ML20236P393

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Rev 0 to Maint Rule Scoping & Risk Significance Determination Results of Expert Panel Meetings
ML20236P393
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 03/22/1996
From: Allen R, Beth Brown, Short M
SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML20236P359 List:
References
NUDOCS 9807160294
Download: ML20236P393 (41)
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Text

_ - _ - - _ _ - - _ _ - - - - _ _ _ __ __-_ _ _ __-__

! .. ATTACHMENT 1

! MAINTENANCE RULE SCOPING ,

RISK SIGNIFICANCE DETERMINATION RESULTS OF EXPERT PANEL MEETINGS l

):

Date: March 22,1996 Revision 0 l Expert Panel Approvals:

Mike Short Y Si;/Iechnical Services Rick Allen [# A Station Technical Bob Brown -

Operations s

Neil Bloo ,

, _ Maintenance Sam Chien A[

Nuclear Oversight TorreyYee 6mdw NuclearMichering Design Organization  ;

. l 9907160294 990714 PDR ADOCK 05000361 .

G PDR l

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l. A
5. Number of abnormat radiological releases l

! a. This criteria is calculated from the number of unplanned radiological releases reported annually under Tech Spec. Section 6.1.9.8, entitled, " Annual Radioactive Effluent Release Report".

b. The current criteria has been set at the following:
1. 2 Events in a 2-yr monitoring period - acceptable
2. 3 Events in a 2-yr monitoring period - not acceptable
6. Core damace freauenev The current performance criteria is set at 3.2E-5/ unit /yr, any excursion beyond this value will be considered not acceptable.

In order to effectively review the balance of systems that were initially not considered by the PRA to be risk significant, the 10 questions listed below were developed to assist in making the final determination as to whether or not there is some non PRA merit for including any of these SSCs as risk-significant. As a panel member, please apply these questions to each of the PRA non risk-significant systems listed in Attacbment 4 and if you believe the SSC to be risk significant indicate so by putting a "Y" in the tnird column adjacent to the system and identify which question number (s) you believe support that conclusion by placing the question number in the forth column. The panel as a group will collectively make the final determination based on the inputs from the group. Use of the fifth column is discussed further down.

Accident Resconse Functions: (Items below were covered by the PRA) i A1. Required to shutdown the reactor and maintain it in a safe shte.down condition?

A2. Required to maintain the reactor coolant pressure boundary?

A3. Required to remove atmospheric heat and radioactivity from containment and maintain containment integrity? l A4. Required to remove heat from the reactor?

Normal Ooeration Functions: (Deviations from normal operations should result in accident l initiation and/or safety system cha!!enges.) l Bl. Required to provide primary side heat removal? l B2. Required for power conversion?

B3. Required to provide primary, secondary, or containment pressure control?

B4. Required to provide cooling water, component or room cooling?

B5. Required to provide electric power (AC,DC power)?

B6. Required to provide other motive or control power (instrument air)?

f Note: The above criteria was taken from NUREG/CR-5695, "A Process for Risk Focused I

o.

M-Rule, Expert Panel Disposition Results Summary,3/19/96 Page 1 of 6 l Bold =STT/STS' renamines ,

Ite OpSys Description ";/N Risk Basis A&B C-1 Y/N PRA/EP m# Code Table Table 1 AFW AUXILIARY Y Y PRA FEEDWATER I

2 BLDG BUILDINGS Y Y Expert A-3, Panel A-1 3 CCS CONTAINMENT Y Y PRA '

1 SPRAY SYSTEM i

4 CCW COMPONENT Y Y PRA l l

COOLING WATER 5 CHW CHILLED WATER Y Y PRA 6 CND CONDENSATE Y Y PRA 1 CR-ANUN CONTROL ROOM Y Y PRA 7

ANNLrNCIATOR SYS 8 CVCS CHEM & VOL CONT Y Y PRA SYSTEM 9 EDG EMERG. DIESEL Y Y PRA GENERATOR 10 DC CLASS IE & NON 1E 125 Y Y PRA VDC & 250 VDC SYS 11 120V IE & NON IE CTRL 120V Y Y Expert B-5 INVERTER SYS Panel l

12 430V LOW VOLTAGE DISTRIB Y Y PRA SYS (IE & NON 1E) 13 4KV 4.16KV SYSTEMS Y Y PRA

[

(IE & NON 1E) 14 XFMRS MAIN: UNIT & RES. Y Y PRA g TRANS. _

l l

+

M-Rule, Expert Panel Disposition Results Summary,3/19/96 Page 3 of 6 Bold =STT/STS' renamines Item OpSys Description M-Rule Risk Basis A& B- C-

  1. Code Y/N Y/N PRA/EP Table Table i

29 CIRC CIRC. WATER Y N Expen B-2 C-1 Panel 30 CMNC COMMUNICATION Y N Expen SYSTEM Panel l 31 FFCPD FULL FLOW COND. POLISHING Y N Expen DEMINERALIZED Panel 32 FHS FUEL HANDLING & Y N Expen i REACTOR SERVICES Panel 33 FPC FUEL POOL COOLING Y N Expen Panel 34 FPG GASEOUS FIRE Y N Expen i PROTECTION Panel l 35 FPS HRE BARRIERS & Y N Expen DETECTORS Panel 36 GH2 GASEOUS HYDROGEN Y N Expen Panel 37 GRW GASEOUS RADWASTE Y N Expen Panel 38 H2-MON HYDROGEN Y N Expen MONITORING Panel 39 HTRS HEAT TRACING Y N Expen Panel

, 40 HVAC-CN CONTMT HVAC SYS Y N Expen Panel 41 HVAC-NM NORMAL HVAC -AUX Y N Expen l BLDGS Panel j 42 HVAC-TB TURBINE BLDG. HVAC Y N Expen Panel l.

