ML20198L613
| ML20198L613 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/31/1998 |
| From: | Nunn D SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20198L593 | List: |
| References | |
| NUDOCS 9901050071 | |
| Download: ML20198L613 (21) | |
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UNITED STATES OF AMERICA-NUCLEAR REGULATORY C0WIISSION Application of SOUTHERN CALIFORNIA ) ,
-EDISON COMPANY, ET AL. for a Class 103 ) Docket No. 50-361 License to Acquire, Possess, and Use )
a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear ) No. 184 Generating Station )
SOUTHERN CALIFORNIA EDIS0N COMPANY, gt al. pursuant to 10CFR50.90, hereby submit Amendment Application No.184. This amendment application consists of Proposed
' Change No. NPF-10-501 to facility Operating License No. NPF-10. Proposed Change No. NPF-10-501 is a request to revise Technical Specification 3.3.5, "ESFAS Instrumentation," to limit operation ~ with a channel of the Refueling Water Storage Tank Level-Low input to the Recirculation Actuation Signal (RAS) or the Steam Generator Pressure-Low or Steam Generator Pressure Difference-High input to the
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1 Emergency Feedwater Actuation Signal (EFAS) in the tripped condition.
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9901050071 981231 "
PDR ADOCK 05000361 i P Pm ~
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l Subscribed on this ay of u Nbh J~ , 1998. 1 Respectfully submitted, SOUTHERN CALIFORNIA EDIS0N COMPANY l
T l By kt l DwighTE.Nunn\
Vice President State of California County f San. Die o On lY 6I Cf appeared before me llb(LQ MllfM, personally NICt h P h Id lA A.-+ Q personally known to me to be the person whose l
name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the ' person, or the entity upon behalf of which the person acted, executed the instrument.
WITNESS my hand and official seal. -------------
( MARIANE SANCHEZ p Commission #1196482 ,
No*ary Futsc-Collfamla !j g
- 4 Son Diego County My Comm.Exphs Oct14,2002 U ~
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. UNITED STATES OF-AMERICA NUCLEAR REGULATORY COMISSION i
. Application'of SOUTHERN CALIFORNIA )
EDISON COMPANY, ET- AL. for a Class 103 ) Docket No. 50-362- t License:to Acquire, Possess, and Use ) !
- a. Utilization Facility as Part of ) Amendment Application l Unit No. 3 of the San Onofre Nuclear ) No. 170
' Generating Station l i
SOUTHERN CALIFORNIA EDISON COMPANY, g131. pursuant to 10CFR50.90, hereby submit 4
. Amendment Application No.170. This amendment application consists of Proposed Change No. NPF-15-501 to Facility Operating License No. NPF-15. Proposed Change No.. NPF-15-501 is a request to revise Technical Specification 3.3.5, "ESFAS Instrumentation," to limit operation with a channel of the Refueling Water Storage Tank Level-Low input to the Recirculation Actuation Signal (RAS) or the Steam Generator Pressure-Low or Steam Generator Pressure Difference-High input to the Emergency Feedwater Actuation Signal (EFAS) in the tripped condition.
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l Subscribed on this day of d fd k b V , 1998.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY I By .L DwightE.Nunn)
Vice President State of California p County 9fSanDie l 'l On W 9} 7 8 before me, q!kO [Ld M M onally appeared b Ab % N b k_ , ersonally known to me to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument. i WITNESS my hand and official seal.
