ML13025A238

Usfar 15.10.6.3.2 SGTR Analysis

ML13025A238
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Site:	San Onofre
Issue date:	06/10/2011
From:	Division of Operating Reactor Licensing
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San Onofre Nuclear Generating Station, Unit 2 & 3 Updated Final Safety Analysis Report Revised Apri I 2011 Chapter 15.0 - Accident Analyses Section 15.10.6.3.2 - Steam Generator Tube Rupture With Concurrent Loss of AC Power

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS The offsite radiological doses for the Primary Sample or Instrument Line Break with an accident-induced iodine spike are "a small fraction" (i.e., do not exceed 10%) of the 10 CFR 100 exposure guidelines, and the Control Room radiological doses are within the 10 CFR SO Appendix A General Design Criterion 19 exposure guidelines.

lS.1O.6.3.2 Steam Generator Tube Rupture with Concurrent Loss of AC Power Introduction A Steam Generator Tube Rupture (SGTR) event is a penetration of the barrier between the Reactor Coolant System (RCS) and the main steam system via the double-ended break of a U-tube. This causes highly radioactive RCS fluid to contaminate the secondary side. The radioactivity is released via the condenser air ejectors, the Main Steam Safety Valves (MSSVs),

and the Atmospheric Dump Valves (ADVs).

This event is analyzed with a concurrent loss of AC power, which increases the radiological release to the environment (see section lS.6.3.2.S). It is this analysis which is presented below.

Ifthe primary to secondary leak is beyond the capacity of the charging pumps, the reactor will eventually trip on a low pressure trip signal. As a result ofthe loss of AC, the electrical power would be unavailable for the station auxiliaries such as the Reactor Coolant Pumps (RCPs) and the Main Feed Water (MFW) pumps. Under such circumstances, the plant would experience a simultaneous loss ofload, normal feed water flow, forced reactor coolant flow and steam generator blowdown capability.

When the reactor is off line, stored energy and fission product decay energy must be dissipated by the reactor coolant and main steam systems. In the absence of forced reactor coolant flow, convective heat transfer is supported by natural circulation reactor coolant flow. Initially, the liquid inventory in the steam generators is used and the resultant steam is released to the atmosphere via the MSSV s. With the availability of stand-by power provided by the automatic start-up ofthe diesel generators, Auxiliary (emergency) Feed Water (AFW) flow is initiated on a low steam generator level signal.

When the reactor plant has been stabilized in Mode 3, the operator achieves plant cool down using remotely operated ADVs. The plant is cooled to 3S0°F at a nominal rate of7soFIhr. At this time, Shut Down Cooling (SDC) is initiated.

The analysis of record conservatively assumes the operator action to isolate the affected steam generator is delayed until 30 minutes after initiation of the event. The operator's diagnosis of the SGTR event is facilitated by the radiation monitors which initiate alarms and signal the existence of abnormal radioactivity levels.

Radiation monitors are found in the blowdown sample lines from each steam generator, in the blowdown processing system neutralization sump discharge sea line which processes blowdown from both steam generators, and in the condenser air ejector discharge line. Additional IS.1O-177 Rev: 44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS diagnostic information is provided by RCS pressure and pressurizer level response indicating a loss of primary coolant. Level in the affected steam generator increases as the primary fluid enters the steam generator driven by the substantially higher primary pressure.

The offsite and control room dose consequences of the postulated steam generator tube rupture are analyzed for the assumed conditions of no iodine spike, a pre-accident iodine spike, and an accident initiated iodine spike in the reactor coolant.

Summary of Methods The CESEC-III code is used to simulate the transient for the first 1800 seconds (i.e., 30 minutes).

The output of the code provides the amount of primary to secondary leak, the amount of steam transported from the steam generators through the MSSVs and the overall Nuclear Steam Supply System (NSSS) response to the event. This information is then used to derive the radiological releases and accompanying doses.

This analysis is primarily performed to establish the parameters, such as the primary to secondary mass transferred during the event, by which the radiological releases are calculated.

There is no specific acceptance criteria for the mass releases.

One computer case was run for this analysis. This was an 1800 seconds CESEC-III simulation of a double-ended SGTR in the right hand (arbitrary designation) steam generator. This case utilizes a 15% MSSV blowdown model to determine the impact on steam released to atmosphere.

The primary transient analysis inputs and assumptions for the analysis are presented below in Table 15.10.6.3.2-1. The sequence of events is provided in Table 15.10.6.3.2-2.

The dose methodology for this event is described in Appendices l5B and l5.10.B. Using this methodology, design basis 0-2 hour Exclusion Area Boundary, 0-30 day Low Population Zone, and 0-30 day Control Room doses were calculated with and without consideration of pre existing and accident induced iodine spikes.

The following release mechanisms that can disperse radioactive material into the atmosphere have been evaluated:

1. Reactor coolant releases via the ruptured tube into the affected steam generator, and eventually to the outside environment.