.. 3 M-Rule, Expert Panel Disposition Results Summary,3/19/96 Page 5 of 6 Bold =STT/STS' rennmine item OpSys Description M-Rule Risk Basis A & B- C-

  1. Code Y/N Y/N PRA/EP Table Table 57 SEIS-IN SEISMIC Y N Expen INSTRUMENTATION Panel SS SRW SOf.ID RADWASTE Y N Expen Panel 59 SUMP NON-RADIOACTIVE Y N Expen pg.gJ 7j -

SUMPS Panel l 4

60 TBN MAIN TURBINE Y N Expen B-2 C-2'8 ) l Panel 61 TGIS TOXIC GAS ISOLATION Y N Expen SYTEM (TGIS) Pane! ,

l 62 TGOV TURBfNE GOVERNOR Y N Expen CONTROLS Panel 63 TPCA TURBINE PLANT CHEM Y N Expen ADD Panel 64 TPCW TURBINE PLANT Y N E.4 pen B-2 C-1 COOLING WATER Panel 65 ADMIN ADMIN BUILDINGS AND N N Expen FACILITIES Panel 66 DWS DOMESTIC WATER N N Expen Panel 67 LLRWS LOW LEVEL RADWASTE N N Expen STORAGE FACILITY Panel l 68 MISC MISCELLANEOUS N N Expen Panel 69 PGRM " PROGRAM" STARTUP N N Expen SYSTEM CODES, NO ID'S Panel 70 SAN SANITARY WASTE N N Expen Panel l

{

L_______-____

b ,, ATmcKMM 4 a

Maintenance Rule l Scoping Summary Matrix Introduction This document provides the Scoping Summary Matrix for the Maintenance Rule Program.

10 CFR 50.65 and Regulatory Guide 1.160 " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" requires monitoring the performance or condition of structures, systems, or components, against licensee established goals, in a manner sufficient to provide reasonable assurance that they are capable offulfilling their intended functions.

The scope of the monitoring pregram includes safety related and non-safety related structures, systems, or components, as follows:

1.

Safety related structures, systems, or components that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe condition, and the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the 10 CFR part 100 guidelines.

2. Non-safety related structures, systems, or components :

a.

That are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures, or

b. Whose failure could prevent safety related suuctures, systems, or components from fulfilling their safety related functions, or
c. Whose failure could cause a reactor scram or actuation of a safety related system.

The Scoping Summary Matrix contains the determination of those structures, systems, and components which must be addressed under the Maintenance Rule. This document is the output of procedure SO123-XIV-5.3.1, " Scoping for the Maintenance Rule."

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%. 1 ATTAMMM G Maintenance Rule <

Scoping Summary Matrix Introduction This document provides the Scoping Summary Matrix for the Maintenance Rule Program.

10 CFR 50.65 and Regulatory Guide 1.160 " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" requires monitoring the performance or condition of structures, systems, or components, against licensee established goals, in a manner sufficient to provide reasonable assurance that they are capable of fulfilling their intended functions.

The scope of the monitoring program includes safety related and non-safety related stmetures, systems, or components, as follows:

1. Safety related structures, systems, or components that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe condition, and the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the 10 CFR part 100 guidelines.
2. Non-safety related structures, systems, or components :

a.

That are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures, or

b. Whose failure could prevent safety related structures, systems, or components from fulfilling their safety related functions, or
c. Whose failure could cause a reactor scram or actuation of a safety related system.

The Scoping Summary Matrix contains the determination of those structures, systems, and components which must be addressed under the Maintenance Rule. This document is the output ofprocedure SO123 XIV-5.3.1, " Scoping for the Maintenance Rule."

j i

This change updates the Matrix to include Containment Isolation and the Steam Generators and includes revisions for the major re-scoping effort for the Low Safety Significant i systems. Some of the High Safety Significant systems were also revised to reflect LSS scoping.

~

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Those systems which are indicated as part of another system have their values removed. The I table has also been resorted by OpSys order.

I STS-SO123-2001, Rev. I i

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  • ATi%cK4 txT 6 l

l MEETING MINUTES Maintenance Rule Expert Panel <

October 17,1997 The follcwing are the minutes from the Maintenance Rule Expert Panel Meeting held October 17,1997 at 1030 Hrs. in conference room D3A.

The following. individuals were present at Ge meeting:

Votine Members Other Attendees Rick Allen, STS Michele Carr, NOD Tom Hook, NOD Kevin Flynn, STEC Steve Atkins, STEC Deau Goodwin, STS Neil Bloom, MAINT John Ramsdell, STS Dave Schafer, STS Sam Chien, NOD l

Marty Cooper, OPS Torrey Yee, NEDO Containment Isolation System Analysis Renort Dave Schafer passed out an updated draft of the Containment Isolation System (CIS) Analysis Report to the MREP members. Mr. Schafer covered the recent updates with the MREP, which included changes to the functions, and the failure criteria.

Due to prior commitments, Station Technical was unable to have Murray Jennex attend this meeting. It was determined that a final vote should not take place without more technical information being supplied to the MREP. Station Technical agreed to try to get Mr. Jennex at the next meeting. Rick Allen shared the results of his informal benchmarking on the issue.

