem)

EAB 3.6E-01 2.5 LPZ 1.8E-01 2.5 Millstone Unit 3 Control Room 3.OE+00 5

Table 3.4-4 TEDE - Pre-accident Iodine Spike Location TEDE (rem)

Limits (rem)

EAB 9.1 E-02 25 LPZ 3.6E-02 25 Millstone Unit 3 Control Room 1.2E+00 5

Serial No.04-285 Discussion of Changes Page 70 of 98 3.5 Locked Rotor Analysis This section describes the methods employed and results of the Locked Rotor Accident (LRA) design basis radiological analysis. The analysis assumes failure of 7% of the fuel rods, due to Departure from Nucleate Boiling (DNB) during the accident. Doses were calculated at the Exclusion Area Boundary (EAB), at the Low Population Zone (LPZ), and in the Millstone Unit 3 Control Room. The methodology used to evaluate the control room and offsite doses resulting from the LRA included Regulatory Guide 1.183 methodology, ARCON96-based control room atmospheric dispersion factors, and Federal Guidance Reports (FGR) No. 11 and 12 dose conversion factors.

3.5.1 Locked Rotor Scenario Description The Locked Rotor Accident (LRA) begins with instantaneous seizure of the reactor coolant pump rotor under 4 loop operation. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer, which results in assumed fuel damage due to Departure from Nucleate Boiling (DNB). Although there is no increase in the leak rate of primary coolant to the secondary side during the LRA, a larger amount of activity (from the failed fuel) is transported to the secondary side via any pre-existing leaks in the steam generators.

A turbine trip and coincident loss of offsite power are incorporated into the analysis.

This results in a release through a stuck open steam generator atmospheric dump valve (ADV) and a parallel release from the intact steam generators. The stuck-open ADV represents the assumed, single, active failure.

Consistent with current licensing bases, operator action is credited with closure of the ADV after 20 minutes.

Serial No.04-285 Discussion of Changes Page 71 of 98 3.5.2 Locked Rotor Source Term Definition The core source term used in the Locked Rotor Analysis is taken from Table 3. 1-

3. Analyses are based on 7% of the gap activity being released.

3.5.3 Release Transport The release scenario uses the Technical Specification primary to secondary leakage limits of 1 gpm total and 500 gpd from the affected steam generator.

The release from the affected steam generator continues for 20 minutes until operator action isolates that release pathway.

The balance of the 1 gpm limit (0.65 gpm) is released from the intact steam generators over the course of 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months until shutdown cooling can be implemented to fully remove decay heat. At this point the release from the intact steam generators is terminated when the operator closes the ADVs.

The RADTRAD-NAI computer code (Reference 3) is used to model the time dependent transport of radionuclides, from the primary to secondary side and out to the environment via ADVs.

3.5.4 Determination of Atmospheric Dispersion Factors (X/Q) 3.5.4.1 Control Room Atmospheric Dispersion Factors (X/Q)

The Millstone Unit 3 Control Room X/Q values were calculated by using the ARCON96 code and guidance from RG 1.194. The control room X/Q values are calculated at the Main Steam Valve Building. No credit is taken for the thermal plume rise. These X/Q values are given in Table 1.3-4.

3.5.4.2 Offsite XQ The EAB and LPZ X/Q values used in the Locked Rotor analysis are the same as those used in the LOCA analysis listed in Table 1.3-3.

Serial No.04-285 Discussion of Changes Page 72 of 98 3.5.5 Locked Rotor Analysis Assumptions and Key Parameters 3.5.5.1 Basic Data and Assumptions All numeric values specific to this evaluation are listed in Table 3.5-1. Generic data such as control room information is available in Tables 1.3-1 and 1.3-2.

Table 3.5-1 Basic Data and Assumptions for the LRA Parameter / Assumption Value Fuel Damage 7% Fuel Failure Radial Peaking Factor:

1.7 Primary to Secondary Leak Rate:

0.35 gpm (affected SG)

(Technical Specifications) 0.65 gpm (intact SGs)

Release from secondary side is coincident with loss of off-site power Release Points:

Steam Generator Atmospheric Dump Valve (ADV)

Credited Operator Actions These operator actions are

a. Operator Action to close ADV after 20 currently part of the Licensing minutes Bases.

b. Operator actions commensurate with cooldown using only safety grade equipment Iodine Chemical Form Released from Steam Elemental 97%

Generators to Environment:

Organic 3%

Fraction of Fission Product Inventory in Gap Halogens: 0.8 Noble Gases: 0.10 Alkali Metals: 0.12 Iodine Partitioning in Intact Steam Generator 100 Intact Steam Generator Tube Uncovery No tube bundle uncovery assumed.

Serial No.04-285 Discussion of Changes Page 73 of 98 Table 3.5-1 Basic Data and Assumptions for the LRA Parameter / Assumption Value Affected Steam Generator Tube Uncovery

Affected steam generator goes dry, immediately

100% flashing is assumed

Conservative assumption because no feedwater is credited to this generator and the mass of water pre-existing in the generator, 9.722E+04 lb

Contents discharged out the ADV at 820,000 lb/hr

SG dries out in approximately 7 minutes.

Release Duration:

Intact Steam Generators - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months based on cooldown using only safety grade equipment

Affected Steam Generator - 20 minutes based on operator action Total Steam Flows to Atmosphere from 3 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: 251,000 Ibm Intact Steam Generators 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s: 1,031,000 Ibm 11-18 hours: 1,915,359 Ibm Mass Flow Rates from 3 Intact Steam 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s: 1.255E+05 lbm/hr Generators 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s: 1.718E+05 Ibm/hr 8 - 11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months s: 2.736E+05 Ibm/hr 11 - 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months s: 2.736E+05 Ibm/hr Moisture carryover in intact Steam 1%

Generators:

Initial Steam Generator Liquid Mass 4.414E+07 grams / Steam Generator Millstone Unit 3 Steam Generator ADV 820,000 Ibm/hr @ 1140 psia Maximum Flow Rate Control Room Ventilation Timing Same as the Fuel Handling Accident (Table 3.2-1)

Assumption:

Dose consequences are from release of initial secondary side liquid

Dose consequences from the release of steam is not significant Control Room Plume and Filter Shine Dose to the Millstone Unit 3 Control Room

Conservatively set the same as the LOCA analyses

Serial No.04-285 Discussion of Changes Page 74 of 98 3.5.6 Locked Rotor Analysis Results The results of the design basis Locked Rotor analysis are presented in Table 3.5-

2. These results report the calculated dose for the worst 2-hour interval (EAB),

and for the assumed 30-day duration of the event for the control room and the LPZ. As stated in Table 3.5-1, plume and filter shine to the control room are conservatively based on LOCA results and will be used here. The doses are calculated with the TEDE methodology, and are compared with the applicable acceptance criteria specified in 10 CFR 50.67 and Regulatory Guide 1.183.

Table 3.5-2 TEDE Results for the Locked Rotor Accident Location TEDE (rem)

Limits (rem)

EAB 2.3E+00 2.5 LPZ 3.7E-01 2.5 Millstone Unit 3 Control Room 3.2E+00 5

Serial No.04-285 Discussion of Changes Page 75 of 98 3.6 RCCA Ejection Accident Analysis This section describes the evaluation of TEDE at the EAB, LPZ and MP3 Control Room from a Millstone Unit 3 Rod Control Cluster Assembly (RCCA) Ejection Accident (REA) using the AST. Two release cases are considered. The first case is a release into the containment. The second release is a release into the primary coolant, which is released through the secondary system.

3.6.1 RCCA Ejection Accident Scenario Description This accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a RCCA and drive shaft.

The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

3.6.2 RCCA Ejection Accident Source Term Definition The core source term used in the RCCA Ejection Accident Analysis are taken from Table 3.1-3. The release of the core source term is adjusted for the fraction of fuel rods assumed to fail during the accident and the fractions of core inventory assumed to be in the pellet-to-clad gap.

The analysis is based on the assumption that 10% failed fuel and 0.25% melted fuel occurs during a RCCA Ejection Accident.

3.6.3 Release Transport Two release paths are considered for the REA: containment leakage and the secondary system.

Serial No.04-285 Discussion of Changes Page 76 of 98 The containment release transport assumptions and methodology are similar to the LOCA and can be found in section 3.1.5, with a few exceptions.

The exceptions are:

1) The core release fractions are based on Appendix H of R.G. 1.183. The core release fractions are 0.010625 for halogens and 0.0125 for noble gases based on the consequences of 10% failed fuel and 0.25% melted fuel.

2) Containment sprays do not initiate due to a REA. Therefore there are no consequences from ECCS leakage and RWST backleakage.

3) Natural deposition in the containment is not assumed.

4) Containment leak rate is reduced by 50% at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for both offsite and control room analyses.

5) The safety injection signal is initiated 2 minutes after a REA. Therefore the control room is not isolated until 2 minutes 10 seconds following a REA.

The second release path is via the secondary system.

The activity in the secondary system release is based on Appendix H of RG 1.183.

The core release fractions are 0.01125 for halogens and 0.0125 for noble gases based on the consequences of 10% failed fuel and 0.25% melted fuel.

The iodines released from the steam generators are assumed to be 97% elemental and 3%

organic. The primary-to-secondary leak rate of 1 gpm, which is specified in the technical specifications, exists until shutdown cooling is in operation and release from the steam generators terminate. All noble gas radionuclides released to the secondary system are released to the environment without reduction or mitigation. The condenser is not available due to a loss of offsite power. A partition coefficient for iodine of 100 is assumed in the steam generators.

The primary-to-secondary leak occurs during the first 1,200 seconds of the REA (until primary system pressure is less than secondary side system pressure).

Steam generator mass releases are a product of updated Westinghouse thermal-

Serial No.04-285 Discussion of Changes Page 77 of 98 hydraulic analyses. The steam released via the safeties/ADVs during the REA and subsequent cooldown is listed in Table 3.6-1.

Serial No.04-285 Discussion of Changes Page 78 of 98 3.6.4 Determination of Atmospheric Dispersion Factors (XQQ) 3.6.4.1 Control Room Atmospheric Dispersion Factors (X/Q)

The Millstone Unit 3 Control Room X/Q values are given in Table 1.3-4.

3.6.4.2 Offsite Atmospheric Dispersion Factors (X/Q)

The EAB and LPZ X/Q values used in the REA analysis are the same as those used in the LOCA analysis and are listed in Table 1.3-3 3.6.5 RCCA Ejection Accident Analysis Assumptions and Key Parameters 3.6.5.1 Basic Data and Assumptions The basic data and assumptions are listed below in Table 3.6.1. Generic data, such as control room information, is available in Tables 1.3-1 and 1.3-2.