'^^^^^Q^^^'^^ cHH Commission #1196482 h Notary PuoGc- Cafifamio l
/ q Son Diego County 7
[ l / 4 MyComm.Emphz Oct14,2CD2 >
Signatu k_) b 7 /~
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ENCLOSURE 1 PCNNPF-10/15-501 Operation with RAS or EFAS in Tripped Condition l
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DESCRIPTION OFPROPOSEDCHANGENPF-10/15-501 OPERATION WITH RAS OR EFAS IN TRIPPED CONDITION SAN ON0FRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 This proposed change is a request to revise Technical Specification (TS) 3.3.5, "ESFAS Instrumentation," to limit operation with a channel of the Refueling Water Storage Tank Level-Low input to the Recirculation Actuation Signal (RAS) or the Steam Generator Pressure-Low or Steam Generator Pressure Difference-High
.inpu,.t to the Emergency Feedwater Actuation Signal (EFAS) in the tripped l condition. I i
Existing Technical Specifications Unit 2: See Attachment "A" Unit 3: See Attachment "B"
-Proposed Technical Specifications Unit-2: See Attachment "C" (Redline and Strikeout shown)
Unit 3:
See Attachment "D" (Redline and Strikeout shown) '
Proposed Technical Specifications Unit 2: See Attachment "E" Unit 3: See Attachment "F" Proposed Bases (for information only)
L Unit 2: See Attachment "G" l Unit 3: See Attachment "H" DESCRIPTION OF CHANGES:
l This proposed change is to revise Technical Specification 3.3.5, "ESFAS l Instrumentation," to provide new Conditions and Actions for the Refueling Water Storage Tank (RWST) Level-Low input to the Recirculation Actuation Signal (RAS) and the Steam Generator (SG) Pressure-Low and SG Pressure Difference-High input
( to the Emergency Feedwater Actuation Signal (EFAS). For these three inputs, operation with one channel inoperable will be allowed only if the channel is
.placed in bypass. Operation with two channels of either of these functions will
- be allowed for up to 7 days with one channel in bypass and one channel in trip.
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Currently, these two Functions are covered by existing Conditions 3.3.5.A and 3.3.5.B which cover all Emergency Safety Features Actuation System (ESFAS)
Functions. The.specified input channels for RAS and EFAS are being removed from these existing Conditions _and placed into separate TS 3.3.5 Conditions. Because of the undesireable consequences of spurious trips for these two signals, more restrictive Completion Time limits are required which will maintain the signals in a single-failure-proof condition.
BACKGROUND In June of 1997, another utility identified a scenario in which a single failure of a channel of the Refueling Water Tank (RWT) Level - Low signal could lead to a spurious RAS actuation with unacceptable results.
A review of the San Onofre Plant Protection System (PPS) confirmed that the same deficiency exists for RAS. In addition, this review determined that a spurious EFAS could lead to undesireable consequences as well. Since this time, San Onofre Units 2 and 3 have been operating under administrative limits which allow the RAS and EFAS functions to' remain OPERABLE. This proposed change will provide the final corrective action /or this situation.
ESFAS Initiation ESFAS initiation is based on a two-out-of-four logic. The ESFAS has a design feature which allows one out of the four logic measurement channels to be placed in bypass. When a channel is in bypass the logic converts to a two-out-of-three configuration logic. With two channels in bypass, an ESFAS function would be in a two-out-of-two configuration, and a single failure of one of the remaining channels could result in failure to actuate when necessary. Therefore, by design the ESFAS permits only one channel at a time of any one function to be placed in bypass. If one channel is inoperable, TS 3.3.5 allows the plant to be operated with the inoperable channel in bypass or trip. Because only one channel can be placed in bypass, if two channels are inoperable then one channel must be placed in bypass and the other channel in trip. This places the ESFAS function in a one-out-of-two mode of operation and is allowed by the current TS 3.3.5 for an indefinite period.
With one channel in a tripped condition it takes only one additional channel trip, either due to channel failure or in response to actual plant conditions, to initiate an ESFAS signal. Previously, a spurious signal due to channel failure had been considered failure in a safe condition, and therefore operation with one channel in trip was acceptable. It has now been determined that there is a scenario in which a spurious RAS actuation could lead to unacceptable results.
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RAS Actuation The purpose of RAS is to change the mode of operation of the-Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) from the injection phase to the recirculation phase. At the onset of the recirculation phase the RAS automatically stops the Low Pressure Safety Injection (LPSI) pumps and shifts the High Pressure Safety Injection (HPSI) pumps suction from the
- Refueling Water Storage Ta'nk (RWST) to the Containment Emergency Sump. The Containment Emergency Sump then supplies the inventory requirements for the ECCS pump suctions. The RAS is designed to occur when the inventory of the RWST is nearly depleted, ensuring the containment sump will have adequate inventory for the recirculation phase.