He talked to Mr. Dan Rains ofINPO, who was instrumental in the development of the Maintenance Rule while at NEI. Mr. Rains indicated that a functional failure should not be counted unless we loose both containment isolation valves in a line at the same time.

Even though the vote was delayed, the MREP continued to discuss the CIS Analysis Report.

It was determined that at least a 36 month historical review would be needed. The subject of blind flanges was also discussed. There is a Tech Spec requirement to have blind flanges installed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if they are needed.

The PRA Group gave an update of the efforts to provide PRA based performance criteria recommendations for CIS. During the previous week, Mr. Schafer had provided them with the equipment scope of the system. Sam Chien indicated that he expected to have the numbers by the next MREP.

I h

I october 29,1997 san Onofre Nuclear Generating station Page I of 3

?* .

The MREP agreed on a preliminary scope for the system:

4

- CIS valves in Appendix J

- Includes personnel and equipment hatches

- Does not include some mechanical penetrations, and all electrical ones.

SSC Candidates for Return to (a)(2) Status A handout was distributed (Attachment 1) that identified 3 SSC's currently in (a)(1) Goal Setting which had met their goals and should be considered for return to (a)(2) status. The three SSC's were: l

- Fire Pump MP222 (AR 960601362)

- ESF Swithgear Room Normal Cooler 3ME430 (AR 961000946)

- Aux Feedwater, Unit 2, Swing Train (AR 970401178)

After discussion, the MREP voted to return all three candidates to (a)(2) status.

Low Safety Significant (LSS) Performance Criteria Rick Allen discussed the 7 action items that were created from the Low Safety Significant (LSS) Performance Criteria meeting held on October 4th. The following is a list of the items, and their current status:

1.

Resolve differences in wording in the PRA sections of the LSS system scoping documents. (NOD /STS/ ERIN)

- Complete i

2. Ensure LRW components supporting the BPS function are mapped to that function.

(STS/ ERIN)

- Working 3.

Review PRA flood analysis to verify that Non-radioactive Sumps and Drains should not be within the scope of the Rule. (NOD)

- Not credited, and should not be in the scope of the Rule.

4. I Determine if the AVR HVAC function of the Turbine Bldg HVAC system should be in the scope of the Rule. (STEC)

- Determination made that it should not be included in the scope of the Rule.

5.

Incorporate the safety related MUD berm into the structures program. (NEDO)

- Working October 29,1997 san Onofre Nuclear Generating station Page 2 of 3 l

1

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6.

Rewrite LSS system scoping documents and out of scope documents to reflect that some components may be listed in the E01's, but they are not significant contributors to mitigating the event. (STS/ ERIN)

- Under review 7.

Determine method of storing the out of scope documents either in CDM or MOSAIC.

(STS)

I - Working l

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October 29.1997 San Onofre Nuclear Generating Station page 3 or3 l

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$? *

  • Q w/

SSC Candidates for Return to (a)(2) Status AR # 960601362 Fire Pump SA230lMP222

Description:

FIRE PUMP SA230lMP222 HAS FAILED TO MEET ITS AVAILABILITY PERFORMANCE CRITERIA OF 90%. DURING THE PAST 12 MONTHS THE AVERAGE AVAILABILITY FOR THIS CHANNEL HAS BEEN 89.8%.

Goals Set:

The goal for Fire Pump P-222 is too re-achieve the availability performance criteria of 90% over four quarters. As of the date of this report, the pump has already returned above the acceptable availability performance criteria but will not be evaluated for return to (A)(2) until the next work window planning and implementation sequence is complete.

Current Performance: 19 Hours of Unavailability with a performance criteria of 876 Hours.

AR # 961000946 ESF Switchgear Room Normal Cooler 3ME430

Description:

THE UNIT 3 HVAC-SG, ME430, FAILED TO MEET ITS AVAILABILITY PERFORMANCE CRITERIA OF 98.0% FOR THE FOUR QUARTERS ENDING SEPTEMBER 30,1996. THE OBSERVED AVAILABILITY FOR THIS PERIOD WAS 81.6%.

Goals Set: The goal for this cooler is to achieve the availability criteria of 98% over 4 t

quarters. The expectation is that the average availability will exceed the {

I performance criterion during the third quarter of 1997.

Current Performance: 0 Hours of Unavailability with a performance criteria of 175 Hours.

AR # 970401178 Unit 2 Aux Feedwater Swing Train

Description:

THE UNIT 2 AUXILIARY FEEDWATER (AFW) SYSTEM SWING TRAIN FAILED TO MEET ITS UNAVAILABILITY PERFORMANCE CRITERIA OF

< 180 HOURS (2%) FOR THE FOUR QUAPsTERS ENDING MARCH 31, 1997. THE OBSERVED UNAVAILABILITY FOR THIS PERIOD WAS 182 HOURS.

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Goals Set: The goal for this system is to achieve the unavailability criterion ofless than 3%

over 4 quarters. No additional goals are required to achieve acceptable performance. Since most of the hours that contributed to this unavailability occurred 2Q96. the train unavailability should decrease to under 3% during 2Q97.

Current Performance: 56 Hours of Unavailability with a perfonnance criteria of 169 Hours.