Table 3.6-1 Basic Data and Assumptions for the REA Parameter / Assumption Value Core Release Fractions for Breached 10% of noble gasses and iodines in the Fuel: (Appendix H of Reference 1) gap Core Release Fractions for Melted 100% noble gasses and 25% iodines Fuel: (Appendix H of Reference 1)

Percentage of Failed Fuel following a 10%

REA:

Percentage of Melted Fuel following a 0.25%

REA:

Core Release Fractions for Secondary 100% noble gasses and 50% iodines in Side Release: (Appendix H of the fraction released to the reactor Reference 1):

coolant Safety Injection (SI) Signal Initiated 2 minutes after a REA:

Serial No.04-285 Discussion of Changes Page 79 of 98 Table 3.6-1 Basic Data and Assumptions for the REA Parameter / Assumption Value Iodine Chemical Form Released from 3% Organic Iodide the Steam Generators to the 97% Elemental Iodine Environment: (Reference 1)

Total reactor-to-secondary leakage 1 gpm through all steam generators:

(Technical Specifications, Section 3.4.6.2.c)

Time for primary system pressure to 1,200 seconds fall below secondary system pressure:

Duration of steam releases:

18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months Steam released from t=0 to 1200 200,000 Ibm seconds (primary system depressurization):

Steam released from 2 - 11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months s:

1.547E+06 Ibm Steam released from 11 - 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months s:

1.916E+06 Ibm 3.6.6 RCCA Ejection Accident Analysis Results The total TEDE to the EAB, LPZ, and the Millstone Unit 3 Control Room from a Millstone Unit 3 RCCA Ejection Accident (REA) is summarized below for the containment pathway (Table 3.6-2) and the secondary side release pathway (Table 3.6-3). The dose to the EAB and LPZ is less than the 6.3 rem TEDE limit stated in 10CFR50.67 and Regulatory Guide 1.183. The EAB dose represents the worst 2-hour dose for each release pathway. The dose to the Millstone Unit 3 Control Room is less than the 5 rem TEDE limit specified in 10CFR50.67 and Regulatory Guide 1.183.

Serial No.04-285 Discussion of Changes Page 80 of 98 Table 3.6-2 TEDE from a Millstone Unit 3 REA (containment)

Location TEDE (rem)

EAB 8.7E-01 LPZ 4.8E-01 Millstone Unit 3 Control Room 8.3E-01 Table 3.6-3 TEDE from a Millstone Unit 3 REA (secondary side)

Location TEDE (rem)

EAB 1.2E-01 LPZ 1.5E-02 Millstone Unit 3 Control Room 5.3E-02

Serial No.04-285 Discussion of Changes Page 81 of 98 4.0 ADDITIONAL DESIGN BASIS CONSIDERATIONS In addition to the explicit evaluation of radiological consequences that had direct impact from the changes associated with implementing the AST, other areas of plant design were also considered for potential impacts. The evaluation of these additional design areas is documented below.

4.1 Impact Upon Equipment Environmental Qualification In the Federal Register notice issuing the final rule for use of alternative source terms at operating reactors (Reference 38), the NRC stated that it will evaluate this issue as a generic safety issue to determine whether further regulatory actions are justified. This issue was subsequently designated as Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification.

Further guidance is provided in SECY-99-240 (Reference 39), which transmitted the final AST rule changes for the Commission's approval.

The following is stated in the 'Discussion' section, regarding evaluation of the equipment qualification issue before its final resolution:

'in the interim period before final resolution of this issue, the staff will consider the TID-14844 source term to be acceptable in reanalyses of the impact of proposed plant modifications on previously analyzed integrated component doses regardless of the accident source term used to evaluate offsite and control room doses.'

In NUREG-0933, Supplement 25 (Reference 40), the NRC staff reported its conclusions concerning the assessment of Issue 187. The staff concluded that there was no clear basis to require that the equipment qualification design basis be modified to adopt the AST. It was stated that there would be no discernable risk reduction from such a requirement.

This issue was thus dropped from further pursuit. Consistent with this guidance, no further evaluation of this issue is presented in support of implementing the AST for Millstone Unit 3.

The

Serial No.04-285 Discussion of Changes Page 82 of 98 existing equipment qualification analyses, which are based upon the TID-14844 source term, are considered acceptable.

4.2 Risk Impact of Proposed Changes Associated with AST Implementation Implementation of the AST is of benefit to licensees because of the potential to obtain relaxation in specific safeguard systems operability or surveillance requirements, since such changes can reduce regulatory burden and streamline operations. Such changes are warranted if they can be pursued without creating an unacceptable impact upon plant risk characteristics as compared with the existing system licensing and operational basis.

The proposed changes associated with implementation of the AST for Millstone Unit 3 have been considered for their risk effects.

A discussion of these considerations is presented below.

The proposed changes are presented here for convenience; these changes are described in report sections 2.2 through 2.6:

a. The definition of Dose Equivalent 1-131 in Section 1.10 of the Technical Specifications Definitions is revised to reference Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, " 1989, as the source of thyroid dose conversion factors (Reference 7).

b. Change Technical Specification 3/4.7.7, "Control Room Emergency Air Filtration System," Surveillance Requirements c.2 and d to reflect a methyl penetration less than or equal to 5% for the Control Room Emergency Air Filtration System filters instead of 2.5%.

Serial No.04-285 Discussion of Changes Page 83 of 98

c. Delete Technical Specification 3/4.7.8, "Control Room Envelope Pressurization System."

The Control Room Envelope Pressurization System is no longer credited in the accident analyses described in evaluation.

d. Change the leakage rate acceptance criteria for all penetrations that are Secondary Containment bypass leakage paths in Section 6.8.4.f, "Containment Leakage Testing Program," from ' 0.042 La to 5 0.06 La.

Item a - This change allows the use of dose conversion factors from FGR 11.

These dose conversion factors have been previously found to be acceptable for use in dose calculations. This change has no impact upon plant risk from severe accident scenarios.

Item b - Item d - The changes to the specified test acceptance criteria for the Control Room Emergency Air Filtration System ensure that system performance remains consistent with the assumptions of the accident analyses. Although the changes represent relaxation of acceptance criteria, this is judged to represent negligible risk impact since the most risk-significant scenarios are for accident sequences in which filtration systems have already lost power or are ineffective in reducing radioactive releases.

In addition, the risk associated with modification and/or elimination of such filtration systems was evaluated during the NRC's rebaselining study (Reference 37). Reference 37 reported that the effect on overall risk from filtration system modifications was small.

This effect was attributed to the fact that filtration systems, which require electrical power for operation, will already not be functional for certain risk-significant accident sequences (e.g., station blackout).

It is concluded that the proposed changes in test acceptance criteria will produce negligible impact upon overall plant risk for such accident sequences.

Serial No.04-285 Discussion of Changes Page 84 of 98 Items c and d - These changes remove the pressurization of the control room early in the accident sequence and increase the leakage rate acceptance criteria for all penetrations that are Secondary Containment bypass leakage paths.

These changes are accomplished while maintaining calculated doses to the Control Room operators, the EAB and the LPZ within the TEDE limits of 10CFR50.67. These limits have been judged to be acceptable consequences.

These changes are expected to have negligible impact on plant risk associated with severe accident sequences.

It is concluded that the proposed changes associated with AST implementation for Millstone Unit 3 will have insignificant effect upon the risk associated with severe accidents. This is primarily due to the fact that the risk significant accident sequences involve the failure of systems or structures (e.g., containment) that are not impacted by the relatively minor operational changes proposed herein.

4.3 Impact Upon Emergency Planning Radiological Assessment Methodology This application of the AST for Millstone Unit 3 replaces the existing design basis source term with the source term defined in RG 1.183. The MIDAS model that is employed for emergency planning radiological assessments includes definitions of source terms for various design basis accidents. Calculated results from MIDAS are used in various emergency preparedness processes. The basis of the existing source term definitions in the MIDAS calculations will be evaluated to determine: 1) the manner in which the source terms used in emergency preparedness activities rely upon the design basis event source term definition and 2) what specific changes may be warranted in the emergency preparedness source terms and their detailed usage. This assessment of potential impact will also include radiation monitor setpoint calculations for accident high range monitors, which use data input similar to MIDAS.

Serial No.04-285 Discussion of Changes Page 85 of 98 5.0 Conclusions The alternative source term defined in Regulatory Guide 1.183 has been incorporated into the reanalysis of radiological effects from six key accidents for Millstone Unit 3. This amendment request represents a full implementation of the alternative source term, making RG 1.183 the licensing basis source term for assessment of design basis events. The analysis results from the reanalyzed events meet all of the acceptance criteria as specified in 10 CFR 50.67 and RG 1.183.

Serial No.04-285 Discussion of Changes Page 86 of 98 6.0 References

1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," USNRC, Office of Nuclear Regulatory Research," July 2000.

2. 10 CFR 50.67, "Use of Alternative Source Terms at Operating Reactors,'

Final Rule, in Federal Register No. 64, p. 71990, December 23,1999.

3. RADTRAD-NAI 1.1 (QA), Numerical Applications Inc.

4. NUREG/CR-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," USNRC, June 1997, S.L.

Humphreys et al.

5. NUREG/CR-6331, Rev. 1, 'Atmospheric Relative Concentrations in Building Wakes, ARCON96," USNRC, 1997.

6. Computer Code SCALE 4.4a Version 1 Mod 0.

7. Technical Information Document TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," United States Atomic Energy Commission, 1962.

8. 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, 1988.

9. Federal Guidance Report No. 12, External Exposures to Radionuclides in Air, Water and Soil," EPA 420-r-93-081, Environmental Protection Agency, 1993.

10. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, June 2003

11. NUREG-0800, "Standard Review Plan," Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," U.S. Nuclear Regulatory Commission, Rev. 2, December 1988.

12.NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," July 01, 1996.

Serial No.04-285 Discussion of Changes Page 87 of 98

13. NRC Letter dated September 16, 2002, Victor Nerses (USNRC) to J. A.

Price (DNC), "Millstone Power Station, Unit 3 - Issuance of Amaendment Re: Revised Final Safety Analysis Report Licensing Masis for the Post-Accident Operation of the Supplementary Leakage Collection and Release System (TAC No. MB3700)."

14. NRC Letter dated November 25, 2002, Victor Nerses (USNRC) to J. A.

Price (DNC), "Millstone Power Station, Unit No. 3 - Correction to Safety Evaluation (SE) for Amendment No. 211 (TAC No. MB3700)."

15. BNP-100, Iodine Removal from Containment Atmospheres by Boric Acid Spray, July 1970.

16. "Radiological Health Handbook", January 1970.

17. WCAP-7828, "Radiological Consequences of a Fuel Handling Accident," December 1997.

18. SCALE Manual, NUREG/ CR-0200, Rev. 6, Volume 3, Section M8, "Standard Composition Library

19. Northeast Nuclear Energy Company Letter dated June 6, 1998, M. H.

Brothers (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No.

3, Proposed License Amendment Request, SLCRS Bypass Leakage, (PLAR 3-98-5)."

20. Northeast Nuclear Energy Company Letter dated April 5, 1999, Raymond P. Necci (NNECo) to USNRC, 'Millstone Nuclear Power Station, Unit No.

3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5), Response to Request for Additional Information."

21. Northeast Nuclear Energy Company Letter dated April 7, 2000, Raymond P. Necci (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No.

3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5), Supplemental Information."

22. Northeast Nuclear Energy Company Letter dated April 9, 2000, M.H.

Brothers (NNECo) to USNRC, 'Millstone Nuclear Power Station, Unit No.

3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5), Supplemental Information."

23. Northeast Nuclear Energy Company Letter dated July 31, 2000, Raymond P. Necci (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No.

Serial No.04-285 Discussion of Changes Page 88 of 98 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5), Supplemental Information."

24. Northeast Nuclear Energy Company Letter dated September 28, 2000, Raymond P. Necci (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No. 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5),

Supplemental Information."

25. Northeast Nuclear Energy Company Letter dated March 19, 2001, Raymond P. Necci (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No. 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5),

Supplemental Information."

26. Dominion Nuclear Connecticut, Inc. Letter dated June 11, 2001, Raymond P. Necci (DNC) to USNRC, 'Millstone Nuclear Power Station, Unit No. 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5),

Clarification of Proprietary Information."

