L14a: Practical examples of SAMG (Krsko NPP) Presentation on IAEA TRAINING WORKSHOP ON Severe Accident Management Guideline Development using the IAEA SAMG-D toolkit, Vienna 19-23 October 2015 Presented by Ivica Bašić APoSS d.o.o.
L14a: Practical examples of SAMG (Krsko NPP)
Presentation on
IAEA TRAINING WORKSHOP ON
Severe Accident Management Guideline
Development using the IAEA SAMG-D toolkit,
Vienna 19-23 October 2015
Presented by
Ivica Bašić
APoSS d.o.o.
2
Overview
• Experience with Krsko AMP
• WOG Generic SAMG Implementation
• Relationship between NEK IPE and SAMG
• Phenomenology of Accidents addressed by Krsko IPE
• Background Documents – Strategies/Setpoints
• ENSREG stress test
• Experience with Krsko AMP
• Conclusions
• References
3
Introduction of Krsko AMP
• Plant specific analysis (IPE – Individual Plant
Examination & IPEEE for external events) - plant
response on Severe accident
– Hazard Analysis (voulnerabilties identification)
– PSA Level 1
– PSA Level 2 including Containment fragility
analysis, deterministic analysis (MAAP),..
• NPP Krško specific full scope simulator with capability
to simulate severe accidents (based on MAAP models)
• Plant specific SAMG based on WOG generic
procedures
4
– Definition of transition (rules of usage)
– SAMG for MCR (should be similar to FR-C1)
– SAMG for Spent Fuel Pool (not available in generic
SAMG, important issue from Fukushima point of view)
– SAMG for shutdown (e.g. loss of SRH on midloop
operation)
– Alternative means (mobile equipment) usage:
• Different fire protection pumps
• Fast connections to the systems (e.g. injection into SGs)
• Source of waters (e.g. amount for flooding the containment to
protect cavity floor from MCCI OR even flooding the Rx cavity
to the top of acctive fuel to establish external cooling)
Additional Plant Specific Issues
5
Relationship between NEK IPE and SAMG
Level 1 PSA
Sequences that lead to
core damage after 24
hours
Dominant core damage sequences from Level 1
study have been grouped and assessed following
the criteria set out in NUMARC 91-04, Severe
Accident Issue Closure Guideline
For beyond 24 hour sequence (loss of SW, loss of
CCW, station blackout), insights were developed
based on the accident scenarios
The Level 2 results have been grouped into
release categories and insights have been derived
based on these categories.
Also, the phenomenological evaluations have
been reviewed to gather additional insights.
Level 2 PSA
Plant-specific Severe Accident Management insights were
developed based on the following:
NEK IPE – Individual Plant Examination & External Events
Evaluation
Hazard Analysis/Vulnerabilities Identification
6
• NEK Internal Review (technical aspects)
• External Review, ERIN (methodological aspects)
• International Reviews
• IPERS Review of Level 1 Assessment (June-July 1994)
• IAEA Engineering Review of External Events PSA (Feb
1996)
• IPERS Review of Level 2 Assessment (Nov 1997)
• IPERS Review of Fire PSA Assessment (June 1998)
• IPSART Review (2000)
• Other Reviews (JŠI - Level 2, Enconet Consuting -
Shutdown Modes PSA)
NEK IPE / IPEEE Reviews
7
• Behavior up to core uncovery
• Cladding oxidation; transport, release and combustion of hydrogen
• Core uncovery, heatup, melt, relocation
• Core melt progression
• Hydrogen generation
• Natural circulation and creep failure phenomena
• Reactor vessel wall attack/melt-through
• Reactor vessel failure
• Effect of Operator actions on Accident progression
• High pressure vessel failures; code debris and coolant ejection
• Core debris dispersal - Direct Containment Heating (DCH)
• Core debris/water and debris/concrete interaction
