ML20199A500
| ML20199A500 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/20/1997 |
| From: | Scott Flanders NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9801270232 | |
| Download: ML20199A500 (29) | |
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f 30 2. l F, - - en as:g j f 4 UNITED STATES l l g , NUCLEAR REGULATORY COMMISSION waswowetow, o.c. seemem j i December 20,1997 i
- LICENSEE
- Florida Power Corporation l
l PLANT: Crystal River Unit 3 l
SUBJECT:
SUMMARY
OF MEETING WITH THE FLORIDA POWER CORPORATION l On Dooomber 15,1997, the Nuclear Regulatory Commission (NRC) staff met with - representatives of the Florida Power Corporation (FPC) in Rockville, Maryland, to discuss ; Technical Specification Change Request Notice (TSCRN) 210,
- Small Break Loss of Coolant ;
Accident,' for Crystal River Unit 3. Enclosure 1 is a list of the meeting attendees and Enclosure , 4 2 is a copy of the material distributed at the meeting. FPC presented on overview of TSCRN ! 210 which included a discussion of the limiting single failure soonarios, operator actions, and a risk analysis. ; Limiting Singia Failures i TSCRN 210 is focused on certain small break loss of coolant accidents (SBLOCAs) that require ! emergency foodwater to maintain Once Through Steam Generator (OTSG) oooling until the t reactor core decay heat can be removed solely by the HiCh Pressure injection (HPI) System )' cooling. For these types of SBLOCAs, concurrent with a loss of offsite power (LOOP), FPC identitled three limiting single failures, (1) loss of Battery A (LOBA),- (2) loss of Battery B ; (LOBB), and (3) loss of the steam driven emergency feedwater pump (EFP-2). ,
- A LOBA would result m a loss of the A train emergency diesel generator (EDG) and emergency I core coohng system (ECCS). One HPl pump via the B train EDG and EFP 2 will be available.
OTSG cooling must be maintained for 36 hours, until decay heat removal can be accomplished by HPl/breek/ safety relief valve (SRV) alone. A LOBB would result in a loss of the B train EDG and ECCS. One HPl pump and the motor F driven emergency foodwater (EFP-1) would be available. EFP 2 would be available as a result , of an 'A' train emergency foodwater initiation and control system (EFIC) signal. EFP-1 use is limited due to the EDG loading capability. The EFP 1 must be secured prior to loading the ! control complex chillers and chilled water pump or the low pressure injection pump (for piggy ) ; back operation or due to LPI actuation at 500 peig). l- l 4
- A loss of EFP 2 would result in both EDGs being available and EFP 1 being available, however, [ !
. due to the EDG cepecity, the EFP.1 (power supplied by the A train EDG) would need to be i escured prior to starting the LPI pump. g Operates Actions ;
- FPC stated, during its presentation, that a total of 17 operator actions 'are included in TSCRN l 210,15 are required to be executed for any one of the 3 scenarios. FPC also stated that only 6 !
of the 17 operator actions are new as result of TSCRN 210, the other 11 were already in there emergency operating procoounse (EO?s). However, only three of the 11 are included in the !
- CR3 lioeneine basis.
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, 2 ; i L Followin0 the presentation, the staff informed FPC that their December 3 and 11,1997, [ responses to the EOP Validation and Venfication (V&V) portion of the stafra November ?6, j l 1997, request for additional information (RAI) did not fully satisfy the staffs request. As a l
- result, the staff requested that: l j (1) FPC expand the Simulator Validation table (Attachment E, Page 1) submitted in !
3 their September 25,1997, supplement to TSCRN 210, to provide completion times : for all 17 of the operator actions, in addition, for operator actions that exceed 20 ! minutes, include the maximum times allowed operators to perform each action !