27. Dominion Nuclear Connecticut, Inc. Letter dated September 21, 2001, J.

Alan Price (DNC) to USNRC, "Millstone Power Station, Unit No. 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5)."

28. Dominion Nuclear Connecticut, Inc. Letter dated December 20, 2001, Raymond P. Necci (DNC) to USNRC, "Millstone Nuclear Power Station, Unit No. 3, License Amendment Related to the Supplementary Leakage Collection and Release System (PLAR 3-98-5),

Supplemental Information."

29. USNRC Letter dated November 4, 1999, John A. Nakoski (USNRC) to R.P. Necci (NNECo), Millstone Nuclear Power Station, Unit No. 3 -

Issuance of Amendment Re: Reactor Water Storage Tank Back Leakage (TAC No. MA1749)."

30. Northeast Nuclear Energy Company Letter dated May 7, 1998, M. H.

Brothers (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit No.

3, Proposed License Amendment Request, Refueling Water Storage Tank Back Leakage (PLAR 3-98-3)."

31. Northeast Nuclear Energy Company Letter dated January 22, 1999, Martin L. Bowling (NNECo) to USNRC, "Millstone Nuclear Power Station, Unit

Serial No.04-285 Discussion of Changes Page 89 of 98 No. 3, License Amendment Request, Refueling Water Storage Tank Back Leakage (PLAR 3-98-3), Request for Additional Information."

32. NSAL-00-004, Westinghouse Nuclear Safety Advisory Letter dated March 7, 2000,, Non-conservatisms in Iodine Spiking Calculations.

33.Westinghouse Report WCAP-13132, 'The Effect of Steam Generator Tube Uncovery on Radioiodine Release," January 1992. (Proprietary)

34. Dominion Nuclear Connecticut Letter dated March 4, 2003, J. Alan Price (DNC) to USNRC, "Millstone Power Station, Unit No. 3, Licensing Basis Document Change Request (LBDCR) 3-01-03, Selective Implementation of the Alternative Source Term - Fuel Handling Accident Analysis."

35.V. Nerses, USNRC to D. Christian, "Millstone Station Unit No. 3 -

Issuance of Amendment Re: Selective Implementation of Alternative Source Term (TAC No. MB8137), dated March 17, 2004.

36. Regulatory Guide 1.145, Revision 01, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," February 1983.

37.SECY-98-154, "Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors," June 30,1998.

38. "Use of Alternative Source Terms at Operating Reactors," Final Rule, in Federal Register No. 64, p. 71990, December 23,1999.

39.SECY-99-240, 'Final Amendment to 10 CFR Parts 21, 50, and 54 and Availability for Public Comment of Draft Regulatory Guide DG-1081 and Draft Standard Review Plan Section 15.0.1 Regarding Use of Alternative Source Terms at Operating Reactors," October 5,1999.

40.NUREG-0933, -A Prioritization of Generic Safety Issues," Supplement 25, June 2001.

Serial No.04-285 Discussion of Changes Page 90 of 98 7.0 Technical Specification And Bases Change The following Technical Specifications for Millstone Unit 3 are revised as noted below to reflect implementation of the NUREG-1465 alternative source term (AST) as the Design Basis Source Term. The AST implementation analyses provide justification for the following changes to the Millstone Unit 3 Technical Specifications and Technical Specification Bases:

a. The definition of Dose Equivalent 1-131 in Section 1.10 of the Technical Specifications Definitions is revised to reference Federal Guidance Report No. 11 (FGR 11),

"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989, as the source of thyroid dose conversion factors (Reference 8). The AST implementation analyses, as described in Section 3, use the thyroid dose conversion factors listed in Federal Guidance Report No. 11 (FGR 11) instead of those listed in NRC Regulatory Guide 1.109, Revision 1.

b. Change Technical Specification 3/4.7.7, "Control Room Emergency Air Filtration System," Surveillance Requirements c.2 and d to reflect a methyl iodide penetration less than or equal to 5% for the Control Room Emergency Air Filtration System filters instead of 2.5%.

The AST implementation analyses assumed Charcoal Filter Efficiencies (%) for Control Room Filtered Recirculation and Intake as 90 aerosol, 90 elemental, and 70 organic.

Millstone Unit 3 adopts ASTM D3803-89 as the standard test method for Nuclear-Grade activated carbon and to determine the acceptance criteria of methyl iodide penetration.

The allowable penetration in accordance with ASTM D3803-89 corresponding to 90% methyl iodide efficiency for charcoal credited in Millstone Unit 3 AST implementation analyses is 5% using a safety factor of 2.

c. Delete Technical Specification 3/4.7.8, "Control Room Envelope Pressurization System."

The Control Room Envelope Pressurization System is no longer

Serial No.04-285 Discussion of Changes Page 91 of 98 credited in the accident analyses described in the AST implementation analyses. In accordance with AST implementation analyses, the requirements contained in this Specification do not meet any of 10 CFR 50.36(c)(2)(ii) criteria on items for which Technical Specifications must be established. This can be justified as follows:

Justification:

The Control Room Envelope Pressurization System provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity.

The Control Room Envelope Pressurization System consists of two banks of air bottles with its associated piping, instrumentation, and controls.

Each bank is capable of providing the control room area with one-hour of air following any event with the potential for radioactive releases.

During normal operations, the Control Room Envelope Pressurization System is required to be on standby. The Control Room Envelope Pressurization System is required to operate during post accident operations to ensure the control room will remain habitable during and following accident conditions.

Technical Specification 3/4.7.8 provide operability requirement, associated actions and surveillance requirement for the Control Room Envelope Pressurization System. The Control Room Envelope Pressurization System is no longer credited in the accident analyses described in the AST implementation analyses as described in Section 3.

Additionally, this specification does not meet any of the criteria of 10 CFR 50.36(c)(2)(ii).

10 CFR 50.36(c)(2)(ii) contains the requirements for items that must be in Technical Specifications. This regulation provides four (4) criteria that can be used to determine the requirements that must be included in the Technical Specifications.

Serial No.04-285 Discussion of Changes Page 92 of 98 Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

This Specification provides for the criteria used in determining operability of the Control Room Envelope Pressurization System. This specification does not cover installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. Therefore this specification does not satisfy criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

This Specification provides for the criteria used in determining operability of Control Room Envelope Pressurization System. This specification does not cover a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore this specification does not satisfy Criterion 2.

Criterion 3 A System, Structure, or Component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Serial No.04-285 Discussion of Changes Page 93 of 98 This Specification requires Control Room Envelope Pressurization System to be OPERABLE in MODES 1 through 6, and during fuel movement within containment or spent fuel pool.

The AST implementation analyses, as described in Section 3, do not assume the Control Room Envelope Pressurization System available in these analyses.

This assumption will provide the basis for removing Technical Specification 3/4.7.8 because it will no longer be credited in the accident analysis. Therefore, this feature does not cover a System, Structure, or component that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This Specification does not satisfy Criterion 3.

Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The Specification, which provides the criteria used in determining operability of the Control Room Envelope Pressurization System, has not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. The subject system is not credited to ensure radiological dose criteria for the EAB, LPZ, or control room is met. With the changes proposed in this license amendment request, this requirement no longer covers a SSC which requires risk review/unavailability monitoring.

This Specification does not satisfy Criterion 4.

In conclusion, the proposed changes to this specification (3/4.7.8) do not cover plant equipment which is credited to function in the event of a DBA.

Additionally, the requirements contained in this Specification do not meet any of 10 CFR 50.36(c)(2)(ii) criteria regarding items for which Technical Specifications must be established. Therefore, the proposed change to delete Technical Specification 3/4.7.8 is consistent with regulation and is safe.

Serial No.04-285 Discussion of Changes Page 94 of 98

d. Change the leakage rate acceptance criteria for all penetrations that are secondary containment bypass leakage paths in Section 6.8.4.f, "Containment Leakage Testing Program," from s 0.042 La to s 0.06 La.

The AST implementation analyses, as described in Section 3, assume leakage rate acceptance criteria for all penetrations that are secondary containment bypass leakage paths to be 5 0.06 La.

e. Index pages x and xiv are revised to reflect the deletion of Technical Specification 3/4.7.8 and the corresponding Bases.

The associated Bases changes are provided for information only. The Technical Specification Bases will be revised in accordance with the Technical Specification Bases Control Program (Section 6.18), following approval of the AST license amendment.

7.1 Specific Technical Specification Changes In this section deleted text is omitted and inserted text is underlined in the To portion of each revision. The Bases changes are included with each technical specification that is changed.

7.1.1 Definitions Revise the current Definition of DOSE EQUIVALENT 1-131 from:

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I."

Serial No.04-285 Discussion of Changes Page 95 of 98 To:

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-13 1, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance No. 11. "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

7.1.2 Technical Specification 3/4.7.7, "Control Room Emergency Air Filtration System" Revise Surveillance Requirement c.2 from:

Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 300C (860F), a relative humidity of 70%, and a face velocity of 54 ft/min; and To:

Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 5.0% when tested in accordance with ASTM D3803-89 at a temperature of 300C (860F), a relative humidity of 70%, and a face velocity of 54 ft/min; and

Serial No.04-285 Discussion of Changes Page 96 of 98 Revise Surveillance Requirement d from:

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal absorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 300C (860F), and a relative humidity of 70%, and a face velocity of 54 ft/min.

To:

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal absorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 5.0% when tested in accordance with ASTM D3803-89 at a temperature of 300C (860F), and a relative humidity of 70%, and a face velocity of 54 ft/min.

7.1.3 Bases 3/4.7.7, "Control Room Emergency Air Filtration System" Revise Background, Post Accident Operation, Item 2 from Delete items 2. and 3., and renumber item 4. as 2.

Revise Background, Applicable Safety Analysis Delete "except a Fuel Handling Accident, the radiation exposure to personnel occupying the control room shall be 5 rem or less whole body, or its equivalent for the duration of the accident, consistent with the requirements of General Design Criterion 19 of Appendix "A,"10 CFR 50.

For a Fuel Handling Accident."

Serial No.04-285 Discussion of Changes Page 97 of 98 Revise 4.7.7.e.2, first paragraph on Page B 3/4 7-16 from During the first hour, the control room pressurization system creates and maintains the positive pressure in the control room. This capability is verified by Surveillance Requirement 4.7.8.C, independent of Surveillance Requirement 4.7.7.e.2. A CBI signal will automatically align an operating filtration system into the recirculation mode of operation due to the isolation of the air supply line to the filter.

To:

A CBI signal will automatically align an operating filtration system into the recirculation mode of operation due to the isolation of the air supply line to the filter.

7.1.4 Technical Specification 3/4.7.8, "Control Room Envelope Pressurization System" Delete Technical Specification 3/4.7.8. The text in Technical Specification pages 3/4 7-18 and 3/4 7-19 is replaced with THIS PAGE INTENTIONALLY LEFT BLANK 7.1.5 Bases 3/4.7.8, "Control Room Envelope Pressurization System" Delete Bases 3/4.7.8. The text in Bases pages B 3/4 7-17, 7-18, 7-19, 7-20, 7-20a, 7-21 and 7-22 will be replaced with THIS PAGE INTENTIONALLY LEFT BLANK 7.1.6 Section 6.8.4.f, "Containment Leakage Testing Program" Revise Section f.1 from

Serial No.04-285 Discussion of Changes Page 98 of 98 Containment overall leakage rate acceptance criterion is s 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and s 0.042 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; To:

Containment overall leakage rate acceptance criterion is s 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and 5 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; 7.1.7 Additional Bases Changes

1. Bases 3/4.6.1.1, "Containment Integrity" Change "10 CFR Part 100" to "10 CFR 50.67", and "GDC 19" to "Regulatory Guide 1.183."