RAS ~ Actuation causes the following to occur:
- 2. ECCS miniflow valves close, and
- 3. LPSI pumps trip. ;
During a LOCA, a' spurious RAS actuation (before the RWST level has reached the Low Level setpoint) will actuate the same equipment with the exception of
- closure of the ECCS miniflow valves. There is an interlock that prevents the miniflow valves from closing unless there is sufficient level in the containment j emergency _ sump.
In this scenario, the pressure in containment will be sufficient to drive the
, HPSI pump suction until the header is pumped dry, and air and steam from the
! containment gas will bind the pumps, potentially damaging them. Because the L miniflow valves remain open, the miniflow lines, which return to the RWST, l effectively breach containment integrity.
EFAS Actuation A review of other ESFAS functions was performed to determine if problems with spurious signals existed similar to RAS. This review identified that a spurious l- SG Pressure-Low or SG Pressure Difference (SGPD) - High input to EFAS also led to undesireable consequences.
- EFAS is comprised of two trains (EFAS-1 and EFAS-2) that provide Auxiliary feedwater (AFW) to the Steam Generators (SGs). EFAS-1 controls the AFW to Steam 4
Generator 1 and EFAS-2 controls the AFW to Steam Generator 2.
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EFAS Actuation.causes the following to occur:
-1.- Starts the AFW pumps
- 2. Determines'whether a SG is intact
'SG, and
valves when the water level is re-established above the low level !
trip setpoint. ;
EFAS-1 is initiated to SG #1 by either a low SG level coincident with no low pressure trip present on SG #1 or by a low SG level coincident with a
' differential pressure between the two generators with the higher pressure in SG
- 1. EFAS-2 is similarly configured to feed SG #2.
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A spurious SG Pressure-Low signal during a Loss of Main Feedwater event- could i i prevent an EFAS signal from being generated to one SG. During this event, SG !
pressure is expected to remain above the SG Pressure-Low setpoint, while level !
decreases in both SGs. EFAS is designed to actuate in this condition, but a l spurious SG Pressure-Low signal would block feedwater to the associated SG, as i the system would treat that SG as ruptured.
, .The SG secondary differential pressure is used, in conjunction with a SG
- l. . Pressure-Low input from each SG, as an input of the EFAS logic to determine if a lt ' generator is intact. The EFAS logic allows feeding a SG if a SG Pressure-Low condition exists in that generator as long as the pressure in that SG is higher than-the pressure in the other SG by the Steam Generator Pressure Difference (SGPD)-High setpoint.
.A spurious SGPD-High signal during a Main Feedwater Line Break (MFLB) or a Main Steam Line Break (MSLB) could cause feeding of a ruptured SG, depleting L Condensate Storage Tank inventory and increasing the load on the Containment l Cooling System.
) DISCUSSION Because of the potential consequences of spurious trips for the RAS and EFAS signals, additional constraints are necessary for channel operability. The current TS 3.3.5 allows operation with one channel in bypass or trip until MODE
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currently allows operation with one channel in bypass AND one channel in trip for all ESFAS functions.
Following identification of this issue, San Onofre Units 2 and 3 placed administrative' limits on channel inoperability for RAS and EFAS signals. These administrative limits allow operation for only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with a channel in trip before requiring a plant shutdown. This reduces the operational flexibility that is available under the current TS 3.3.5. With the current ability to operate with two inoperable channels, if one channel were to fail outright in such a manner that it could not be repaired, continued operation would be possible, even considering the monthly CHANNEL FUNCTIONAL TEST which requires ,
placing a second channel in bypass. Under the current administrative limits for '
RAS and EFAS, this would not be possible. l Southern California Edison (SCE) has performed a risk-informed assessment to determine acceptable limits for operation with a channel of RAS or EFAS in trip.
The proposed 7 day Completion Time is based on the engineering evaluation below. l Assuming that one channel of RWST Level-Low for RAS would be inoperable due to l outright failure, this channel would need to be placed in trip during the l
performance of CHANNEL FUNCTIONAL TESTING of the other three channels of RAS so '
that the channel being tested could be placed in bypass. CHANNEL FUNCTIONAL TESTING is performed monthly, on a staggered basis. l l
Also, for a case where two channels of RWST Level-Low input to RAS have failed l outright, the 7 day Completion Time allows sufficient time to restore one of the {
two channels to OPERABLE status before a plant shutdown is required. The l operating history at San Onofre shows that outright failure of more than one channel of a particular Function occurs less than once per year.