- without exceeding any safety limits. ;
i i (2)- FPC discuss how the avera9e times identified in Attachment B (Table of EOP Step Changes) of their Oooember 11,1997, RAI response were derived, and (3) FPC provide information that demonstrates that all crews ca accomplish each of i the operator actions required, for all three scenarios, in the maximum time allowed l operators to perform each ac; ion without exceeding safety limits. 3 l This information should include a discussion of the operator training and evaluations completed l
- for each of the three scenarios, and should make clear the distinction between EOP V&V end i operator training and evaluation (i.e., the purpose of each and how each was conducted, i emphasizing how licensed operators were involved. [e g., whether they were selected on an as- i available basis or participated as a functioning crew; whether there was instructor assistance provided durin0 the simulator exercises or not, etc.) ). Subsequent to the meeting, FPC
, laformed the staff that all crews were trained for each of the three scenarios, but none of the crews were evaluated (i.e., simulator tested) for a I.OBB or failure of EFP 2. Crows were > evaluated for a loss of EDG A, which FPC states is very similar to a LOBA. FPC stated that it ! would verify this information and provide it in writing to the staff. . t 4 Rink Analysis l During FPC's discussion on risk analysis, it was stated that the initiating event frequency ;
. assumed for SBLOCA-induced LOOP (i.e., SBLOCA frequency multiplied by a conditional probability of LOOP)is 2.24 E 5/ year. FPC stated that it performed a simplified bounding analysis that assumed a human error probability of 1.0 for each of the new operator actions. i The results of that analysis showed that the core damage frequency (CDF) (for all intomal i
, events) without accounting for new operator actions was 7.19 E 6/ year (baseline CDF) and 7.51 E-6/ year accounting for the new operator actions (with an a* vami human error probability of :
1 0). - , The staff questioned how FPC derived a CDF that is lower (by an order of a magnitude or two) . ! than the initiating frequency of SBLOCA/ LOOP when all the new operator actions were f assumed have failed (i.e., human error probability of 1.0). In response, FPC stated that taking j credit for a non-safety related feedwater pump (FWP 7) allowed them an attomate success : .- path to mitigate the analyzed accident scenarios, which resulted in a CDF estimate lower than ! the initiating event frequency. FWP 7 is powered by a newly installed diesel generator {
. _ = _ _ . _ .; _ m _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ m._
__ . . _ . _ . _ . _ . _ _ . _ . . _ _ _ _ ~ . _ . . _ _ . _ _ _ . _ . _ _ _ . _ _ _ _ . - . I 3 dedicated to the pump and both the generator and FWP 7 can be started from the control room. The staff stated that in TSCRN 210, FWP 7 was only discussed as defense in depth. The staff 1 l !. asked if FWP 7 is included in the scope of the maintenance rule. FPC stated that FWP 7 is not included in the maintenance rule scope but it believed that it would be included in the future. i 1 As a result of the information provided in FPC risk analysis discussion, the staff requested that FPC provide: j 1 (1) A detailed discussion on how FWP 7 affe f.s the results of FPC's risk analysis
- (providing quantitative as well as qualit/Jve description), including the assumed ;
unavailability / failure probability of FWP 7 l
'(2) A discussion on how the estimated risk at the plant would change if FWP 7 is not ;
credited in the risk analysis, in other words, what would be the estimated overall ; CDF if FWP 7 was not crooi.ed for SBLOCA/ LOOP sequences? (3) - Justification and references for the single failure probabilities presented in the risk analysis table provided in the meeting, and to expand the table to include single ; failures that are less limiting but more likely than the LOBB, LOBA, and EFP-2, ! (i.e., failure of EDG), and !
-(4) Justification for the frequency assumed in FPC December 3, igg 7, RAI response r for a SBLOCA occurring in the HPl (7E.7/ year).
General Dealgn Criterion (GDC) 20 > I 4 The discussion of operator actions led the staff to request that FPC address Crystal River i design criterion 14 " Core Protection Systems' (similar to GDC 20,
- Protection System -l Functions'), which is described in the Final Safety Analysis Report (FSAR) and requires the i high pressure injection to be initiated automatically. The analysis used to support Amendment 210, for some of the small-break loss-of-coolant accidents, credits manual initiation of HPl by ;
- the operators, rather than automatic initiation. The licensee indicated that they would correct their licensing documentation (FSAR and Technical Specifications Bases) to clarify the credit i
,- given to manualli)ltiation of HPl. ' Scott . landant, r ro]ect@ Project Directorate PD ll-3 ; Division of Reactor Projects - t/ll Office of Nuclear Reactor Regulation Docket No. 50-302: ;
Enclosures:
1.- Attendees List 4 2L Meeting Handout
.oc w/oncWures: See next page j
a a .. - . . . - . - . _ - _ _ _ _ _ _ . - - - - - - . . - . - . = . . .