2. Bases 3/4.6.6.2, "Secondary Containment" Change "10 CFR Part 100" to "10 CFR 50.67."

3. Bases 3/4.7.1.4, "Specific Activity" Replace "a small fraction of 10 CFR Part 100" with "10 CFR 50.67 and Regulatory Guide 1.183."

4. Bases 3/4.9.10 and 3/4.9.11 "Water Level -

Reactor Vessel and Storage Pool" Add "at least' and delete '1O%."

ENCLOSURE 1 to ATTACHMENT 1 ELECTRONIC COPY OF SITE METEOROLOGICAL DATA TAKEN OVER THE YEARS 1997-2001 and CALCULATION OF ONSITE X/Q VALUES TO THE MILLSTONE UNIT 3 CONTROL ROOM USING ARCON96 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM MARK-UP PAGES OF TECHNICAL SPECIFICATIONS CHANGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

JGnuar, 2, 2OO3t INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 TABLE 4.7-1 3/4.7.2 3/4.7.3 3/4.7.4 3/4.7.5 3/4.7.6 3/4.7.7 3/4.7.8 3/4.7.9 3/4.7.10 TABLE 4.7-2 FIGURE 4.7-1 3/4.7. 11 3/4.7.12 Table 3.7-4 Table 3.7-5 3/4.7.13 3/4.7.14 TABLE 3.7-6 STEAM LINE SAFETY VALVES PER LOOP Auxiliary Feedwater System.............

Demineralized Water Storage Tank..........

Specific Activity SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.................

Main Steam Line Isolation Valves...

Steam Generator Atmospheric Relief Bypass Lines DELETED REACTOR PLANT COMPONENT COOLING WATER SYSTEM..

SERVICE WATER SYSTEM................

ULTIMATE HEAT SINK.................

DELETED CONTROL ROOM EMERGENCY VENTILATION SYSTEM AUXILIARY BUILDING FILTER SYSTEM..........

SNUBBERS......................

SNUBBER VISUAL INSPECTION INTERVAL.........

SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.....

DELETED DELETED DELETED DELETED DELETED AREA TEMPERATURE MONITORING AREA TEMPERATURE MONITORING 3/4 7-3 3/4 7-4 3/4 7-6 3/4 7-7 3/4 7-8 3/4 7-9 3/4 7-9a 3/4 7-10 3/4 7-11 3/4 7-12 3/4 7-13 3/4 7-14 3/4 7-15 3/4 7-18 3/4 7-20 3/4 7-22 3/4 7-27 3/4 7-29 3/4 7-30 3/4 7-32 3/4 7-33 MILLSTONE - UNIT 3 x

Amendment No. !A

9pR, Af, ml$

JM=

Daber-1% 26O3'L BASES TABLE B3/4.4-1 REACTOR VESSEL FRACTURE TOUGN PROPERTIES.

B 3/4 4-9 FIGURE B 3/4.4-1 FAONSFLUEN>EMeV)WAS AFUCTIONOFB FUI1 PO tR.

SERVICE

.....................4...................... B 314 4-10 3/4.4.10 DELETED............

B 3/44-15 314.4.11 DEIEFED.................

B 3/4 4-l5 3t4.S CELNG SYSTM 314.5.1 A C IMUL A O S......................................

3/4.52 and 3/4.53 ECCSSUBSYSnh{...................

3W4.5A 1REF[EING WATERt STOkA& TANK.....................

3/455 pH TRISODIUMPHOSPHAI STORAGEBASKET

.:. !.. 4

_3/4S-1 3J4 S5-2

.................................... B 3t4 S-4 I-.1**

3J4.6.1 3t4.6.2 314.63 314.6.4 314.6.5 314.6.6 I[NTSYSMS PRIM RY CONTANM T._._....................................... __

DEPRESSURIZATION AND COOLING SYSTEM...

CONTAINMENT ISOLAION VALVES COMBUSTIBLE GAS CONTROL..

SUBATMOSPHERIC PRESSURE CONTROL SYSTEM SECONDARY CONTAINM ENT....

...._E I......................

i 3 3/4 6-1 3 3/4 6-2 3 3/4 6-3

,............_...... B 3/4 6-3d

....... B 34 6-4 3/47 PLANT SYSTEMS 3/4.7.1 TllRBINE CYCLE B 3/4 7-1 3/4.72 DELEED.

B..

3/4 7-7..

3/4.7.3 REACTOR PLANT COMPONENT COOLING WAER SYSTEM.

.B 3/4 7-7 3/4.7A4 SERVICE WAER SYSTEM B 3/4 7-7.

3/4.75 ULTIM E lEAl SINK.

B 3/4 7-8 3/4.7.6 D.~..4.

.4___......

dB 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................................. _ 3/4 7-10 3/4.7.8 B 3/4 71 7.,.,,,

.B 3/ 7-17 OEL-01eim 3/4.7.9 A.........YB..ER.SYSTEM.B 3/4 7-23 3/4.7.10 SNUBBERS..............

B 314 7-23 MILLSTONE -UNIT 3 liv Anmdmt No. 49,.,

14-5 449, 4M, 04,6.M, 44,26,DCIL-

-O2

-be3-0

DEFINMONS Septonber 29,2003 9

1.7 CONTARWENT INTEGRITY shall aeistwen:

a.

All penetrations required to be closed during accident conditions are either:

l.

Capable of being closed by an OPERABLE containment automatic isolation valve system* or

~-

7.

laO9dby irnVlves bliffllange, ofdt i their closed positions, except for vahles that are opened under adm ive.

control as permitted by Specification 3.63.

b.

All equipment hatches are closed and scaled,

c.

Each air lock is in compliance with the requiroments of Spectification 3.6.1.3,

d.

,The contaimnt leakage t

rates are whithe mitsof the Contnment Leakage Rate Testing Program, and

' e; T

- Ih. e sealingmechanism associated with each penetration (e.g., welds, bellows, or 0ings) is 0P~ABl^E.'

CONMROLLED LEAKAGME 1.8 CONTROLLEDLEAKAGE shall bethat sealwaterflow.pliedto theeactor coolantpupseals.

C OEALTEM4TIONS 1.9 CORE ALTERAIIONS shall be the movement of any fuel, sources, reactivity controlcomponents, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel Suspension of CORE ALTERTONS shall not precl4de completion of movement o a component to a safe positio DOSED HOTJIVAENT 1-131' 1.10 DOSE EQUIVALENT 1-131 sball be that cncentton of 1-131 (microCurietgram) which *lne would produce the same thyroid dose as the quantity and isotopimixture of 1-131,1-132,1-133, 1-134,'and I-135 c le sent. The thyroid dose

-en on factor used for this calculation shall be those listed in diem^bV

_,>.6 of E"vef COPM

." lo w,. 22

Rv wsv I n MODE 4, e qurment for an OPERABLE contamment isolation valve system is satisfied by use of the containent isolation actuation pushbuttons.

/

hflLLSTONE - UNIT 3 1-2 Amendment No. 28, 40, 46,*,

Insert A to Page 1-2 Federal Guidance No. 11, 'Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

ma

^^rw

_sto__Xr PLANT SYTEMS ho r Zk{r-+

Ag~,aJ10 f 314.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM February 20, 2002 LIMITING CONDITION FOR OPERATION 3.7.7 Two independent Control Room Emergency Air Filtration Systems shall be OPERABLE.* #

APPLICABILITY:

MODES 1, 2, 3, 4, 5, and 6.

During fuel movement within. containment or the spent.fuel..

pool.'

ACTION:

MODES 1, 2, 3 and 4:

a. With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the -next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within -the following. 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s-.

b. With both Control. Room. Emergency Air Filtration. Systems inoperable, except as specified:-in.'ACTION c.j immediately-suspend the-movement of fuel assemblies-within the spent fuel pool-.

Restore at least one inoperable system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C. With both Control Room Emergency Air Filtration Systems inoperable due to-an inoperable-Control Room boundary, immediately suspend the movement of fuel assemblies within the spent-fuel pool and restore the Control Room boundary to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6, and fuel movement within containment or the spent fuel pool:

d. With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days.

After 7 days, either initiate and maintain operation of the remaining OPERABLE Control Room Emergency Air Filtration System in the recirculation. mode of operation, or immediately suspend CORE ALTERATIONS and the movement of fuel assemblies.

e. With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency:. Air Filtration System required to be in the'recirculation'mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, immediately suspend CORE ALTERATIONS and the movement of fuel assemblies.

The requirements of Surveillance Requirement 4.7.7.e.2 do not apply during pressure testing of the Cable Spreading Room. This exception is valid until the first entry intoMODE 4 following the completion of refueling operations associated with the seventh Refueling Outage.

The Control Room boundary may be opened intermittently under administrative control.

MILLSTONE -,UNIT 3 3/4 7-15 Amendment ND. A, IFJ, 203

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is less than or equal to 95F;

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA fiilters and charcoal adsorbers and verifying a system flow rate of 1,120 cfm*+20% and that the system operates for at least 10 continuous hours with the heaters operating;

c.

At least once per 24 months or (1) after any structural maintenance on the HEPA filter of charcoal adsorber housings, or (2)following 'painting, fire, or chemical.release.in 'any.

ventilation zone communicating'with'the system by:-

1)

Verifying that the system satisfies the in-place penetration and bypass leakage.

testing acceptance criteria of less than 0.05% and-uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the system flow rate is 1,120 cfm +/-20%;

2)

Verifying., within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulator-OD Guide 1.52, Revision 2, March 1978,* showst eityl iodide penetration less than or equal tongs when tested in accordance with ASTM D3803-89 at a temperature of 30*C (86'F), a relative humidity of 70%, and a face velocity of 54 ft/min; and

3)

Verifying a system flow rate of 1,120 cfm +/-20% during system operation when tested in accordance with ANSI

N510-1980.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of. a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, 'March 1978,* s penetration less than'or equal to when tested in accordance with ASTM D3803-89 at a temperature of 30C (86F), and a relative humidity'of 70%, and a face velocity of 54 ft/min.

e.

At least once per 24 months by:

1)

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm +20%;

MILLSTONE - UNIT 3 3/4 7-16 Amendment No. A, Z

, Jp, 7p}. t#

PLANT SYSTEMS Fbr O4jl o94 February 20, 2002 SURVEILLANCE REQUIREMENTS (Continued)

2)

Verifying.-,that the system maintains the control room at a.

positive pressure of greater than or equal to.1/8 inch Water Gauge at less than or equal to a pressurization flow of 23O0 cfm relative to adjacent areas and outside atmosphere during positive pressure system.operation; and.

3)

Verifying that.the heaters dissipate 9.4 +/-l1 kW when tested in accordance with ANSI N510-1980.-

f.

After each complete or partial replacement of-a IEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than. 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system'at a flow rate of 1120 cfm +20%; and g..

After each -complete or partial replacement of a charcoal adsorber bank, by verifying that the :cleanup.system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1120 cfm +20%.

', I

ANSI N510-1980 shall be used in place of ANSI NS1O-1975 referenced in Guide 1.52, Revision 2, March 1978.