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4TihC MCAIT 7 San Onofra 2&3 FSAR

. Updsted TRANSIENT ANAI,YSIS f

l The offsite radiological doses for the Primary Sample or'Instrumenc Line Break k j with an accident-induced iodine spike are "a small fractien" (i.e., do not l exceed lot) cf the 10 CFR 100 exposure guidelines, and the Control Room

] radiological deses are within the 10 CFR 50 Appendix A General Design l Criterion 19 exposure guidelines.

I l 15.10.6.3.2 stesm cen rater Pube puntura with cencurrant railure of a sinele l Active Cerrenent 1

l Introduction I

l A Steam Generator Tube Rupture (SGTR) event is a penetration of the barrier l 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 l contaminate the secondary side. The radioactivity is released via the l condenser air ejectors, the Main Steam Safety valves (MSSVs), and the l Atmospheric Dump Valves (ADvs).

l l This event was previously analy:ed with and without a single failure. Both l situations are classified as limiting faults. The worst case is for the single l f ailure of a less of normal AC power which increases the radiological release l to the environment (see section 15.6.3.2.1.5). It is this analysis whien is l presented below.

l l If the primary to secondary leak is beyond the capacity of the charging pumps, j the reactor will eventually trip en a low pressure trip signal. As a result of -

l the loss of A.C., the electrical power would be unavailable for the station f l auxiliaries such as the Reactor Coolant Pumps (RCPs) and the Main Feed Water

] (MFW) pumps. Under such circumstances, the plant would experience a l simultaneous loss of lead, normal feed water flew, forced reactor coolant flow l and steam generator blow down capability.

I l 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  ;

) icsence of forced reactor coolant flew, cenvective heat transfer is supported j by natural circulation reactor coolant flow. Initially, the liquid inventory l in the steam generators is used and the resultant steam is v? leased to the l atmosphere via the Main Steam Safety Valves (MSSVs). With the availability of l stand-by pcwer provided by the autematic start-up of the diesel generators, l Auxiliary (emergency) Feed Water (AFW) flow is initiated on a low steam l generator level signal.

l When the reactor plant has been stabilized in Mode 3, the operator achieves l plant cool down using remotely cperated Atmospheric steam Dump Valves (ADVs). j j The plant is cooled to 350*F at a nominal rate of 75'F/hr. At this time, Shut l l Down Cooling (SDC) is initiated.  !

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1/98 15.10-136 Revision 13

l San Onofre 2&3 FSAR Updated j TRANSIENT ANAI,YSIS The analysis of record conservatively assumes the operato/ action to isolate l the affected steam generator is delayed until 30 minutes after initiation of l the event. The operator's diagnosis of the SGTR event is f acilitated by the l radiation monitors which initiate alarms and signal the existence of abnormal l radioactivity levels.

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Radiation monitors are found in the blow down sample lines from each steam l generator, in the blow down processing system neutralization sump discharge l l sea line which processes blow down from both steam generators, and in the l f condenser air ejector discharge line. Additional diagnostic information is j ]

provided by RCS pressure and pressurizer level response indicating a loss of l J primary coolant. Level in the affected steam generator increases as the l primary fluid enters the steam generator driven by the substantially higher l primary pressure.

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

Summarv of Methods l

1 l The CESEC-III code is used to simulate the transient for the first 1800 l seconds (i.e., 30 minutes). The output of the code provides the amount of l

primary to secondary leak, the amount of steam transported from the steam l generators through the MSSVs and the overall Nuclear Steam Supply System l (NSSS) respense to the event. This information is then used to derive the l radiological releases and accompanying doses. l 1

This analysis is primarily performed to establish the parameters, such as the l primary to secondary mass transferred during the event, by which the l radiological releases are calculated. There is no specific acceptance criteria i for the mass releases.

l l

One computer case was run for this analysis. This was an 1800 seconds l CESEO-III simulation of a double-ended SGTR in the right hand (arbitrary l designation) steam generator. This case utili:es a 15% MSSV blow down model to l determine the impact on steam released to atmesphere. l l

The primary transient analysis inputs and assumptions for the analysis are l presented below in Table 15.10.6.3.2-1. The sequence of events is provided in l Table 15.10.6.3.2-2. l 1/98 15.10-137 Revision 13 ,

l

.q San Onofre 2&3 FSAR

,,, Updated TRANSIENT ANALYSIS

.. 3 l The dose methodology for this event is described in Appendices'15B.and .

l 15.10.B. Using this methodology, design basis 0-2 hour Exclusion Area l Boundary, 0-30 day Low Population Zone, and 0-30 day Control Room doses were l calculated with and without consideration of pre-existing and accident induced l iodine spikes.

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

I l 1. Reactor coolant releases via the ruptured tube into the affected j l steam generator, and eventt.? ly to the outside environment.

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

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

l l 4. Turbine-driven auxiliary feed water (ATW) pump venting of secondary l steam from the affected and steam generators to the outside j environment.

I l S. Atmospheric dump valve (ADV) releases of secondary steam from the l intact steam generator to the outside environment.