SCE considers it prudent to avoid the undesireable effects of an inadvertent EFAS by minimizing the probability of occurrence. Therefore, the Completion !
Time for a channel of SG Pressure-Low or SGPD-High input for EFAS in trip was conservatively chosen to be equal to that for RAS.
SCE has established and implemented a Configuration Risk Management Program (CRMP) in TS 5.5.2.14 in accordance with prior approved risk-informed TS changes. The Bases for proposed Action 3.3.5.0 requires implementation of TS 5.5.2.14, " Configuration Risk Management Program." The main purpose of the CRMP is to ensure that a proceduralized Probabilistic Risk Assessment-informed process is in place that assesses the overall impact of plant maintenance on plant risk. Implementation of the CRMP enables appropriate actions to be taken or decisions to be made to minimize and control risk when performing on-line maintenance for Systems, Structures, and Components with a r'sk-informed
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Completion Time. TS 5.5.2.14 is applicable to TS 3.8.1 because the Completion
' Times for proposed Action 3.3.5.D is a " risk-informed Completion Time."
ENGINEERING EVALUAT.10N ,
l In accordance with the guidance provided in Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specif.ications," the' following topics are discussed relative to the proposed change:
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Compliance with Current Regulations
- Defense-in-Depth
. Safety Margins I Evaluation of Risk Impact
. Quality of the PRA I
Comoliance with Current Reaulations The proposed change provides additional time limits on operation with a channel l of the RWST Level-Low input to RAS or the SG Pressure-Low or SG Pressure Differential-High inputs to EFAS in trip. Operation in this condition is currently allowed indefinitely. The proposed restrictions reduce the possibility of an inadvertent actuation of RAS or EFAS. The proposed change is l in compliance with all current regulations and orders while meeting all license conditions.
I L Defense-in-Depth I The San Onofre Nuclear Generating Station (SONGS) 2/3 defense-in-depth is unaffected by the proposed change due to the following: (1) no physical change will take place in plant design; (2) the possibility of an inadvertent RAS or EFAS.is reduced; and (3) the current system redundancy, independence, and diversity are maintained. The proposed change provides limits for operating with a channel of the RWST Low-Level input to the RAS in trip as well as the'SG {
result of this change, the potential for an inadvertent actuation of either of
'these two functions is reduced.
L The proposed 7 day Allowed Outage Times (A0Ts) are based on Probabilistic Risk Assessment (PRA) considerations and are conservative compared to the current TS 3.3.5 completion times. Furthermore, the proposed TS change has a small impact i on plant risk in terms of both changes in the core damage frequency (CDF) and the large early release frequency (LERF) according to RG 1.174, "An Approach for
, Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific t
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Changes to the Licensing Basis," as well as in the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP) according to RG 1.177.
Safety Margins This proposed change provides additional time limits on operation with a channel of the RWST Level-Low input to RAS in trip or with a channel of the SG Pressure-Low or SG Pressure Differential-High inputs to EFAS in trip. The proposed 7 day limits are conservative compared to the current TS 3.3.5 requirements, i.e.,
operating indefinitely with one RAS or EFAS channel in bypass and one channel in trip. As a result, the proposed change with more restrictive allowed outage times (A0Ts) has a positive impact on the current plant risk (i.e., reduces CDF and LERF). Therefore, the current plant safety margins are maintained as a result of the proposed risk-informed TS change.
Evaluation of Risk Imoact 1 This section addresses the impact of the proposed change on plant risk following guidelines provided in RG 1.174 and RG 1.177.
RAS Evaluation 1
The San Oncfre Living PRA model contains a detailed plant protection system (PPS) model. For the RAS evaluation, each channel is assessed for four major types of failures. The failure rate of each type of failure was assessed based on the component failure rate and the number of components in the channel (failure rate times the number of components). The events of concern for this scenario are loss-of-coolant accidents (LOCAs) including induced pressurizer pressure safety vaive failure to close during other plant transients.