Regulatory MILLSTONE - UNIT 3 0832 3/4 7-17 Amendment No. Z, 711, IFI, 203

DI AUX CYCT&MCIi 3 4.7.8 CONTROL ROOM ENVELOPE PRESSURI TION SYSTEM February 20, 2002 LIMITING COND ION FOR OPERATION 3.7.8 Two indepn'dent Control Room Envel'op Pressurization Systems shall be OPERABLE.*

APPLICABILITY:

MODES

, 2 3, 4, 5, and 6.

During uei.movement within containm t or the spent fuel pool...

ACTION:\\

MODES 1, 2, 3, and 4-

a.

With one Control Roo Envelope Pressurization Syst inoperable restore i

the system to OPERABLE status within 7 days or be IPHOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and CNLD SHUTDOWN within the folio ng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With. both Control Room \\nveldpe Pressurization Syst s inoperable, except as specified in ACTION c. or ACTION d., immediate suspend the w

movement of fuel assemblies within the spent fuel pool.

Restore at least-one inoplerable system OPERABLE status within.1 ho or be in HOT STANDBY within the next hours and COLD SHUTDOWN w thin the u

f ollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

W h both Control Room Envelope Pr ssurization Systems inoperab due to an inoperable Control Room boundary, immediately suspend the move ent of fuel assemblies within the spent fuel Pool.

Restore e

Contr Room boundary to-OPERABLE stat s within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H STANDB within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD HUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

With both ontrol Room Envelope Pressuri tion Systems inoperable i

during the rformance of Surveillance Requ'rement 4.7.8.c and the system not b ing tested under administrative control, immediately suspend the mo ment of fuel assemblies within the spent fuel pool.

Restore at leas one inoperable system to OPERLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in HO STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the followi 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and, and fuel movement

'thin containment or the spent f I pool:

e.

Wi one Control Robm nvelope Pressurization System inoperable, res re the inoperable sy em to OPERABLE status within 7 d s. After 7 day immediately suspen CORE ALTERATIONS and the moveme of fuel assemb es.

f.

With both Control Room Envel e Pressurization Systems inope ble, immediatel suspend CORE AL TIONS and the movement of uel assemblies.

The requirements Surveillance Requl ments 4.7.8.c.2 and 4.7.8.c.3 do not apply during ressure testing of t Cable Spreading Room. This exception Is valid til the first entr into MODE 4 following the completion of refue ing operations asso iated with the seventh Refueling Outage.

I The Control Room bounda may be opened intermittently under administrative control.

MILLSTONE - UNIT 3 3/4 7-18 Amendment No. REX7, Xp 0833

I, JuMy 24, 2002 PLANT SYSTEMS\\\\\\

SURVEILLANCE REQkRMENTS\\\\

4.7.8 Each Control 0oom Envelope Pressuriz ion System shall e.demonstrated PERABLE:

a.

At least on per 7 days by verifyl that the st ra air bottles are pressurizd to greater than or e al to 2200 pstg, At least once p 31 days on a STAGG.ERE TEST BASIS by v ifying that each valve anual, power operated a aUtomatic) In ti e flow path not locked, s aled or otherwise secur d in position, i in its correct positio and t least once per 24 mnths or following.a ma r alteration of. e c trol room envelope p ssure boundary by:

\\

1.

Verifying that the ntrol room envelope is solated -in

\\response to a'Control uilding Isolation.test ignal,

2.

V ifying that after a second time delay fol wing a Co rol Building Isolatia test signal, the contr I room env ope pressurizes to gr ater than or equal to 1 inch W.G..

lative to adjacent a as and outside atmosphe, and

3.

Verifyl that the positive pr ssure of Specific tion 4.7.8.c.2 is main ained for greater than equal to minutes.

Ree w;so "T1 s.s PA 5JXN1TbPALL L~9 T 6f~k I TE UI3 1-9 AmendmentPo.

MILLSTONE -UNIT 3

3/4 7-19 Amendment No.

Z.Z.

ADMINSTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and

3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

f. Containment Leakage Rate Testing Program, A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10.CFR 50, Appendix J, Option B, as modified by approved exemptions.. This program.shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,' dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 38.57 psig.

The maximum allowable containment leakage rate C., at P., shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Leakage rate acceptance criteria are:

1) Containment overall leakage rate acceptance criterion is

< 1.0 L,.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance a are < 0.60 L. for the combined Type B and type C tests, a L. for all penetrations that. are Secondary Containment bypass leakage paths, and <0.75 L. for Type A tests;

2) Air lock testing acceptance criteria are:.

a.

Overall air lock leakage rate is < 0.05 L. when tested at

> P.

b.

For each door, seal leakage rate is < 0.01 L, when pressurized to > Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified.in the Containment Leakage Rate Testing Program.

The provisions of -Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

An exemption to Appendix J, Option A, paragraph JII.D.2(b)(il), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.

MILLSTONE - UNIT 3 6-17 Amendment No. #i, #A,)

Serial No.04-285 Page 1 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM RETYPED PAGES OF TECHNICAL SPECIFICATIONS CHANGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP........................................ 314 7-3 Auxiliary Feedwater System...............

................................. 3/4 7-4 Demineralized Water Storage Tank..............................

/..................

34 7-6 Specific Activity................................................

314 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................................................ 3/4 7-8 Main Steam Line Isolation Valves..........................

...................... 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines................................ 3/4 7-9a 3/4.7.2 DELETED................................................................................................

3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM..... 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM....................

............................ 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK.................................................

3/4 7-13 3/4.7.6 DELETED................................................

3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM............

3/4 7-15 3/4.7.8 DELETED.................................................

3/47-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM......................................... 3/4 7-20 3/4.7.10 SNUBBERS................................................

3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL................................... 3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.................. 3/4 7-29 3/4.7.11 DELETED.................................................

3/4 7-30 3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING.............................

3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING.............................

3/4 7-33 MILLSTONE - UNIT 3 x

Amendment No. 62,84,400,460, 24

INDEX BASES SECTION PAGE TABLE B3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES.

B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>lMeV) AS A FUNCTION OF FULL POWER SERVICE LIFE...................................................

B 3/4 4-10 3/4.4.10 DELETED.....................................................

B 3/4 4-15 314.4.11 DELETED.....................................................

B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS..................................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...................................................

B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.................................................. B 3/4 5-2 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS............................. B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.

B 3/4 6-1 3/4.6.2 DEPRESS URIZATION AND COOLING SYSTEMS.B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES.

B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL.

B 3/4 6-3a 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM.B 3/4 6-3d 3/4.6.6 SECONDARY CONTAINMENT.B 3/4 6-4 314.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.B 3/4 7-1 3/4.7.2 DELETED.

B 3/4 7-7 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM.

B 3/4 7-7 3/4.7.4 SERVICE WATER SYSTEM.B 3/4 7-7 3/4.7.5 ULTIMATE HEAT SINK.B 3/4 7-8 3/4.7.6 DELETED.

B 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.B 3/4 7-10 3/4.7.8 DELETED.

B 3/4 7-17 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM.B 3/4 7-23 3/4.7.10 SNUBBERS.B 3/4 7-23 MILLSTONE - UNIT 3 xiv Amendment No. 4X, 9, 14,4419, 436, 204,26P, 244,246,

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation valve system*, or

2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b.

All equipment hatches are closed and sealed,

c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d.

The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and

e.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.

M[ILLSTONE - UNIT 3 1-2 Amendment No. 28,407, 4-86, 26

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is less than or equal to 950F;

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying a system flow rate of 1,120 cfm +/-20% and that the system operates for at least 10 continuous hours with the heaters operating;

c.

At least once per 24 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1.

Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the system flow rate is 1,120 cfm +/- 20%;

2.

Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 5.0% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (860F), a relative humidity of 70%, and a face velocity of 54 ft/min; and

3.

Verifying a system flow rate of 1,120 cfm +/- 20% during system operation when tested in accordance with ANSI N510-1980.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 5.0%

when tested in accordance with ASTM D3803-89 at a temperature of 30'C (860F),

and a relative humidity of 70%, and a face velocity of 54 ft/min.

e.

At least once per 24 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm +/- 20%;

MILLSTONE - UNIT 3 3/4 7-16 Amendment No.2, 423, 41, 44,203, 206

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-18 Amendment No. 48+, 2, 249

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-19 Amendment No. 423, 203,20

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

2.

Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and

3.

Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

f.

Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix 1, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pas is 38.57 psig.

The maximum allowable containment leakage rate Lap at Pa' shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Leakage rate acceptance criteria are:

1.

Containment overall leakage rate acceptance criterion is c 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests;

2.

Air lock testing acceptance criteria are:

a.

Overall air lock leakage rate is < 0.05 La when tested at 2 Pa.

b.

For each door, seal leakage rate is < 0.01 La when pressurized to 2 Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

An exemption to Appendix J, Option A, paragraph m.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.

MILLSTONE - UNIT 3 6-17 Amendment No. 69, 486

Serial No.04-285 Page 1 ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.04-285 No Significant Hazards Consideration Determination Page 1 of 2 Significant Hazards Consideration Determination This license amendment proposes full implementation of an alternative source term (AST) and changes to the Technical Specifications. Changes are proposed for the following Technical Specifications:

Definition of Dose Equivalent 1-131 - revised to allow use of Federal Guidance Report No. 11 (FGR 11) dose conversion factors

Technical Specification 3/4.7.7, Control Room Emergency Air Filtration System -

changed the value used for methyl iodide penetration test acceptance criteria.

Technical Specification 3/4.7.8, Control Room Envelope Pressurization System -

deleted the specification in its entirety.

Section 6.8.4.f, the leakage rate acceptance criteria for all penetrations that are secondary containment bypass leakage paths - changed the value used in the acceptance criteria.

We have reviewed the proposed Technical Specifications changes relative to the requirements of 10 CFR 50.92 and determined that a significant hazards consideration is not involved.

Specifically, operation of Millstone Power Station Unit 3 with the proposed changes will not:

1.

Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase in the probability or consequence of an accident previously analyzed. The Millstone Unit 3 Control Room Emergency Air Filtration System only functions following the initiation of a design basis radiological accident. Therefore, the change to the value used for methyl iodide penetration test acceptance criteria following a design basis accident will not increase the probability of any previously analyzed accident. The Millstone Unit 3 Control Room Envelope Pressurization System is no longer credited in the accident analyses described in the Alternative Source Term (AST) implementation analyses. In accordance with AST implementation analyses, the requirements contained in this Specification do not meet any of 10 CFR 50.36(c)(2)(ii) criteria on items for which Technical Specifications must be established.

Deletion of this Technical Specification will not increase the probability of occurence of any previously analyzed accident and does not impact the consequences of any evaluated accident since it is no longer analytically credited.

The Millstone Unit 3 containment and the containment systems function to prevent or control the release of radioactive fission products following a postulated accident. Therefore, the change to the value used for the leakage rate acceptance criteria for all penetrations that are secondary containment bypass leakage paths following a design basis accident will not increase the

Serial No.04-285 No Significant Hazards Consideration Determination Page 2 of 2 probability of any previously analyzed accident and is limited to ensure it does not increase any accident consequence.

These systems are not initiators of any design bases accident. Revised dose calculations, which take into account the changes proposed by this amendment and the use of the alternative source term, have been performed for the Millstone Unit 3 design basis radiological accidents. The results of these revised calculations indicate that public and control room doses will not exceed the limits specified in 10 CFR 50.67 and Regulatory Guide 1.183. There is not a significant increase in predicted dose consequences for any of the analyzed accidents.