2. Normal primary to secondary leakage releases into the affected and intact steam generators, and eventually to the outside environment.

3. Main steam safety valve releases from the affected and intact steam generators to the outside environment.

15.10-178 Rev: 44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS

4. Turbine-driven auxiliary feed water pump venting of secondary steam from the affected and intact steam generators to the outside environment.

5. Atmospheric dump valve releases of secondary steam from the intact steam generator to the outside environment.

6. Leakage past one or more of the affected steam generator MSSVs and/or its ADV (subsequent to operator action to isolate the affected steam generator).

Conservatively, the total leakage is modeled as being equivalent to the flow capacity of a MSSV.

The principal assumptions and inputs for the dose analysis are presented below in Table 15.10.6.3.2-3.

15.10-179 Rev: 44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.10.6.3.2-1 Principal Assumptions and Inputs for SGTR Parameter Unit 2 Unit 3 Core Power 3478 MWth 3478 MWth Inlet Temperature 560°F 560°F R CS Pressure 2300 psia 2300 psi a SG Pressure 900 psia 900 psia Core Flow, Total 376,200 gpm 376,200 gpm BOC Doppler Uncertainty Multiplier 0.86 0.86 Moderator Temperature Coefficient -3.7 x 1O-4.6.p/oF -3.7 x 1O-4.6.p/oF

~

SCRAM Worth -6.0 %.6.p CPC (range -low pressure) Trip Set Point 1785 psi a Loss of AC Power Coincident with Coincident with Reactor Trip Reactor Trip Steam Generator (S/G) U-Tube Break Size 45% Double 45% Double Ended Guillotine Ended Guillotine Safety Injection Actuation System - Set 1785 psia 1785 psia Point High Pressure Safety Injection - Response 15.0 seconds 15.0 seconds Time Main Feed Water (MFW) - Flow Rate 102% of Design 102% of Design Main Feed Water (MFW) Enthalpy 425 Btu/Ibm 425 Btu/Ibm (pre-trip) (pre-trip)

Auxiliary Feed Water (AFW) - Response 57.7 sees. electric 57.7 sees. electric Time and steam driven and steam driven Auxiliary Feed Water (AFW) - Flow Rate 601 gpm at 1000 601 gpm at 1000 psia psi a Auxiliary Feed Water (AFW) - Enthalpy 68 Btu/Ibm 68 Btu/lbm Charging Flow Rate 135gpm 135gpm Letdown Flow Rate O.Ogpm O.Ogpm 15.10-180 Rev:44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.10.6.3.2-1 (continued)

Principle Assumptions and Inputs for SGTR Parameter Unit 2 Unit 3 Main Steam Safety Valves (MSSV) 1067 to 1120.4 1067 to 1120.4 Opening Set Points pSla pSla (9 valves at 7 psi increments. -3% Tolerance -3% Tolerance Includes set point tolerance)

MSSV Accumulation Set Point 0% 0%

MSSV Blow Down 15% (to fully 15% (to fully close) close)

Atmospheric Dump Valves (ADVs) Inoperative Inoperative Feed Water Control System (FWCS) Not Required to Not Required to Mitigate Event Mitigate Event Pressurizer Pressure Control System (PPCS) Not Required to Not Required to Mitigate Event Mitigate Event Steam Bypass Control System (SBCS) Inoperative Inoperative 15.1 0-181 Rev:44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.l0.6.3.2-2a Sequence of Events for SGTR, Unit 2 Time Chronological Event Set Point or Value (Seconds) 0.0 S/G Tube rupture occurs 1329.4 CPC Reactor Trip (low pressurizer pressure) 1785 psia setpoint reached, SIAS initiated 1330.3 Reactor Trip breakers open ---

Turbine Stop Valves close Loss of Normal AC 1333.1 MSSV s begin to open on both S/Gs 1067 psia 1337.9 Maximum S/G pressure on both generators 1112 psia 1356.1 Low S/G level signal generated, AFW initiated 115610 Ibm 1795.8 MSSV s close on both S/Gs 907 psi a 1800.0 Damaged S/G isolated, ---

ADV on unaffected S/G opened to begin system cool down to Shut Down Cooling (SDC) 11880.0 Shut Down Cooling (SDC) initiated Temperature 350°F Total steam release 739,034 Ibm 15.10-182 Rev: 44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANAL YSIS Table 15.1O.6.3.2-2b Sequence of Events for SGTR, Unit 3 Time Chronological Event Set Point or Value (Seconds) 0.0 S/G Tube rupture occurs ---

1329.4 CPC Reactor Trip (low pressurizer pressure) 1785 psia setpoint reached, SIAS initiated 1330.3 Reactor Trip breakers open ---

Turbine Stop Valves close Loss of Normal AC 1333.1 MSSVs begin to open on both S/Gs 1067 psi a 1337.9 Maximum S/G pressure on both generators 1112 psia 1356.1 Low S/G level signal generated, AFW initiated 115,610 Ibm 1795.8 MSSVs close on both S/Gs 907 psia 1800.0 Damaged S/G isolated, ---