The inadvertent RAS actuation has unacceptable consequences following a LOCA ev nt, only before a valid automatic RAS actuation occurs. The exposure time for other RAS channel failures is the time interval between initiation of the LOCA event and the switch over time from injection (RWST source) to l recirculation (containment sump source). This time is estimated to be 20 i minutes for a large LOCA to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for medium LOCA based on Modular Accident Analysis Program (MAAP) calculations. Since this time is expected to be longer j for a small LOCA, conservatively a time interval of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is used here. The ]
operators are not expected to remain in an injection phase for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> >
before they initiate shutdown cooling operation. Therefore, the conditional CDF !
reflects an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exposure time for RAS channel random failures. This results l in a conditional CDF of 5.4E-6/yr as a result of inadvertent RAS actuation l following a LOCA.
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In order to estimate the average annual CDF increase and compare it with the l
acceptance criteria stated in RG 1.174, an estimate of the cumulative time that l a RAS channel is in the tripped condition is needed.
Char.nel functional testing is performed monthly, on a staggered basis.
Therefore, a second channel would be tested 3 out of every 4 months, or 9 times l a year for a period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per test. Conservatively assuming a RAS channel is failed for an entire year and must be placed in the trip condition (from the i byass condition) during channel functional testing, cumulative operation with I the increased potential for an inadvertent RAS actuation during channel testing l 1s estimated as follows: l Annual CDF Increase = 5.4E-6 /yr x (9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> /8760 hours / year) l Annual CDF Increase = 5.4E-9 /yr '
According to the acceptance criteria stated in RG 1.174, the proposed change falls in Region 111 for CDF which represents a "very small risk impact" change.
In addition to the risk impact criteria in RG 1.174, RG 1.177 requires that the l risk-informed submittal show that the proposed TS A0T change has only a small quantitative impact on plant ICCDP. RG 1.177 defines a "small" impact on plant risk associated with a single A0T change to be an ICCDP of 5.0E-7 or less. The ICCDP for the proposed A0T of 7 days is calculated as follows:
l ICCDP = 5.4E-6 /yr x 7 days x (1/365 days /yr)
ICCDP = 1.1E-7 According to the acceptance criterion stated in RG 1.177, the proposed change represents a "small risk impact" change with respect to the ICCDP.
The increase in LERF as a result of inadvertent RAS actuation when a channel is in the tripped condition is estimated to be equivalent to the LERF of having a i Safety injection Actuation Signal (SIAS) event concurrent with a loss of injection and recirculation capability times the probability of an inadvertent RAS. Therefore, the conditional LERF is calculated as follows:
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LERF Increase = LERF of SIAS w/o injection / recirculation x Probability of Inadvertent RAS LERFIncrease=1.8E-05/yrx4.8E-3 LERF Increase = 8.6E-8/yr Therefore, the average annual LERF increase as a result of inadvertent RAS actuation when a channel is in the tripped condition is estimated as follows:
Annual LERF Increase = 8.6E-8 /yr x (9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> /8760 hours / year)
Annual LERF Increase = 8.8E-11 /yr According to the acceptance criteria stated in RG 1.174, the proposed change falls in Region III for LERF which represents a "very small risk impact" change.
In addition to the risk impact criteria in RG 1.174, RG 1.177 requires that the risk-informed submittal show that the proposed TS A0T change has only a small quantitative impact on plant ICLERP. RG 1.177 defines a "small" impact on plant risk associated with a single A0T change to be an ICLERP of 5.0E-8 or less. The l
ICLERP for the proposed A0T of 7 days is calculated as follows:
ICLERP = 8.6E-8 /yr x 7 days x (1/365 days /yr) l l
ICLERP = 1.7E-9 l According to the acceptance criterion stated in RG 1.177, the proposed change ,
represents a "small risk impact" change with respect to the ICLERP. !