Therefore, the proposed changes do not involve a significant increase in the consequences of any previously analyzed accident.

2.

Create the possibility of a new or different kind of accident from any accident previously evaluated.

The implementation of the proposed changes does not create the possibility of an accident of a different type than was previously evaluated in the UFSAR.

Although the proposed changes could affect the operation of the Control Room Emergency Air Filtration System, and containment and the containment systems following a design basis radiological accident, none of these changes can initiate a new or different kind of accident since they are only related to system capabilities that provide protection from accidents that have already occurred.

These changes do not alter the nature of events postulated in the UFSAR nor do they introduce any unique precursor mechanisms.

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

3.

Involve a significant reduction in the margin of safety.

The implementation of the proposed changes does not reduce the margin of safety. The proposed changes for the Control Room Emergency Air Filtration System, and containment and the containment systems do not affect the ability of these systems to perform their intended safety functions to maintain dose less than the required limits during design basis radiological events.

The revised dose calculations also indicate that the change to the containment depressurization times will continue to maintain the dose to the public and control room operators less than the required limits. The radiological analysis results, when compared with the revised TEDE acceptance criteria, meet the applicable limits.

These acceptance criteria have been developed for application to analyses performed with alternative source terms.

These acceptance criteria have been developed for the purpose of use in design basis accident analyses such that meeting the stated limits demonstrates adequate protection of public health and safety.

It is thus concluded that the margin of safety will not be reduced by the implementation of the changes.

ATTACHMENT 5 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM ENVIRONMENTAL IMPACT EVALUATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.04-285 Environmental Impact Determination Page 1 of 2 ENVIRONMENTAL IMPACT EVALUATION 10 CFR51.22(c)(9) provides criteria for and identification of licensing and regulatory action eligible for categorical exclusion from performing an environmental assessment.

A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:

1) involve a significant hazards consideration,

2) result in a significant change in the type or a significant increase in the amounts of any effluents that may be released offsite, or

3) result in a significant increase in individual or cumulative occupational exposure.

Dominion Nuclear Conecticut, Inc. (DNC) has reviewed this license amendment and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 52.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment. The basis for this determination is as follows:

1) The proposed license amendment does not involve a significant hazards consideration as described previously in Attachment 4 of this letter.

2) As discussed in the significant hazards evaluation, the changes proposed by this amendment and full implementation of an alternative source term do not result in a significant change or significant increase in the public dose consequences for Millstone Unit 3 design basis radiological accidents. Approval of a new alternative source term for Millstone Unit 3 establishes a new licensing and design basis for assessment of accident consequences.

It does not change actual accident sequences; only the regulatory assumptions regarding radiological accidents change. The adoption of an alternative source term, by itself, will not result in plant changes that involve any significant increase in environmental impacts.

The proposed changes affect the operation of the Control Room Emergency Air Filtration System during radiological accidents, the requirement for a Control Room Envelope Pressurization System, and the acceptance criteria for the Containment Leakage Testing Program.

These systems do not interface with any plant system that is involved in the generation or processing of effluents during normal plant operations.

The proposed changes will affect the radioactive effluents during a radiological accident. However, the dose to the public will not exceed the limits specified in 10 CFR 50.67 and Regulatory Guide 1.183.

Therefore, implementation of the proposed change and a full alternative source term will not result in a significant change in the types or increase in the amount of any effluents that may be released offsite.

3) The changes proposed by this amendment and full implementation of an alternative source term do not result in a significant increase in control room operator doses during design basis radiological accidents. In addition, the proposed changes do not

Serial No.04-285 Environmental Impact Determination Page 2 of 2 require operators or other actions that could increase occupational radiation exposure.

The proposed changes will affect the radioactive effluents during a radiological accident. However, the dose to the operator will not exceed the limits specified in 10 CFR 50.67 and GDC-19. Therefore, the proposed amendments and implementation of an alternative source term will not result in a significant increase in individual or cumulative occupational radiation exposure.

ATTACHMENT 6 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM MARKED-UP PAGES OF TECHNICAL SPECIFICATION BASES (FOR INFORMATION ONLY)

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

December 18,2003 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR 5o64.7 Qg accident conditions and the control room operators dose to within the guidelinesof~

J XRe.tIe s G xCU 1.Ipi.

Primary CONTAINMENT INTEGRITY is required in MODES 1 through 4. This requires an OPERABLE containment automatic isolation valve system. In MODES 1, 2 and 3 this is satisfied by the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure. In MODE 4 the automatic containment isolation signals generated by high containment pressure, low pressurizer pressure and low steamline pressure are not required to be OPERABLE. Automatic actuation of the containment isolation system in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. Since the manual actuation pushbuttons portion of the containment isolation system is required to be OPERABLE in MODE 4, the plant operators can use the manual pushbuttons to rapidly postion all automatic containment isolation valves to the required accident position. Therefore, the containment isolation actuation pushbuttons satisfy the requirement for an OPERABLE containment automatic isolation valve system in MODE 4.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates, as specified in the Containment Leakage Rate Testing Program, ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than 0.75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The Limiting Condition for Operation defines the limitations on containment leakage.

The leakage rates are verified by surveillance testing as specified in the Containment Leakage Rate Testing Program, in accordance with the requirements of Appendix J. Although the LCO specifies the leakage rates at accident pressure, Pa, it is not feasible to perform a test at such an exact value for pressure. Consequently, the surveillance testing is performed at a pressure greater than or equal to Pa to account for test instrument uncertainties and stabilization changes. This conservative test pressure ensures that the measured leakage rates MILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. 59, 89, 44, *44, 46, iBDCP 03-N 24-60

Febrm7-,199G CONTAINMENT SYSTEMS BASES 3/4.6.6.2 SECONDARY CONTAINMENT The Secondary Containment is comprised of the containment enclosure building and all contiguous buildings (main steam valve building [partially], engineering safety features building

[partially], hydrogen recombiner building [partially], and auxiliary building). The Secondary Containment shall exist when:

a.

Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit,

b.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

Secondary Containment ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with operation of the Supplementary Leak Collection and Release System, and Auxiliary Building Filter System will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR 65.6 7 TEiuring accident conditions.

The SLCRS and the ABF fans and filtration units are located in the auxiliary building.

The SLCRS is described in the Millstone Unit No. 3 FSAR, Section 6.2.3.

In order to ensure a negative pressure in all areas within the Secondary Containment under most meteorological conditions, the negative pressure acceptance criterion at the measured location (i.e., 24'6" elevation in the auxiliary building) is 0.4 inches water gauge.

LCO The Secondary Containment OPERABILITY must be maintained to ensure proper operation of the SLCRS and the auxiliary building filter system and to limit radioactive leakage from the containment to those paths and leakage rates assumed in the accident analyses.

Applicability Maintaining Secondary Containment OPERABILITY prevents leakage of radioactive material from the Secondary Containment. Radioactive material may enter the Secondary Containment from the containment following a LOCA. Therefore, Secondary Containment is required in MODES 1, 2, 3, and 4 when a design basis accident such as a LOCA could release radioactive material to the containment atmosphere.

MILLSTONE - UNIT 3 B 314 6-7 Amendment No. 8U, (tJ')

Juo *3, ^03 --

PLANT SYSTEMS BASES 3/4.7.1.3 DEMINERALIZED WATER STORAGE TANK (Continued)

If the combined condensate storage tank (CST) and DWST inventory is being credited, there are 50,000 gallons of unusable CST inventory due to tank discharge line location, other physical characteristics, level measurement uncertainty and potential measurement bias error due to the CST nitrogen blanket. To obtain the Surveillance Requirement 4.7.1.3.2's DWST and CST combined volume, this 50,000 gallons of unusable CST inventory has been added to the 334,000 gallon DWST water volume specified in LCO 3.7.1.3 resulting in a 384,000 gallons requirement (334,000 + 50,000 = 384,000 gallons).

3/4.7.1.4 SPECIFIC ACTIVITY sA epie SO.

7 a-i Re;r A

1-iiP3 The limitations on Secondary CoolanSstem activi ensure that the resultant offsite radiation dose will be limited to of 10 CFR Part 1 ose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident I gpm primary-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

MILLSTONE - UNIT 3 B 3/4 7-2b Amendment No. 402, 49, 450, "Revised by NRC Letaf A.15'10",

PLANT SYSTEMS January 2, 2003 BASES SURVEILLANCE REQUIREMENTS For T 0G*s<4-b Col4j For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section.

Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature.

The 24-hour frequency is based on operating experience related to trending of the parameter variations during the applicable modes. This surveillance requirement verifies that the average water temperature of the UHS is less than or equal to 75-F.

Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70F during the applicable modes. The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70F. The frequency is based on operating experience related to trending of the parameter variations during the applicable modes.

3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity.

Additionally, the system provides temperature control for the control room during normal and post-accident operations.

The control room emergency ventilation system is comprised of the control room emergency air filtration system and a temperature control system.

The control room emergency air filtration system consists of two-redundant systems that recirculate and filter the control room air. Each control room emergency air filtration system consists of a moisture separator, electric heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan.

Additionally, ductwork, valves or dampers, and instrumentation form part of the system.

Normal Operation A portion of the control room emergency ventilation system is required to operate during normal operations to ensure the temperature of the control room is maintained at or below 95F.

MILLSTONE - UNIT 3 B 3/4 7-10 Amendment No. 777, Of, 7i, 214

ELANTSYSIEMS BASES 314.7.7 CONTROL ROOM EMERGENCY VENTELATION SYSTEM (Continued)

BACKGROUND (Continued)

Post Accident Operation The control room emergency ventilation system is required to operate during post-accident operations to ensure the temperature of the control room is maintained and to ensure the control room will remain habitable during and following accident conditions.

The following sequence of events occurs upon receipt of a control building isolation (CBI) signal or a signal indicating high radiation in the air supply duct to the control room envelope.

1.

The control room boundary is isolated to prevent outside air from entering the control room to prevent the operators from being exposed to the radiological conditions that may exist outside the control room. The analysis for a loss of coolant accident assumes that the highest releases occur in the first hour after a loss of coolant accident 60 seconds, the control room envelope pressurizes to 1/8 inch water gauge A

control room emergency pressurization system. This action provides a continuous purge of the control room envelope and prevents ineakage from the outside environment.

ol Technical Specification 3/4.7.8 provides the requirements for the control room envel presuriatin sstem.

\\ 3. Contol rom pessrization continues for the first hour.

After one hour, the control room emergency ventilation system will be placed in service in fi either the 100% recirculation mode (isolated from the outside environment) or filtered pressurization mode (outside air is diverted through the filters to the control room envelope to maintain a positive pressure). The mode of service for the filtration will be based on the radiological conditions that exist outside the control room. To run the control room emergency air filtration system in the filtered pressurization mode, the air supply line must be manually opened.

APPLICABLE SAFETY ANALYSIS The OPERABILITY of the Control Room Emergency Ventilation System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for a the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The b OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to Innel occuRYing the control room.For all postuled

\\

desg basis acciden xcet a Rel HandligAcdnte rdito exouet esn f q~I room shall be 5 rem or less whole body, or its equivalent for the duration of

\\

l te acidntconsistent with the requiremen!of(e~neral Design Criterion -19 of Apni A"

\\ 0CR5.For a Fuel Handling Accidendthe radiatinexpsr to perone mshall be rem IED-gor less, consistent with the requirements of 10 CFR 50.67.