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! l The principal assumptions and inputs for the dose analysis are presented belew -

l in Table 15.10.6.3.2-3. k'"

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1/98 15.10-138 Revision 13 l l

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't San Onofre 263 FSAR Updated TRANSIENT ANALYSIS ,

1 Table 15.10.6.3.2-1 <

P,inciple Assumptions and Inputs for SGTR, Cycle 5 Parameter Unit 2 Unit 3 l Core Power 3478 Mwth 3478 Mwth l Inlet Temperature 553*r 553*r l l

l RCS Pressure 2400 psia 2400 psia l SG Pressure 900 psia -

900 psia l Core Flow, Total 424,000 gpm 424,000 gpm l Physics Parameters Boc Doppler x 0.85 x 0.85 MTC (Uncertainty) (Uncertainty) i SCRAM Worth -3. 3 x 10-* Ao/

  • F -3. 3 x 10-* Ap/ *r

-6.0 t AD -6.0 % ao CPC (range - icw pressure) Trip Set Point 1785 psia 1785 psia l Loss of A.C. Power Coincident with Coincident with Reactor Trip Reacter Trip l Steam Generator (S/G) U-Tube Break Size Double Ended Double Ended i

Guillotine Guillotine l

Safety Injection Actuation System - Set 1560 psia 1560 psia Point High Pressure Safety In3ection - Response 31.2 seconds 31.2 seconds l Time Main Teed Water (MrW) - Flow Rate 102% of Design 102% of Design l Main reed water (MrW) - Enthalpy 425 Btu /lbm 425 stu/lbm (pre-trip) (pre-trip)

Auxiliary Feed Water (ATW) - Response Time 52.7 secs. 52.7 secs. l electric electric 1 42.7 secs. 42.7 sees. I Steamer Steamer I Auxiliary Feed Water (AFW) - Flow Rate 4% of MTW 4% of MFW Design

  • Design
  • j Auxiliary Feed Water (AFK) - Enthalpy 48 Btu /lbm** 48 Btu /lbm** l Charging Tiow Rate 128 gpm*** 128 gpm*** l I

Let Down Flow Rate 0.0 gpm 0.0 gpm l l 1/98 15.10-139 Revision 13 l

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San Onofre 2&3 FSAR Updated t

, TRANSIENT ANALYSIS ,...

Table 15.10.6.3.2-1 (continued)

Principle Assumptions and Inputs for SGTR, Cycle 5 j Parameter Unit 2 Unit 3 Main Steam Safety Valves (MSSV) - cpening 1089 to 1143 1089 to 1143 Set Points psia psia (9 valves at 7 psi increments.

Includes 11 set point tolerance) l MSSV Accumulation Set Point 3% Rated Flow 3% Rated riew MSSV Blow Down 15% Rated Flow 15% Rated riew (to fully close) (to fully close) l Atmospheric Dump valves (ADVs) Inoperative Inoperative Feed Water Control System (FWCS) I Not Required to- Not Required to

( Mitigate Event Mitigate Event Pressurizer Pressure Centrol System (PPCS) Not Required to Not Required to Mitigate Event Mitigate Event l Steam Bypass Control System (SBCS) Inoperative Inoperative This ATW flow is consistent with cycle 5 analysis. Ard flows ;r500 gpm i-l are adequate to maintain secondary heat sink. _

AFW enthalpy of up to 60 BTU /LBM has an insignificant impact on analysis '

results.

Cycle 5 analysis was performed using 128 gpm. Later analysis evaluated acceptability to 135 gpm.

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1/98 15.10-140 Revision 13 I

- _ _ _ . - _ _ _ - _ _ _ _ _ __ _ _ - _ _ _ - - - _ _ _ - . _ I

San Onofra 2&3 FSAR Updated TRANSIENT ANALYSIS 4

Table 15.10.6.3.2-2 Sequence of Events' for SGTR, Units 2 and 3, Cycle 5 Time (Seconds) Chronological Event Set Point or Value 0.0 S/G Tube rupture occurs ----

l 520.6 Pressurizer Heaters de-energired 334 ft' ,j CPC Reactor Trip (low system pressure),

l 985.1 Turbine stop Valves close, 1785 psia l

Loss of Normal AC 985.5 CEAs begin to drcp into the reactor core ----

l 991.3 MSSVs begin to open on both S/Gs 1099 psia j j.

995.3 Maximum S/G pressure on both generators 1131 psia l 1000.0 Pressurizer empties ----

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W O.2 8"f**Y I")***i*" A*t"*ti*" 31 9 ""1 I3IA3I 1

initiated 1560 psia 1020.9 Low S/G 1evel signal generated 134,540 lbm \

1073.8 Auxiliary Feed Water (ADT) flow initiated ----

l 1305.6 MSSVs close en both S/Gs 926 psia l Damaged S/G isolated, 1800.0 ADV on unaffected S/G opened to begin system cool ----

down to Shut Down Cooling (SDC)

Shut Down Cooling (SDC) initiated 11242.0 Temperature 350*F Total steam release 683,653 lbm l

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l 1/98 15.10-141 Revisien 13 l

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San Onofra 2&3 FSAR

, Updated TRANSIENT ANALYSIS Table 15.10.6.3.2-3 t

C'.

Principal Assumptions and Inputs for Steam Generator Tube Rupture Dose Analysis

\

l Parameter Unit 2 Cycle 9 Unit 3 Cycle 9 i Single Failure Loss of AC Loss of AC Power Power RCS Iodine Activity (Dose Equivalent I-131), 1.0 1.0 gCi/gm Increase in Iodine Release Rate from Fuel 500 500 j for Accident Induced Iodine Spike RCS Pre-Existing Iodine Spike Iodine '0 6 60 Activity (Dese Equivalent I-131), yCi/gm l RCS Non-Iodine Activity, yC1/gm 100/E 100/E Secondary Liquid Iodine Activity (Dese 0.1 0.1 Equivalent I-131), yCi/gm l Steam Generator Iodine Partition Coefficient 0.01 0.01 Primary to Secondary Leak Rate into each SG, 0.5 0.5 l gpm Integrated primary to secondary rupture 1 800 ecends) 3,323 73,323 Portion of Primary Leak Flow Which Flashes 40% 40%

to Steam Integrated MSSV flow, lbm (1,800 secends)