ADV on unaffected S/G opened to begin system cool down to Shut Down Cooling (SDC) 11880.0 Shut Down Cooling (SDC) initiated Temperature 350°F Total steam release 739,034lbm 15.1 0-183 Rev:44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.10.6.3.2-3 Principal Assumptions and Inputs for Steam Generator Tube Rupture Dose Analysis Parameter Unit 2 Unit 3 AC Availability Loss of AC LossofAC Power Power RCS Iodine Activity (Dose Equivalent 1-131), IlCilgm 1.0 1.0 Increase in Iodine Release Rate from Fuel for Accident 500 500 Induced Iodine Spike RCS Pre-Existing Iodine Spike Iodine Activity (Dose 60 60 Equivalent 1-131), }lCi/gm RCS Non-Iodine Activity, }lCi/gm 10010 100/0 Secondary Liquid Iodine Activity (Dose Equivalent 0.1 0.1 1-131), }lCi/gm Steam Generator Iodine Partition Coefficient 0.01 0.01 Primary to Secondary Leak Rate into each SG, gpm 0.5 0.5 Integrated primary to secondary rupture flow, Ibm 70,563 70,563 (1,800 seconds)

Additional primary to secondary rupture and normal 150,000 150,000 leakage flow available for release from 30 minutes to shut down cooling due to MSSV / ADV valve seat leakage, Ibm Integrated MSSV flow, Ibm (1,800 seconds)

LH - Unaffected 57,560 57,560 RH - Affected 57,664 57,664 Total MSSV Flow 115,224 115,224 AFW Flow (steam driven pump), Ibm (1,800 seconds) 4,922 4,922 Steam Release (30 - 120 minutes), Ibm 331,547 331,547 Total steam release (0 - 120 minutes), Ibm 451,693 451,693 Total steam release to Shut Down Cooling, Ibm 739,034 739,034 Additional affected steam generator steam release from 2,400,000 2,400,000 30 minutes to shut down cooling due to MSSV/ADV valve seat leakage, Ibm Control Room Isolation Signal High Radiation High Radiation Control Room Isolation Time, min 3 3 Offsite Dose Evaluation Model Appendix 15B Appendix l5B and Appendix and Appendix l5.l0B l5.10B Control Room Dose Evaluation Model Appendix 15B Appendix I5B and Appendix and Appendix I5.lOB I5.lOB 15.10-184 Rev:44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANAL YSIS Results The primary to secondary mass transfer and steam release data required to perform radiological calculations for the steam generator tube rupture event are presented in Table 15.10.6.3.2-3.

The RCS and secondary system pressures remain below the 110% of the design pressure limits, thus, assuring the integrity of these systems.

The results of the most recent analysis of the potential off site and control room personnel doses from a steam generator tube rupture with concurrent loss of normal AC power are presented in Table 15.10.6.3.2-4. These results are compared against the NRC approved acceptance criteria in section 15.6.3.2.

15.1 0-185 Rev:44 I

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.10.6.3.2-4 Results for Steam Generator Tube Rupture Analysis Results Parameter Acceptance Criteria Unit 2 Unit 3 Design Basis Case with No Iodine Spike 0-2 hr EAB Doses, Rem Thyroid 30 0.8 0.8 Beta Skin N/A 0.1 0.1 Whole Body 2.5 0.2 0.2 0-30 day LPZ Doses, Rem Thyroid 30 <0.1 <0.1 Beta Skin N/A <0.1 <0.1 Whole Body 2.5 <0.1 <0.1 0-30 day Control Room Doses, Rem Thyroid 30 1.8 1.8 Beta Skin 30 1.6 1.6 Whole Body 5 <0.1 <0.1 Design Basis Case with Pre-Existing Iodine Spike 0-2 hr EAB Doses, Rem Thyroid 300 8.2 8.2 Beta Skin N/A 0.1 0.1 Whole Body 25 0.2 0.2 0-30 day LPZ Doses, Rem Thyroid 300 0.2 0.2 Beta Skin N/A <OJ <0.1 Whole Body 25 <0.1 <0.1 0-30 day Control Room Doses, Rem Thyroid 30 2.0 2.0 Beta Skin 30 1.6 1.6 Whole Body 5 <0.1 <0.1 Design Basis Case with Accident Induced Iodine Spike 0-2 hr EAB Doses, Rem Thyroid 30 4.1 4.1 Beta Skin N/A 0.1 0.1 Whole Body 2.5 0.2 0.2 0-30 day LPZ Doses, Rem Thyroid 30 OJ 0.1 Beta Skin N/A <OJ <0.1 Whole Body 2.5 <0.1 <0.1 0-30 day Control Room Doses, Rem Thyroid 30 1.9 1.9 Beta Skin 30 1.6 1.6 Whole Body 5 <0.1 <0.1 15.10-186 Rev:44 I