This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.

MILLSTONE - UNIT 3 B 314 7-11 Amendment No. 4A6, t0

-PLANT SYSTEMS BASES 3477 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

LIMlNG CON ON FOR OPERATION Two independent control room emergency air filtration systems are required to be operable to ensure that at least one is available in the event the other system is disabled.

A control room emergency air filtration system is OPERABLE when the associated:

a.

Fan is OPERABLE;

b.

HEPA filters and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions; and

c.

moisture separator, heater, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

The integrity of the control room habitability boundary (i.e., walls, floors, ceilings, ductwork, and access doors) must be maintained such that the control building habitability zone can be maintained at its design positive pressure if required to be aligned in the filtration pressurization mode. However, the LCO is modified by a footnote allowing the control room boundary to be opened intermittently under administrative controls. For entry and exit through doors the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in constant communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated.

APPLICABILITY In MODES 1, 2, 3,4, 5, and 6.

During fuel movement within containment or the spent fuel pool.

Actions a., b., and c. of this specification are applicable at all times during plant operation in MODES 1, 2, 3, and 4. Actions d. and e. are applicable in MODES 5 and 6, and whenever fuel is being moved within containment or the spent fuel pool. The fuel handling accident analyses assume that during a fuel handling accident some of the fuel that is dropped and some of the fuel impacted upon Is damaged. Therefore, the movement of either new or irradiated fuel (assemblies or individual fuel rods) can cause a fuel handling accident, and this specification is applicable whenever new or irradiated fuel is moved within the containment or the storage pool.

MUIMSONE - UNIT 3 B 3/4 7-12 Amendment No. -36, I, 219

PLANT SYSTEMS BASES Fo y,

°4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

ACTIONS Modes 1. 2. 3. and 4

a.

With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. In this condition, the remaining control room emergency air filtration system is adequate to perform the control room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE train could result in a loss of the control room emergency air filtration system function. The 7-day completion time is based on the low probability of a DBA occurring during this time period, and the ability of the remaining train to provide the required capability.

If the inoperable train cannot be restored to an OPERABLE status within 7 days, the unit must be placed in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . These completion times are reasonable, based on operating experience, to reach the required unit condition from full power conditions in an orderly manner and without challenging unit systems.

b.

With both control room emergency air filtration systems inoperable, except due to an inoperable control room boundary, the movement of fuel within the spent fuel pool must be immediately suspended. At least one control room emergency air filtration system must be restored to OPERABLE status within I hour, or the unit must be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . These completion times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

c.

With both control room emergency air filtration systems inoperable due to an inoperable control room boundary, the movement of fuel within the spent fuel pool must be immediately suspended. The control room boundary must be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the unit must be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

If the control room boundary is inoperable in MODES 1, 2, 3, and 4, the control room emergency air filtration systems cannot perform their intended functions. Actions must be taken to restore on OPERABLE control room boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the period that the control room boundary is inoperable, appropriate compensatory measures (consistent with the intent of GDC 19) should be utilized to protect control room operators from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be MILLSTONE - UNIT 3 B 3/4 7-13 Amendment No. 4-36, I0, 219

PLANT SYSTEMS BASES Fof :Q&..,.ees 0),

314.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

ACIIONS (Continued) l available to address these concerns for intentional and unintentional entry in to this condition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed outage time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed outage time is a typically reasonable time to diagnose, plan, and possibly repair, and test most problems with the control room boundary.

MODES 5 and 6. and fuel movement within containment or the spent fuel pool

d.

With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the recirculation mode or suspend the movement of fuel. Initiating and maintaining operation of the OPERABLE train in the recirculation mode ensures:

(i) operability of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected.

e.

With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'd' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving the movement of fuel. This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position.

SURVEILLANCE REQUIREMENTS 4.7.7.a The control room environment should be checked periodically to ensure that the control room temperature control system is functioning properly. Verifying that the control room air temperature is less than or equal to 950F at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient. It is not necessary to cycle the control room ventilation chillers. The control room is manned during operations covered by the technical specifications. Typically, temperature aberrations will be readily apparent.

4.7.7.b Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing the trains once every 31 days on a STAGGERED TEST BASIS provides an adequate check of this system. This surveillance requirement verifies a system flow rate of 1,120 cfm +/- 20%.

Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized. These operations are sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters due to the humidity in the ambient air.

MILLSTONE - UNIT 3 B 3/4 7-13 a Amendment No. 436,48+, 23, 219

PLANT'SYSTEMS July 24, 2002 BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued)

Foy O 04) 4.7.7.c The performance of the control room emergency filtration systems should be checked periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. The frequency is at least once per 24 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.

ANSI N510-1980 will be used as a procedural guide for surveillance'testing.

4.7.7.c.1 This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of'less than O.05% In accordance with Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 1,120 cfm t 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in the regulatory guide.

4.7.7.c.2 This surveillance requires that a representative carbon sample be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the representative carbon sample meets the laboratory testing criteria of ASTH D3803-89 and Millstone Unit 3 specific parameters. The laboratory analysis is required to be performed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm t 20%,

during system operation when testing in accordance with ANSI N510-1980.

4.7.7.d After 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters.

MILLSTONE - UNIT 3 B 3/4 7-14 Amendment No. 11f, JIf, 206

PLANT SYSTEMS RY.

0A)

July 24, 2002 BASES 3/4.7.7 CONTROL ROOMKEMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued)

The laboratory analysis "is required to.be performed within 31 days after removal of the simple... ANSI. N510-1980 is 'used in lieu of ANSI N510-1.975 referenced' in Revision 2 of Regulatory Guide 1.52.

The maximum surveillance interval 'is'900 hours0.0104 days 0.25 hours 0.00149 weeks 3.4245e-4 months , per Surveillance'Requirement 4.0.2. The 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of operation requirement originates from Nuclear Regulatory Guide 1.52., Table 2, Note C. This testing ensures that the charcoal adsorbenrcy capacity has-n'ot degraded belbw acciptable limits as well as providing trending

data.

4.7.7;e.1

.This -surveillance verifies that the pressure drop across the combined HEPA filters and cha'rcoal adsotbers banks' at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm i 20%.

The frequency is at least once per 24 months.

I 4.7.7.e.2

'This surveillance verifies that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch water gauge at less than or equal to a pressurization flow of 230 cfm relative to adjacent areas and outside atmosphere during positive pressure system operation. The frequency is at least once per 24 months.

I The intent of this surveillance is to verify the ability of the control room emergency air filtration system to maintain a positive pressure while running in the filtered pressurization mode.

MILLSTONE - UNIT 3 B 3/4 7-15 Amendment No. Aid, Aid, ZIf, IF1,206

Jal, LA4 f2tO2 PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued)

During the first hour, the control room pressurization system creaes and maintains..the posit'ivetpresiure in th-econtrol'room.

This capability Jis verified by Surveillance Re ulrement

8.C de endent of Surveillance Requirement 4.7.7.e..2. ACBI.signal will adtomatically align an operating filtration system into the recirculation mode of'bperati-on due to the isolation of the air supply line to the filter.

After the first hour of ain'evint with the potential for a radiological' r'elease,'.the control room.emergeiky entiThtin'ysted wi'll be aligned in either the recirculation mode (isolated from the outside environment) or filtered pressurization mode (outside air is diverted through the filters to the control room'envelope to maintain a' posftive'pressure)..

The mode of service for the control room.emergency air filtration system will be based on the radiological conditions that exist outside the control room.

Alignment to the filtered pressurization mode requires manual operator action to open the air supply line.

4.7.7.e.3 This surveillance verifies that the heaters can dissipate 9.4 t I kW at 480V when tested in accordance with ANSI N510-1980.

The frequency is at least once per 24 months. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can, lead to measurements of kW which cannot be compared to the nameplate riting'because the output kW is proportional to the square of the voltage.

4.7.7.f Following the complete or partial replacement of a HEPA filter bank, the operability of the cleanup system should be confirmed. This is accomplished by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1,120 cfm +/- 20%.

MILLSTONE'- UNIT 3 B 3/4 7-16 Amendment No. Afi, jpj, lop 0934

PLANT SYSTEMS June_3,_2W2_a_

BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.

Following the complete or partial replacement of a charcoal adsorber bank, the operability of the cleanup system should be.confirmed. This is accomplished

-by verifying that the cleanup system satisfied the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow of 1,120 cfm +/- 20%.

References:

(1)

(2)A 'I Nuclear Regulatory Guide 1.52, Revision 2 MP3 UFSAR, Table 1.8-1, NRC Regulatory Guide 1.52 NRC Aeneric Letter 91-04 (4) Condition Report (CR) #M3-99-0271 P eL 314.7.8 ONTROL cl_

I BACKiURUND 211 The control room envelope pressurration system provides a protected environment from which operators can cgtrol the unit following an uncontrolled release of radioactivity.

The control room envelope pre urization system consists of two ban of air bottles with its associated piping instrumentation, and controls. Each ank is capable of providing the control om area with one-hour of air followi any event with the potential for rad active releases.

Control Room Envelope OPERABIL Y is satisfied while:

I d

Door 352 (C-49-1 Is closed (East door)

Door 351 (C is closed, but C-47-IA, ATD/Missile closed (West d ors)

Normal Operation During normal op ations, the control room envelope pre sur required to be on sta dby.

Post Accident 0 eraaion is not ization system is The control oom envelope pressurization system is.equired to operate during post-acci ent operations to ensure the control r m will remain habitable during and foll ing accident conditions.

The seq ence of events which occurs upon receip of a control building isolation (C I) signal or a signal indicating high r diation in the air supply duct to the control room envelope is described in B ses Section 3/4.7.7.

I MILLSTONE - UNIT 3 0894 B 3/4 7-17 I

Amendment No.

ifd, "Revised by NRC Letter A157104

, PLANI SYSIEMS BASES ROOM Eb^gkUM PRESSURIATION SYSEM (Continued)

AFTICARLED SAFEiTY AAYI I

The OPERAB of the control room envelope pres-rization system ensures that:

(1) breathable air is s plied to the control room, instrumenta n rack room, and computer room, and (2) a positive essure is created and maintained withie control room envelope during control building olation for the first hour following any vent with the potential for radioactive releases. Eac system is capable of providing an ad air supply to the control room for one hour foll g an initiation of a control building iso on signal. After one hour, operation of the contr om emergency ventilation system woul initiated.

ON FOR OPE Two independent control room nvelope pressurization systems operable to ensure that at least one iavailable in the event the other yd A control room when the associated:

a.

air storage

b.

piping and OPERABLE; and The integrity of!e control room habitability bo dary (i.e., walls, floors, ceilings, ductwork, and access doors ust be maintained. However, e LCO is modified by a footnote allowing the control ron boundary to be opened inte ently under administrative controls. For entry and exit throtgh doors the administrative co of the opening is performed by the person(s) entering or exiting the area. For othe penings, these controls consist of stationing a dedicated individual at the opening who is i'onstant communication with the control room. This individual will have a method pidly close the opening when a need for control room isolation is indicated.

I

.d L-tL4sdX,/Lr L=+

64k MULSTONE -UNIT 3 B 3147-18 Amendment No. 436,,3 M

LANT SYSTEMS BASES 1 S4OM ENVELOPE PRESSURIZATION SYSTEM (Continued)

In MOE,2,4, 5, and 6.\\

During movement within containmentrthe spent fuel pool.