3
- dedicated t) the pump and both the generator and FWP.7 can be started from the control room.
l The staff stated that in TSCRN 210, FWP-7 was only discussed as defense in depth. The staff asked if FWP-7 is included in the scope of the maintenance rule. FPC stated that FWP 7 is not included in the maintenance rule scope but it believed that it would be included in the future. 4 As a result of the information provided in FPC risk analysis discussion, the staff requested that FPC provide: (1) A detailed discussion on how FWP 7 affects the results of FPC's risk analysis (providing quantitative as well as qualitative description), including the assumed unavailability / failure probability of FWP-7 (2) A discussion on he v the estimated risk at the plant would change if FWP 7 is not credited in the risk ant!Psis, in other words, what would be the estimated overall CDF if FWP 7 wa' at credited for SBLOCA/ LOOP sequences? (3) Justification and references for the single failure probabilities presented in the risk analysis table provided in the meeting, and to expand the table to include single failures that are less limiting but more likely than the LOBB, LOBA, and EFP 2, (i.e., failure of EDG), and (4) Justification for the frequency assumed in FPC December 3,1997, RAI response for a SDLOCA occurring in tne HPl (7E 7/ year). General Deslan Criterion (GDC120 The discussion of operator actions led the staff to request that FPC address Crystal River design criterion 14 " Core Protection Systems'(similar to GDC 20, " Protection System Functions *), which is described in the Final Safety Analysis Report (FSAR) and requires the high pressure injection to be initiated automatically. The analysis used to support Amendment 210, for some of the small break loss-of-coolant accidents, credits manualinitiation of HPl by the operators, rather than automatic initiation. The licensee indicated that they would correct their licensing documentation (FSAR and Technical Specifications Bases) to clarify the credit given to manualinitiation of HPl. ORIGINAL SIGNED IW: Scott C. Flanders, Project Manager Project Directorate PD ll-3 Division of Reactor Projects - l/II Office of Nuclear Reactor Regulation Docket No. 50-302
Enclosures:
- 1. Attendees List
- 2. Meeting Handout cc wienclosures: See next page Document Name: G ACRYSTA01ll15. SUM Office PM PDil-3 LA PDil 3 PD PDil-3 i Name sFlande DClayton # FHebdon dI note ww /2Nfl'19 cmr ahehn .nu1er it -h7 Copy Yet/No Yes (YisJNo OF FICIAL RECORD COPY
i Florida Power Corporation CRYSTAL RIVER UNIT NO. 3 GENERATING PLANY y cc: Mr. R. Alexander Glenn Mr. Robert E. Grazio, Director Corporate Counsel Nuclear Regulatory Affairs (SA2A) Florida Pcwer Corporation Florida Power Corporation MAC ASA Crystal River Energy Complex P.O. Box 14042 15760 W. Power Line Street St. Peteraburg, Florida 33733-(042 Crystal River, Florida 34428-6708 Mr. Charles Pardee, Director Senior Resident inspector Nuclear Plant Opera' ions (NA2C) Crystal River Unit 3 Florida Power Corporation U.S. Nuclear Regulatory Commission Crystal River Energy Complex 6745 N. Tallahassee Road 15760 W. Power Line Street Crystal River, Florida 34428 Crystal River, Florida 34428-6708 Mr. James S. Baumstark Mr. Bruce J. Hickle, Director Director, Quality Programs (SA2C) Director, Restart (NA2C) Florida Power Corporation Florida Power Corporation Crystal River Energy Complex Crystal River Energy Complex 15760 W. Power Line Street 15760 W. Power Line Street Crystal River, Florida 