l EFAS Evaluation i
The core damage and large early release risk impact of inadvertent EFAS i actuation leading to feeding both SGs, the faulted SG, or no SGs in a Feedwater Line Break (FWLB) or Main Steam Line Break (MSLB) were considered and determined
! to be negligible based on engineering judgement. Thermal-hydraulic best
! estimate analyses performed for the Individual Plant Examination (IPE) indicate that feeding a faulted SG in a FWLB or MSLB event will not result in core damage due to overcooling. A failure to automatically initiate AFW requires the additional failure of 3 or 4 SG differential pressure channels to provide an actuation signal which reduces the likelihood of the event even further. For the loss of Main Feedwater event, a failure of one channel of Steam Generator Pressure - Low could cause a delay of automatic EFAS actuation. In the event
that. AFW automatic actuation fails, at least 30 minutes from initiation is
.available for operator action to manually control AFW flow. The likelihood of the operators failing to manually initiate AFW to any SG within 30 minutes is assumed to be negligible for the purposes of limiting the time a channel can be placed in. trip.
~ 0uality of PRA A comprehensive independent peer review of the SONGS 2/3 Level 1 and Level 2 internal events living-PRA for full power and shutdown operations was conducted-between August 1996 and April 1997 by Dr. Parviz Moieni of SCIENTECH, Inc. The review was mainly based on the guidance provided in the PRA procedure guides suchasNUREG/CR-2300,"PRAProceduresGuide,"andNUREG/CR-4550,"Analysisof Core Damage Frequency." The scope of the peer review is outlined in detail below.
System Fault Trees
'The following system fault trees with their assumptions and associated basic event (BE) calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:
Low Pressure Safety Injection (LPSI)
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Containment Spray / Containment Emergency Cooling (CS/ CEC) l .
High Pressure Safety Injection (HPSI)
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Heating, Ventilation, and Air Conditioning (HVAC)
Component Cooling Water (CCW)
Chemical and Volume Control System (CVCS)
Main Feedwater (MFW) and Condensate j e Main Steam System (MSS) '
Electric Power (EP)
Instrument Air (IA)
- Plant Protection System ( PPS)
Safety Injection Tank System (SIT)
- Reactor Coolant System -(RCS) Pressure Control Containment Isolation System (CIS)
. Saltwater Cooling System (SWC)
The above list represents all the system fault trees in the SONGS 2/3 Living PRA.
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l Event Trees-The following event trees with their assumptions and associated BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:
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LossofPowerConversionSystem(PCS). '
Transients with PCS Initially Available (TT) !
Loss of Offsite Power (LOP)
Station Blackout (SB0)
Main Steam Line Break (SLB)
Small LOCA (SL) l Small Small LOCA (SSL)
Steam Generator Tube Rupture (SGR)
Loss of 125V DC Bus (LDC)
Loss of Component Cooling Water (CCW)
Interfacing System LOCA (VL)
- Reactor Pressure Vessel Rupture (VR)
Anticipated Transient Without Scram (TWS)
- i. . Internal Flooding Analysis T'he above list represents all the event trees in the Level-1 SONGS 2/3 Living i PRA.
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Basic Event (BE) Calculation Files I.
As part of the SONGS 2/3 Individual Plant Examination (IPE) study, a large number of basic event (BE) calculation files had been developed to support a variety of tasks such as human reliability analysis (HRA) and common cause failure (CCF) analysis. All the BE calculation files related to the following topics were reviewed and comments were entered in the SON 3S PRA Review Punch l- List.
. Fault tree analysis
! . Event tree analysis
- CCF analysis j .
Pre-initiating event operator actions (Type A) HRA Post-initiating event operator actions (Type C) HRA
.- Plant-specificequipmentdataanalysis'(i.e.,Bayesianupdateof
- equipment ~ failure rates)
. Plant-specific maintenance unavailability calculations Over 200 BE calculation files were reviewed.
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Level-2 Extended Event Trees (EETs)
The following Level-2 extended event trees with their assumptions were reviewed and comments were entered in the SONGS PRA Review Punch List:
Loss of Power Conversion System (PIE and P2E)
Transients with PCS Initially Available (TIE and T2E)
Loss of Offsite Power (LPE)
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Station Blackout (SBE)
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Main Steam Line Break (MSE)
= Large LOCA (LLE)
Medium LOCA (MLE)
Small LOCA (SLE)
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Steam Generator Tube Rupture (GIE, G2E, and G3E)
Loss of Component Cooling Water (CWE)
Interfacing System LOCA (VLE)
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Reactor Pressure Vessel Rupture (VRE)
Anticipated Transient Without Scram (AWE)
The above list represents all the extended event trees in the Level-2 SONGS 2/3 Living PRA.