Acti s ab., c., and d. of this specific ion are applicable at all times during plant operation in ES 1,2, 3, and 4. Actions e.

f. are applicable in MODES 5 and 6, and whenever fuel is ing moved within containment o the spent fuel pool. The fuel han ing accident analyses assume that during a fuel handli accident some of the fuel that is opped and some of the fuel that is impacted upon is dama d. Therefore, the movement of e Cr new or irradiated fuel (assemblies or individual fuo rods) can cause a fuel handling ident, and this specification is applicable whenever new r irradiated fuel is moved within containment or the storage pool.

MODES 1, 2, d4

a.

Wi ne control room envelope press *zation system inoperable, action must be taken

i to restore the inoperable syste to an OPERABLE status within 7 da s, or place the t in HOT STANDBY within s'hours and COLD SHUTDOWN wi n the next

/0hours.//

The remaining control roo, nvelope pressurization system is a uate to perform the control room protectio nction. However, the overall reliabi y is reduced because a single failure in the ERABLE train could result in a loss the control room envelope pressurization sys The 7-day completion time is bas on the low probability of a design basis ac ent occurring during this time perind the ability of the remaining train to provae the required capability.

The co etion times for the unit to be placed' HOT STANDBY and COLD S

OWN are reasonable. They are bas on operating experience, and they permit

\\thenit to be placed in the required condi ons from full power conditions in an orderly Iihinner and without challenging unit s ems.

b.

With both control room envelope essurization systems inoperable, except due to an inoperable control room bound or durng performance of Surveillance Requirement 4.7.8.c, the movement of fuel ithin the spent fuel pool must be immediately suspended.

At least one control room e elope pressurization system must be restored to OPERABLE status withi hour, or the unit must be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD OWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . These completion times are reasonabi ased in operating experience, to reach the required unit conditions from full power nditions in an orderly manner and without challeng unit systems.

MILLSTONE -UNIT 3 B 3/4 7-19 Amendment No. 4A6, 203, MO%

-FAI-Ave

-"c+*

He

-, R

PLANTSYSTEMS BASES 3/4.7.8 CONTROL ROOM EN*

E PRESSURIZ AWIO VA (Continued)

ACtIONS (Continued)

c.

With both tIo room envelope pressuriza nsystems inoperable due to noperable control m boundary, the movement offi within the spent fuel poo ust be i

ately suspended. The control r m boundary must be restor to OPERABLE sta s within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the unit at be in HOT STANDBY wi nte next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within e following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

/ f the control room boun is inoperable in MODES,,

3, and 4, the control m

envelope pressurizati systems cannot perform the tended functions. Acti smust be taken to restore PERABLE control room bo dary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . D ug the period that the ntrol room boundary is ino able, appropriate compe ory measures (consistent the intent of GDC 19) sho be utilized to protect co ol room operators from po a1 hazards such as radioac&e contamination, toxic ch icals, smoke, tempe hire and relative humidity, d physical security. Prepl ed measures should be av able to address these conce for intentional and uninte onal entry in to this ondition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allow outage time is reasonable ased on the low probability of aDBA occurring during s time period, and the use compensatory measures. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed outage e is a typically reasonable to diagnose, plan, and poss y repair, and test m problems with the control rm boundary.

d.

With both ntrol room enveope pressur don systems inoperable dunn e

perfo ce of Surveillance Requireint 4.7.8.c and the system not be g tested under ad istrative control, the move t of fuel within the spent fuel p must be

mediately suspended. At Ieone control room envelope press zation system must be restored to OPERABLE atus within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or the unit mu in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months d in COLD SHUTDOWN within foilowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The administrative contr for the system not being tested co st of a dedicated operator, in constant communation with the control room, who ca rapidly restore this system to OPERABLE sas. Allowing both control room cnv ope pressurization systems to be inoperable f 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under administrative contro acceptable since the system not being test is inoperable only because it is isola

. Therefore, the system can be rapidly restore f needed. The other completion times reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manne and ithot challenging unit systems.

Ems e<<

n i-c.-WPKAR Let+ 044ac MILLSTONE - UNIT 3 B 3/4 7-20 Amendment No. +36,48i i20I, W

"PLANT SYSTEiMS BASES

L4JL

AcIL

CONTROL ROOM ENrEoPE~w PRESSURIZATION SYSTEM (Continued)

(

US (Continu MODES 5 and 6.

d fuel movement within contaient or t pe spent fuel nool I

e.

With one co o room envelope pressurizatio ystem inoperable, action must be taken to restore the operable system to an OPE status within 7 days. After 7 da,

immedi ely suspend the movement of fu This action places the unit in a co tion that mini zes potential radiological expos e to Control Room personnel. This don does no reclude the movement of fuel safe position.

The remaining control room en lope pressurization system is ad e

to perform the control room protection fuc on. However, the overall reliabiity reduced because a single failure in the 0 LE train could result in a loss of e control room envelope pressurization system.

e 7-day completion time is based the low probability of a design basis acciden curring during this time period the ability of the remaining train to provide th equired capability.

Stud tensiongmay continue in MODE 6 and ODE change to MODE 5 is permitted with a co ol room envelope pressurization ter inoperable (Reference 1).

f.

Wit oth control room envelope pres don systems inoperable, immediately sus hdi vement of fuel. This action ples the unit in a condition that minimizes pot ial

ological exposure to Control om personnel. Ihis action does not preclude e

/movement of fuel to a saepsjin/

SURVILLACE REQIE NT 4.7.8.a This surveillan requires verification that the air bottles are poperly pressurized.

Verfying that the air ottles are pressurized to greater than or equal 2200 psig will ensure that a control room enve4pe pressurization system will be capable of s plying the required flow rate.

The frequency the surveillance is at least once per 7 days. I s ased on engineering judgment and has bee own to be appropriate through operating cx rience.

This surveillance requires verification of correct position of each valve

anual, power operated, or automatic) in the control ro envelope pressurization syste ow path. It helps ensure that the control room envelope essurization system is capable of rforming its intended safety function by verifying that i appropriate flow path will exist.he surveillance applies to those valves that could be mi sitioned. This surveillance does ot apply to valves that have been locked, sealed, or sec d in position, because these positi s are verified prior to locking, sealing, or securing.

The frequency of the veillance is at least once per 31 da on a STAGGERED TEST BASIS. It is based on engin ring judgment and has been shown to be appropriate through operating experience.

MILLSTONE -UNIT 3 B 3/4 7-20a Amendment No. 36.4,8+. 2.

21L9 I

1'brs p-ec l*'+V<e-"Mttor LugeP H6a"

PLANT SYSTE' S BASES T-ASSe M7k2

)~nVteAMY Lft zM k

. 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (C niued).

SURVEILLANCE REQUIREMENTS (Continued) 4..7.8.c The performance of the control room envlope presstirization..system sho ld be checked periodically.. The frequency is least once per 24 months an following any major alteration of the con ol. room envelope pressure bo dary.

A major alteration is a change a-the control room envelope pr sure-boundary that:

(1) results in a b ach greater than analyzed for cceptable pressurization and requires nonr tine work evolutions to restpr the boundary.

A non'routine wbrk evolution i ' ne which miakes it difficult-t - determineeAs-Found and As-Left conditiois. Ex lestof rdutffie w6okevolution riclu'de. '(1) opening and closing a door, and'(

repairing cable and pipe pene ations because the repairs are conducted I accordance with procedures and. re verified.via inspections. for th X two examples, there 'is a high evel of a'ss~utance that the boundary is restore to the As-Found condlfiorr;

/

This sur I lance requires at least once r 24 months or following a major alteration o the control room envelope press e boundary by:

Ve fying.t~he control.room envelop

isolated in response to.a, Control ilding Isolation Test signal, Verifying, after a 60 second ime delay following a Control Buildin Isolation Test signal, the ontrol room envelope pressurizes to gr ater than or equal to 0.125 iwater gauge relative to'adjacent arcs and outside atmosphere; an Verifying the posi ye pressure of Technical Specificatio

.7.8.c.2'is maintained for r ater than or equal'to 60 minutes.

Changes in conditions outside the control room envel e cause pressure spikes which are eflected on the differential pressure dicator, 3HVC-PDI 113.

Pressu spikes or fluctuations which result i the. differential pressure.

momentaril dropped below the 0.125 inch water gafe acceptance criteria are acceptab providing the following conditions a met:

1.

Differential pressure remains positiv at all times.

2.

Differential pressure is only tra itorily below the acceptance criteria.

3.

Differential pressure returns o a value above the acceptance criteria.

M4ILLSTONE - UNIT 3 5 3/4 7-21 Amendment No. Ad0, *0L !MT.

Di ART CYCTMIZ.

A-r, -

BASES W

fre'r t a

EMTrN LfT 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (Con igd)

SURVEILLANCE REQUIREMENTS (Continued)

The control room envelope pressurizatpn-system design.basis cr-iteriais set at 2 0.125 inch water gauge criteria to ccount for wind effects,. thermal col n effects, and barometric pressure cha Jes. Pressurizing the control roomen" ape of 0.125 inch water gauge above t initial atmospheric pressure ensures will remain at a po'sitive.pressure jhking subsequent changes in outside con ions over the next 60 minutes. Sincep e surveillance requirement is verified*

. actual reference to'outside pre ssre, allowances are provided for differeniai pressure fluctuations caused b xternal forces. The 0.125 inch water ga e acceptance.

criteria provides We margin for these fluctuations. This.mee

. the requirements of Regulatory GTe. 1.78 and NUREG-800, Sectiond.4 and is

.~.sisteht assumptions the Control Room Operator DBA. dose calcula n.

/This surveillance verifies that the control om envelope is isolated fllowing a control building isolation (CBI) te signal.

4.7.8.c.2 This surveillance verifies that t control room enveloge pressurizes greater than or equal to 1/8 inch wa r gauge, relative to t e outside atmsphere, after 60 seconds following receipt f a IBI test signal.

4.7.8.c.3//

This surveillance ve ies that the positive pressure dev oped in accordance with Surveillance Requi ment 4.7.8.c.2 is maintained for greer than or equal to 60 minutes. This cap ility is independent from the requl ents regarding the control room emerge. y filtration system contained in Te nical Specification 3/4.7.7.' Also,'f lowing the first hour, the control om emergency ventilation system is resp sible for ensuring that the control am envelope remails habitable.

Referenc (1) RC Routine Inspection Report 50-423/ -33,.dated February 10, 1988.

(2) NRC Generic Letter 91-04.

MILLSTONE - UNIT 3A B 3/4-742 knendment No. &

3/4.9 REFUELNG OPERATIONS BASES 3/4.9.10-AND 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL

( g~ ) The restrictions on minimum water level ensure that sufficient water depth is available to remov>99% of the assumed((ene gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

MILLSTONE - UNIT 3 B 3/4 9-8 Amendment No. 9, 4i, i49, W,

-t8A49, 20a, 77,,

1XI)

v t e Letter Sequence Other
TAC:MC3333, (Approved, Closed) TAC:MC0498, (Open)
Initiation Request Acceptance... Supplement Administration Meeting Questions 20 August 2004 22 September 2004 29 December 2004 9 February 2005 18 November 2005 Answers Results Approval Other: ML041560464
MONTHYEAR2004-08-20 [Table View]

ML041560464

Text

1.0 INTRODUCTION

1.1 INTRODUCTION

SUMMARY

SUMMARY

5.0 CONCLUSION

6.0 REFERENCES

References:

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