LH - Unaffected 43,642 43,642 RH - Affected 43,703 43,703 Total MSSV Flow 87,345 87,345 AIW Flow (steam driven pu=p), lbm (1,800 8,071 8,071 seconds) l Steam Release (30 - 120 minutes), lbm 336,420 336,420 l Total steam release (0 - 120 minutes), lbm 431,836 431,836 Total steam release to Shut Down Cooling,' 638,653 683,653 lbm l Control Rocm Isolation Signal High Radiation High Radiation l Control Room Isolation Time, min 3 3 offsite Dose Evaluation Model Appendix 15B Appendix 15B l

and Appendix and Appendix 15.10B 15.10B Control Room Dose Evaluation Model Appendix 15B Appendix 15B and Appendix and Appendix 15.10B 15.10B i

1/98 15.10-142 Revision 13 l

San Onofra 2f3 FSAR Updated TRANSIENT ANALYSIS Results l

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l The primary to secondary rass transfer and steam releane data required to j perform radiological calculations for the steata generator tube rupture event l are presented in Table 15.10.6.3.2-1. j i

The minimwm DNBR during the transient was not explicitly evaluated for this l analysis. The work evaluates the impact of the extended blow down of the MSSVs l on the parameters which are important to the evaluation of radiological dose l release. Violation of the DNER SATDL was judged not to be affected by the l change in the MSSV modeling since the time of minimum DNER precedes the times l at which the MSSVs epen and close. l l

The RCS and secondary system pressures remain below the 110% of the design l j pressure lindts, thus, assuring the integrity of these systerm. l t

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, The analysis work reported in this section was performed for both Units 2 and l

! 3 Cycle 5. This work has been reviewed and deternined to be bounding for Units l i

j 2 and 3 Cycle 9 operation in all respects..As part of the cycle 9 work, the l {

l SGTR event has been evaluated to bound the use of the new +2% and -3% MSSV set l l point tolerances.

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

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s 1/98 15.10-143 Revision 13 1

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San Onofre 2&3 FSAR Updated i

i TRANSIENT ANALYSIS ,

l Table 15.10.6.3.2-4 l Results for Steam Generator Tube Rupture p, Acceptance Analysis Results Criteria Unit 2 l Unit 3 l Design Basis Case with No Iodine Spike

] 0-2 hr EAB Doses, Rem l Thyroid 30 <0.1 <0.1 l Beta Skin N/A l Whole Body

<0.1 <0.1 2.5 <0.1 <0.1 l 0-30 day LP: doses, Rem l Thyroid 30 <0.1 <0.1 l Beta Skin N/A l Whole Body

<0.1 <0.1 2.5 <0.1 <0.1 l 0-30 day Control Room doses, Rem l Thyroid 30 0.2 0.2 l Beta Skin 30 0.5

)

0.5 )

l Whole Body 5 <0.1 <0.1 l Design Basis Case with Pre-Existing Iodine Spike l 0-2 hr EAB Deses, Rem l Thyroid 300 2.8 2.8 l Beta Skin N/A <0.1 <0,1 l Whele Body 25 <0.1 <0.1 l 0-30 day LP: doses, Rem l Thyroid 300 <0.1 <0.1 l Beta Skin N/A Whole Body

<0.1 <0.1 l 25 <0.1 <0.1 l 0-30 day control Room doses, Rem l l Thyroid 30 0.7 0.7 l Beta Skin 30 0.5 0.5 l Whole Body 5 <0.1 <0.1 l Design Basis Case with Accident Indu~.d Iodine Spike l 0-2 hr EAB Doses, Rem l Thyroid 30 0.4 0.4 l Beta Skin N/A Whole Body <0.1 <0.1 l 2.5 <0.1 <0.1 l 0-30 day LP: doses, Rem l Thyroid 30 i

<0.1 <0.1 l l Beta Skin N/A <0.1 <0.1 l Whole Body 2.5 <0.1 <0.1 l 0-30 day Control Room doses, Rem l Thyroid 30 0.2 0.2 l Beta Skin 30 0.5 0.5 l l Whole Bodv 5 <0.1 <0.1

("

1/SB 15.10-144 Revision 13

Sen Onofre 2&3 FSAR

. Updated .

TRANSIENT ANALYSIS When the reactor is off line, stored energy and fission product decay energy l must be dissipated by the reactor coolant and main steam systems. In the l 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 Main Steam safety Valves (MSSys). With the availability of stand-by power provided by the automatic start-up of the diesel generators, Auxiliary (emergency) Feed Water (ATW) flow is initiated on a low steam i generator level signal.

I When the reactor plant has been stabilized in Mode 3, the operator achieves plant cool down using remotely operated Atmospheric steam Du=p Valves (ADVs).

The plant is cooled to 350*F at a nominal rate of 75*F/hr. At this time, Shut Down Cooling (SDC) is initiated.

l 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 l radiation monitors which initiate alarum and signal the existence of abnormal radioactivity levels.

l Radiation monitors are found in the blow down sample lines from each steam generator, in the blow down processing system neutrali:ation su=p discharge j

sea line which processes blow down from both steam generators, and in the condenser air ejector discharge line. Additional diagnostic information is provided by RCS pressure and pressurizer level response indicating a less of primary coolant. Level in the affected steam generator increases as the l j

t primary fluid enters the steam generator driven by the substantially higher primary pressure.