ChangestoSONGS2/3PRAModel In addition to the comprehensive independent review described above, San Onofre has established a proceduralized process to provide an assurance of quality for required modifications to the SONGS 2/3 Living PRA through the use of independent reviews and approvals. The following steps are utilized in this PRA change process:
Step 1: PRA Review Punchlist Database. Upon identifying a required change or concern, the PRA engineer describes and records the change or concern that needs to be addressed in this database for tracking purposes.
Based upon the importance of the issue, the item is assigned a priority and a PRA Engineer to address the issue.
i Step 2: PRA Change Package (PRACP) Process. The PRA engineer's l resolution is generated utilizing an internal PRACP procedure. Per procedure, the proposed change is reviewed and approved by an independent reviewer (at a minimum). The change is then implemented in the actual
f model (whether it be a basic event value change or a model change). After the implemented change is reviewed for accuracy, the change is reflected in the PRA Relational Database (see below). Finally, the PRACP process is completed once the package is approved by the supervisor or his designee.
Step 3: PRA Relatiocal Database. The purpose of this database is to document electronically all the assumptions and basic event information used in the PRA model. Once a change has been identified, resolved, and approved, the change is reflected in either the assumptions, basic events, and/or references (Basic event calculations, HRA calculations, systems analysis, etc.) sections of the database.
NO SIGNIFICANT HAZARDS CONSIDERATION:
The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the. facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility l of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. A discussion of these standards as they relate to this amendment request follows:
(1) Will operation of the facility in accordance with this proposed change
!_ involve a significant increase in the probability or consequences of an j accident previously evaluated?
Response: No.
This change provides limits for operating with a channel of the Refueling Water Storage Tank (dWST) Level-Low input to the Recirculation Actuation Signal (RAS) or the Steam Generator (SG)
Pressure-Low or SG Pressure Difference (SGPD)-High input to the Emergency Feedwater Actuation Signal (EFAS) in trip. As a result of this change, the potential for an inadvertent actuation of
- either of these two signals is reduced. The proposed Completion Times are based on Probabilistic Risk Assessment (PRA) considerations, and are conservative compared to the current
- - unlimited Completion Times.
The consequences of an inadvertent actuation of EFAS or RAS are unaffected by this change.
Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) Will operation of the facility in accordance with this proposed I change create the possibility of a new or different kind of accident from any accident previously evaluated? l Resnonse: No. 1 This proposed change provides additional time limits on operation with a channel of the RWST Level-Low input to RAS or the SG Pressure-Low or SGPD-High inputs to EFAS in trip. Operation in this condition is currently allowed indefinitely. The proposed restrictions reduce the possibility of an inadvertent actuation of '
RAS or EFAS, and do not allow operation in any configuration not ;
currently allowed by the Technical Specifications (TSs). !
, Therefore, this proposed change will not create the possibility of a new or different kind of accident from any accident that has been previously evaluated.
(3) Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Resnonse: No.
This proposed change provides additional time limits on operation l ,
with a channel of the RWST Level-Low input to RAS or the SG l Pressure-Low or SGPD-High inputs to RAS or EFAS in trip. The l proposed limits are conservative compared to the current i requirements, where the time limit is unrestricted. The overall impact of the change will be a increase in the margin of safety.
Therefore, there will be no significant reduction in a margin of
( safety as a result of this change.
Based on the responses to these three criteria, Southern California Edison (SCE) has concluded that the proposed amendment involves no significant hazards consideration.
ENVIRONMENTAL CONSIDERATION:
SCE has determined that the proposed amendment involves no changes in the amount or type of effluent that may be released offsite, and results in no increase in. individual or cumulative occupational ;
radiation exposure. As described above, the proposed TS amendment i involves no significant. hazards consideration and, as such, meets the eligibility criteria for categorical exclusion set forth in l 10CFR51.22(c)(9). !
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