34428-6708 Crystal River, Florida 34428-6708 Regional Administrator, Region Il Mr. Robert B. Borsum U.S. Nuclear Regulatory Curamission Framatome Technologies Inc. 61 Forsyth Street, SW., Suite 23T85 1700 Rockville Pike, Suite 525 Atlanta, GA 30303-3415 Rockville, Maryland 20852 Mr. John P. Cowan Mr. Bill Passetti Vice President - Nuclear Production Office of Radiation Control (NA2E) Departroent of Health and Florida Power Corporation Rehabilitative Services Crystal River Energy Complex 1317 Winewood Blvd. 15760 W. Power Line Street Tallahassee, Florida 'J2399-0700 Crystal River, Florida 34428-6708 Attomey General Mr. Roy A. Anderson Depariment of Legal Affairs Senior Vice President The Capitol - Nuclear Operations Tallahassee, Florida 32304 Florida Power Corporation ATTN: Manager, Nuclear Licensing Mr. Joe Myers, Director Crystal River Energy Complex (SA2A) Division of Emergency Preparedness 15760 W. Power Line Street Department of Community Affairs Crystal River, Florida 34428-6708 2740 Center /lew Drive Tallahassee, Florida 32399-2100 Mr. Keny Landis U.S. Nuclear Regulatory Commission Chr.rman . 61 Forsyth Street, SW., Suite 23T85 Board of County Commissioners Adanta, GA 30303-3415 Citrus Court /. 110 North Apopka Avenue Ivemess, Florida 34450-424S
a Distribution HARD COPY < en, Docket Filo-PUBLIC Crystal River Reading OGC ACRS L.Raghavan S. Flanders J Johnson, Ril E:Mau S. Collins /F. Miraglia (SJC1.FJM) ; R. Zimmerman (RPZ) B, Sheron B. Boger (BAB2,RCN) G. Holahan
- J. Zwolinski (JAZ)
S. Newberry (SFN) K. Landis (KDL) F. Hebdon (FJH) B. Clayton (BAC2) J. Jaudon, Ril (JPJ) T. Martin (e-mail to SLM3) L. Trocine, EDO
- A. El-Bassioni J. F'ack J. Bongarra Samuel Lee -
W. LaFave C. Jackson F.Orr R. Landry W. Lyon G. Galleti
-- -, - . - . . .-- -- .- ~-
MEETING ATTENDEES December 15.1997 MEETING WI,TH FLORIDA POWER CORPORATIQN NAME- ORGANIZATION Scott Flanders NRR/PD ll-3 L. Raghavan NRR/PD ll 3 Adel El-Bassioni NRR/DSSA/SPSB John H. Flack NRR/DSSA/SPSB Jim Bongarra NRR/DRCH/HHFB Samuel Lee NRR/DSSA/SPSB William T. LeFave NRR/DSSA/SPLB Christopher Jackson NRR/DSSNSRXB Fank Orr NRR/DSSNSRXB Ralph Landry NRR/DSSA/SRXB Federick J. Hebdon NRR/PD ll-3 Mike Lalor FPC/ Licensing Paul Fleming FPC/ Engineering Mike Rencheck FPC/ Director, NEP Tim Catchpole FPC/ Licensing Mark Averett FPC/ Engineering, PSA Robert Grazio FPC/ Director, Nuclear Regulatory Affairs Warren Lyon NRR/DSSNSRXB Roy Zimmerman NRR/ Associate Director, Reactor Projects Greg Galleti NRR/DRCH/HHFB - Encolsure 1
- . - . - - - - - -------------,-,------,w---,----------- -
A g 0 Enclosure 2
TSCRN 210 I l NUCLEAR REGULATORY COMMISSION 4 i FLORIDA POWER CORPORATION DECE MBER 15,1997 i 1 ~
~
24(p AGENDA -
- Introduction - Mike Rencheck
- SBLOCA Solution Sets - Paul Fleming
- Operator Actions - Tim Catchpole
- Risk Analysis - Mark Averett
- Conclusion - Mike Rencheck
)
~
l - LIMITING SINGLE FAILURES THE LIMITING TRANSIENTS AFFECTING THE EMERGENCY DIESEL GENERATOR LOADING AND ' EMERGENCY FEEDWATER WERE IDENTIFIED AS: SBLOCA & LOOP AND - LIMITING SINGLE FAILURES
28# LIMITING SINGLE FAILURES ~
- LOBA or LOBN - Loss of a Battery j i
> Disables ONE Emergency Diesel Generator
> Disables ONE Train of Emergency Core Cooling
> Disables ONE Train of DC Control Power !
I EFP-2 J l The Motor Driven Emergency Feedwater Pump (EFP-l) is designed to be load shed (during the accident) to protect the 'A' Emergency Diesel Generator from an overload condition
l
%& Startup Team Solution Set
- ~
<b
~ .