The offsite and control roem 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 l l the reactor coolant.

l 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 l (NSSS) response to the event. This information is then used to derive the radiological releases and acco=panying doses.

l l 1/98 15.10-145 Revision 13 l

t ___

San Onofra 2&3 FSAR

, Updated TRANSIENT ANALYSIS This analysis is primarily performed to establish the parameters, such as the C 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 s monds CESEC-III simulation of a double-ended SGTR in the right hand (a b.ttrary designation) steam generator. This case utilizes a 15% MSSV blow down model to determine the impact on steam released to atmosphere.

I The primary transient analysis inputs and assumptions for the analysis are

! l presented below in Table 15.10.6.3.2-1. The sequence of events is provided in l Table 15.10.6.3.2-2. .

The dose methodology for this event is described.in Appendices 15B and l 15.10.B. Using this methodology, design basis 0-2 hour Exclusion Area i

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 five release mechanisms that can disperse radioactive material I into the atmosphere have been evaluated:

l f

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

l

4. Turbine-driven auxiliary feed warer (AFW) pump venting of secondary steam from the affected and steam generators to the outside environment.

l l 5. Atmospheric dump valve (ADV) releases of secondary steam from the intact steam generator te the outside environment.

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

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1/98 15.10-146 Revision 13 1

San Onofra 2&3 FSAR

, Updated TRANSIENT AIALYSIS 6

l Table 15.10.6.3.2-1 l Principle Assumptions and Inputs for SGTR, Cycle 5 Parameter Unit 2 Unit 3 l Core Power 3478 Mwth 3478 Mwth l Inlet Temperature 553 *T 553 *F l RCS Pressure I 2400 psia 2400 psia l l SG Pressure 900 psia 900 psia j i Core Flow, Total 424,000 gpm 424,000 gpm l Physics Parameters Boc Deppler x 0.85 MTC x 0.85 (Uncertainty) (Uncertainty)

SCRAM Worth -3.3 x 10-* Ac/*F -3.3 x 10'* Ao/*F

-6.0 L 40 -6.0 % ap CPC (range - low pressure) Trip Set 1785 psia 1785 psia Point Loss of A.C. Power Coincident with Coincident with Reactor Trip Reactor Trip Steam Generator (S/G) U-Tube Break Size Double Ended Double Ended Guillotine Guillotine Safety Injection Actuation System - Set 1560 psia 1560 psia Point High Pressure Safety Injection - 31.2 seconds 31.2 seconds Response Time Main reed Water (MFW) - Flow Rate 102% of Design 102% of Design l Main Feed Water (MFW) - Enthalpy 425 Btu /lbm 425 Btu /lbm I (pre-trip) (pre-trip) l Auxiliary Feed Water (ArW) - Response 52.7 secs. 52.7 secs.

TLme electric electric 42.7 secs. 42.7 secs.

Steamer Steamer Auxiliary Feed Water (AFW) - Flow Rate 4% of MFW 4% of MFW Design

  • Design
  • Auxiliary Feed Water (ATW) - Enthalpy 48 Btu /lbm** 48 Btu /lbm?* l Charging Flow Rate 128 gpm*** 128 gpm*** l Let Down Flow Rate 0.0 gpm 0.0 gpm l 1/98 15.10-147 Revision 13

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San onofra 263 FSAR Updated ,

TRANSIENT ANALYSIS Table 15.10.6.3.2-1 (continued) f '

Principle Assumptions and Inputs for SGTR, Cycle 5 l Parameter Unit 2 Unit 3 Main Steam Safety Valves (MSSV) - 1089 to 1143 1089 to 1143 Opening Set Points psia psia (9 valves at 7 psi increments.

Includes 1% set point tolerance) l MSSV Accumulation Set Point 3% Rated Flow 3% Rated Flow MSSV Blow Down 15% Rated Flow 15% Rated Flow (to fully close) (to fully close) l Atmospheric Dump Valves (ADVs) Inoperative Inoperative Feed Water Control System (T4CS) Not Required to Not Required to Mitigate Event Mitigate Event Pressurizer Pressure Control System Not Required to Not Required to (PPCS) Mitigate Event Mitigate Event l Steam Bypass Control System (SBCS) Inoperative Inoperative This AIW flow is consistent with cycle 5 analysis. AFW flows 2500 gpm l are adequate to maintain secondary heat sink. i i

AFW enthalpy of up to 60 BTU /LBM has an insignificant impact on analysis results.

Cycle 5 analysis was performed using 128 gpm. Later analysis evaluated acceptability to 135 gpm.

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1/98 15.10-148 Revision 13

Sen Onofre 2&3 FSAA Updated TRANSIENT ANALYSIS Table 15.10. 6.3.2-2 Sequence of Events for SGTR, Units 2 and 3, Cycle 5 Time (Second Chronological Event

  • int or s) y,y ,

0.0 S/G Tube rupture occurs ----

l 520.6 Pressurizer Heaters de-energized 334 ft8 l CPC Reactor Trip (low system pressure),

985.1 Turbine stop valves close, 1785 psia Loss of Normal AC 985.5 CEAs begin to drop into the reactor core ----

l 991.3 MSSVs begin to open on both S/Gs 1089 psia l 995.3 Maximwn S/G pressure on both generators 1131 psia l 1000.0 Pressurizer empties ----

l 1010.2 Safety Injection Actuation Signal (SIAS) initiated .1560 psia l-l 1020.9 Low S/G 1evel signal generated 134,540 lbm l l 1073.8 Auxiliary Feed Water (ArW) flow initiated ----

! l.

1305.6 MSSVs close en both S/Gs 926 psia l Damaged S/G isolated, 1800.0 ADV on unaffected S/G opened to.begin system cool ----

down to shut Down Cooling (SDC)

Shut Down Cooling (SDC) initiated

. 11242.0 Temperature 350 'T Total steam release 683,653 lbm 1

1/98 15.10-149 Revisicn 3 3

Sen onofre 243 FSAR Updated TPRISIENT ANAI,YSIS Table 15.10.6.3.2-3 Principal Assumptions and Inputs for Steam Generator Tube Rupture Dose Analysis

~

Parameter Unit 2 Unit 3 Cycle 9 Cycle 9 Single Failure Loss of AC Loss of AC Power Power l RCS Iodine Activity (Dese Equivalent I-131), uCi/gm 1.0 1.0 Increase in Iodine Release Rate f cm Fuel for 500 500 Accident Induced Iodine Spike RCS Pre-Existing Iodine Spike Iodine Activity (Dese 60 60 Equivalent I-131), yCi/gm l RCS Non-Iodine Activity, uCi/gm 100/E 100/E Secondary Liquid Iodine Activity (Dese Equivalent I- 0.1 0.1 131), yci/gm l Steam Generator Iodine Partition coefficient 0.01 0.01 l Primary to Secondary Leak Rate into each SG, gpm 0.5 0.5 Integrated primary to secondary rupture flow, ihm (1,800 seconds) 73,323 73,323 l Portion of Primary Leak Flow Which Flashes to Steam 40% 40%

Integrated MSSV flow, lbm (1,800 seconds)

LH - Unaffected 43,642 43,642 RH - Affected 43,703 43,703 Tetal MSSV Flow 87,345 87,345 l AFW Flow (steam driven pu=p), Ibm (1,800 seconds) 8,071 B,071 l Steam Release (30 - 120 minutes), lbm 336,420 336,420 l Total steam release (0 - 120 minutes), lbm 431,F36 431,836 l Total steam release to Shut Down Cooling, Ibm 638,653 683,653  !

Control Room Isolation Signal High High Radiation Radiation l Control Room Isolation Time, min 3 3 offsite Dose Evaluation Model Appendix Appendix 15B and 15B and Appendix Appendix 15.10B 15.10B Control Room Dese Evaluation Model Appendix Appendix 15B and ISB and Appendix Appendix 15.10B 15.10B 1/98 15.10-150 Revision 13 i

F San Onofra 2&3 FSAR

,. Updated TRANSIENT ANALYSIS Pesults The primary to secondary nass transfer and steam release data required to perform radiological calculations for the steam generator tube rupture event are presented in Table 15.10.6.3.2-1.

The minimwn DNBR during the transient was not explicitly evaluated for _ this analysis. The work evaluates the impact of the extendeo blow down of the MSSVs on the parameters which are important to the evaluation of radiological dose release. Violation of the DNBR SAFDL was judged not to be affected by the

[ change in the NESV modeling since the time of rdnimum DNBR precedes the times I

at which the MSSVs open and close.

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

The analysis work reported in this section was performed for both Units 2 and 3 Cycle 5. This work has been reviewed and detersdned to be bounding for Units 2 and 3 Cycle 9 operation in all respects. As part of the Cycle 9 work, the SGTR event has been evaluated to bound the use of the new +2% and -34 MSSV set point tolerances.

The results of the most recent analysis of the potential off site and control l

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.

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1/98 15.10-151 Revision 13

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.. l San Onofre 2&3 FSAR Updated i

TRANSIENT ANALYSIS Table 15.10.6.3.2-4 .

Results for Steam Generator Tube Rupture Acceptance Analysis Results Parameter Criteria Unit 2 l Unit 3 j l Design Basis Case with No Iodine Spike 0-2 hr EAB 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 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 Rocm doses, Rem Thyroid 30 0.2 0.2 Beta Skin 30 0.5 0.5 Whole Body 5 <0.1 <0.1 l Design Basis Case with Pre-Existing Iodine Spike 0-2 hr EAB Doses, Rem Thyroid 300 2.8 2.8 Beta Skin 'N/A <0.1 <0.1 Whole Body 25 <0.1 <0.1 0-30 day LP: doses, Rem Thyroid 300 <0.1 <0.1 Beta Skin N/A <0.1 <0.1 .

Whole Body 25 <0.1 <0.1 0-30 day Control Room doses, Rem Thyroid 30 0.7 0.7 Beta skin 30 0.5 0.5 Whole Body 5 <0.1 <0.1 l Design Basis Case with Accident Induced Iodine Spike 0-2 hr EAB Doses, Rem Thyroid 30 0.4 0.4 Beta Skin N/A <0.1 <0.1 Whole Body 2.5 <0.1 <0.1 0-30 day LPZ doses, Rem J Thyroid 30 <0.1 <0.1  !

Beta Skin N/A <0.1 <0.1  !

Whole Body 2.5 <0.1 <0.1 l

0-30 day Control Room doses, Rem l Thyroid 30 0.2 0.2 l Beta Skin 30 0.5 0.5 Whole Bodv 5 <0.1 <0.1 l

1/98 15.10-152 Revisici 13 l

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