LIMITING SINGLE FAILURES
- LOBA Loss of "A" Battery
!NITIAL CONDITIONS r , r , r , r '
- MODE
+ SBLOCA + & +
l LOOP F 1 UR S LON :
; < ; < > ' Loss of "B" Batterv g
E
- EFP-2 :
- S
' d
G J 24 e 4 LOSS OF BATTERY "A" Wa i i i 4- i PU
- MUP-1B or MUP-1C S
i
, , U >
MITIGATION i STRATEGY
^* 1 HPI PUMP
' ' C i
> EFP-2 ---+ FRAMATOME
& EFW 5 ADMIN -> AND +
n EFW FLOW ' 2 ANALYSIS e ; C i I i EFW ! FLOW i { , 1 l l i i I S FWP-7 DEFENSE IN DEPTH 8---> FWP-7 --> DIESEL ----- 3
' 2 MOD S i
I SUCCESS PATH " l DEFENSE IN DEPTH ' i
9 24 Y Y$, t LOSS OF BATTERY "B" l PUM :( MUP-1A or MUP-18 ) - sL INSUFFICIENT l OPERATIONAL
-+ EFP-1 EDG -* ~~~~~'
UMITATION I CAPACITY
.- ............. I S
-CONTROt. COMPT _EX Ih COOUNG (-1.0 hr) , [ o
- P MITIGATION g STRATEGY W E DEPENDE y _ ,
-BWST DEMEDOM
, , U l
LOBB->
-500 PSIG . M 1 HPl PUMP .
(Operator Coosdown) . 1$
..............- i ;
C
,, I o
- g u
- i C
' ppw ' m FRAMATOME ' r ,
- + EFP-2 -* CROSSTIE EFW TRAINS -* EFW y ADMIN + AND +
FLOW E
' " " ANALYSIS >
l l I l s I I I s
! r , FWP-7 DEFENSE IN DEPTH
~
SUCCESS PATH O DEFENSE IN DEPTH- - - - h
Xn M
$"3Fb v".)
LOSS OF EFP-2
~
I 7 HPI . HPI FLOW PJMPS ANALYSIS " S U MITIGATION STRATEGY FRAMATOME TIME EFP-2-+ 2 HPI PUMPS + ANALYSIS b l & EFW FLOW '
'OPERATIONA' -
LOAD EFP-1 TRIP g C
-> EFP-1 + L + + DEFEAT 7 + AND +
MANAGEMF4T N
, LIMITATION SWITCH w
' ,I I
EFW ~ mod E FLOW _ swsy DEPLETION
-{"MeMOD (PTLS l
l
# _w I 3
l (operator cooldown) RW l l l s DEFENSE IN DEPTH 7 D L SUCCESS PATH MOD , DEFENSE IN DEPTH ----- v (_
~
OPERATOR ACTION .l
~
y;a_
# OVERVIEW ,
'W l
Less than 20 mmutes_
- No new operator actions prior to 20 minutes t
i l I 4 9
~
~
OPERATOR ACTION 1 L OVERVIEW l Greater than 20 minutes
- 5 new operator actior s for load management
> Contingent upon single failures
> Conditions occur no earlier than 1 hour
> At most 3 of the 5 required for any of the three limiting single failures
- 1 new contingency for HPI pinch break
> Contingency for RCS re-pressurization
G y:a
-. _.. - g OPERATOR ACTION -
L 9 . OVERVIEW OA Operator Action 11& W l Itequired L.lL-N fl Previously Specilie to Potentially Generic fier ISCM Included in 210 Single Itesolveti Next Guidance 53mptom liOl% Failures Outage 10 Minute Actions ' 1 Trip RCI's V 2 Initiate llPl/ltillC 4 V V V V
~ T i 3 Open all 4 IIPI Injection Valves V 20 Minute Actions V V V y I 4 isolate RCP Seal Injection V V V V 5 Isolate llroke., llPl line V V V '
V 6 Itaisc 01 SG level to 90%
~
V V Greater Tlian 20 Minute Actions V y 7 linsure/ Start CC Vent in timergency V V X Transier liCCS Pump Suction to Sump V
*("199);
V V 9 Crosstic liFP-2 to "A 't rain (lil V-12) V V V V i 10 Close lil:W lilock Valves 4 1i Place ItWP/SWP in Pull-to-l.ock and V V
!!FP-1 in Trip Dellat 12 linsure/ Start CC Chiller V V 13 Isolate Itll Sump V V
~
14 Cycle lil:P-2 on I,ow OTSG Pressure V 15 Place lil:P-l in Trip Del' eat Prior to V V l.Pl/liFP-l Interhick (500 psig) 16 ItCS Cool <fown using 'lIIVs or ADVs V V V
!7 Periottically livaluate llPI line lireak V V criteria on Itepressuriration i 4 kJ3F E C L$ . ,
l 3
#.qp4 RISK ANALYSIS
, , a e .
- Bounding simplified risk analysis shows minimal impact on Core Damage Frequency l
- Conservative Assumptions:
> Conditional LOOP given a SBLOCA
> Initiating event frequency greater than NUREG/CR-6538
> Operator fails to perform tLle new actions 100% of the time
> Loss of EFW given loss of"A" or "B" DC power
. - . .- .. .... .. .. -.. . --. . . . . .. )
l ga _m INITIATING EVENT . ~
.:#.. FREQUENCY SBLOCA Frequency 1.6 E-3 9er year (CR-3 PSA) (EPRI and ANS)
Conditional Probability of LOOP 1.4 E-2 - given SBLOCA j (NUREG/CR-6538) SBLOCA/ LOOP frequency for 2.24 E-5 per year TSCRN 210 SBLOCA/ LOOP frequency in 1.4 E-3 per year NUREG/CR-6538
24qp RISK ANALYSIS
~
- Risk Modeling:
> 5 Operator actions for load management I
> Event frequency for HPI pinca break is low
- CORE DAMAGE FREQL EN CY(aiiinternaievents):
> CDF without new operator actions 7.19 E-6/ year
> CDF with new operator actions 7.51 E-6/ year
RISK ANALYSIS l . i l
- Simplified risk analysis is bounding
- Analysis does not consider:
L
> Reasonable Human Error Probabilities i
> Sequence frequencies due to required failures l > Reduction for Human Error Double Counting I
i _ - e mm_, . . , w - -.- ,
l .. RISK ANALYSIS I
- Reasonable Human Error Probabilities are justified:
> Trained Operators
> Explicit Procedures
> Adequate operator response time L . .-
..i 20 9 RISK ANALYSIS ~!
i .
- Sequence frequency for each operator action: j
- > Frequencies range from 4.25 E-8 to 8.11 E-9
> Frequencies have no appreciable effect on CDF l
4 i l
RISK ANALYSIS
- Human Error double counting: .
> Reduces CDF for TSCRN 210 to 7.42 E-6
( 1,
RISK ANALYSIS
- Bounding simplified risk analysis shows minimal impact on Core Damage Frequency
- Reasonable assumptions for risk analysis shows new operator actions do not have an appreciable effect on Core Damage Frequency l
Conclusion l
- At most 3 of the 5 new operator actions required for any of the three limiting single l failures
- No appreciable impact on core damage frequency
- Reasonable engineering approach on mtenm basis
. _ _ _ - - _ _ _ _ _ __--__ __--__ ______ _ __ _ __.i
i 2:;_: s
- S8LOCA/ LOOP- Human Error Frequency (per.
-OA Frequency (per year) Single failure Probotnhty Other Fadures Modeled Probainhty ProbotnHey year) Comments 9 2.24E-05 LOB 8 3.62E-04 None- N/A 1.0 8.11E-09
-10 2.24E-05 LOBB' 3.62E-04 None N/A 1.0 8.11E-09
- HPI (given EGDG-1 A and
'11 2.24E-05 EFP-2 5.70E 5.09E-02 1.0 6.50E-06 SU sequence EFW failed)
HPR (given EGDG-1A and 11_ 2.24E-05 EFP 5.70E-02 EFW failed) 1.93E-03 1.0 2.46E-09 SX sequence FWP-7/PORV (given 11 2.24E-05 ' EFP-2 5.70E-02 EGDG-1 A and EFW failed) 3.33E-02 1.0 4.25E-06 SBP sequence 14 2.24E-05 LOBA 3.62E-04 None N/A 1.0 8.11 E-09 e 14- 2.24E-05 LOBB 3.62E-04 None N/A 1.0 8.11 E-09 HPI (given EGDG-1A and 15 2.24E-05 EFP-2 5.70E 5.09E-02 1.0 6.50E-08 SU sequence EFW failed) HPR (given EGDG-1A and
'15 2.24E-05 EFP-2 5.70E-02 1.93E-03 1.0 2.46E-09 EFW failed) SX sequence FWP-7/PORV (given 15 2.24E-05 FFP-2 5.70E-02 3.33E-02 1.0 - 4.25E EGDG-1A and EFW failed) SBP M mx:e
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