• Cladding oxidation; transport, release and combustion of hydrogen
• Fission product behavior
• Containment failure
Phenomenology of Accidents addressed by Krsko IPE
IN VESSEL
IN CONTAINMENT
RELEASES
8
Krsko NPP MAAP 4.0.5 NSSS Model
4
56
RCP 87
1
6
17
2
1
20
19
18
11
1312
15 14
222
4
23
RCP
9
2
3
1
10
9
2
2726
28 29
ACCACC
SI SI
RHRRHR
9
9
Analyses of 3 LOAF cases:
LPI recover just before HLs creep failure
(CREEP1)
HLs creep failures prevented by user intervention
(CREEP2)
recovery of AFW (CREEP3) just before SG U-
tubes failiure in case CREEP2
Deterministic Analysis of Severe Accidents Phenomena – example CREEP failure and influence on SAMG
Surface temperature of SG hot tubes
0
100
200
300
400
500
600
700
800
900
1000
0.00E+00 5.00E+03 1.00E+04 1.50E+04 2.00E+04
Time (s)
Su
rface t
em
pera
ture
(K
)
CREEP1 SG hot tubes
CREEP2 SG hot tubes
CREEP3 SG hot tubes
CREEP2 SG tubes
creep failure
CREEP1 LPIS ON
CREEP3 AFW ON
Deterministic Analysis of Severe Accidents Phenomena – example CREEP failure and influence on SAMG
TCRHOT - core hotest node temperature
0
500
1000
1500
2000
2500
3000
3500
0.00E+00 5.00E+03 1.00E+04 1.50E+04 2.00E+04
Time (s)
Tem
pera
ture
(K
)
creep1
creep2
creep3
creep1&3, HLs creep
failures
11
Conclusions:
• BE SG tube temperatures do not approaches values
affecting the integrity without non-conservative
assumptions (degraded tube thicknesses, lower
material creep temperature, etc.)
• HLs failure may not be ruled out. It was necessary to
develop a probability distribution to quantify this failure
mode for CET quantification.
• Krsko NPP specific results using MAAP 4.0.5 are seen
to be in general agreement with analyses performed for
other W plants
Deterministic Analysis of Severe Accidents Phenomena – example CREEP failure and influence on SAMG
RC
no.
Release Category
Definition
Base case
(plant as-is)
Sensitivity
case 1 (wet
cavity)
Sensitivity
case 2 (filtered
vent)
Sensitivity
case 3 (both)
1
Core recovered in-vessel, no
containment failure
1.82
1.82
1.82
1.82
2
No containment failure
26.71
38.28
26.72
38.28
3A
Late (time frame IV) containment
failure, no molten core-concrete attack
0.36
0.53
0.0
0.0
3B
Late (time frame IV) containment
failure, molten core-concrete attack
23.26
14.62
10.20
5.92
4
Basemat penetration (no overpressure failure)
14.14
7.84
14.14
7.84
5A
Intermediate (time frame III) containment failure, no molten core-
concrete attack
8.79
12.98
0.0
0.0
5B
Intermediate (time frame III) containment failure, molten core-
concrete attack
2.47
1.47
2.47
1.47
6
Early (time frame I or II) containment
failure
0.03
0.03
0.03
0.03
7A
Isolation failure, no molten core-concrete attack
4.04
6.33
4.04
6.33
7B
Isolation failure, molten core-concrete attack
5.78
3.49
5.78
3.49
8A
Bypass, scrubbed
9.98
9.98
9.98
9.98
8B
Bypass, unscrubbed
2.61
2.61
2.61
2.61
VENT
Containment filtered vent release
-
-
22.22
22.22
Level 1 / Level 2 Results Summary
13
Evaluate the plant modifications (and other relevant issues) that took place from the beginning of 1993 (IPE “freeze date”) up to the Outage of 2000 inclusively
Product: updated Baseline PSA model referred to as “NEK98”
Substantial effort involved: more than 1500 issues evaluated
• AMSAC Modification
• Replacement of 125 Class 1E Batteries
• Four hours control gas supply to AFW and MS valves
• FPAP modification related to area CB-3A
• Modification of ESW system
• Seismic upgrade
• SGs Replacement and Power Uprate
• EOP changes
• Parameter update
• etc.
Major PSA Level 2 Update: 1992 – 2000
14
Post Modernization Krsko NPP CDF Profile
CDF by IE Categories
0,00E+00
5,00E-05
1,00E-04
1,50E-04
2,00E-04
2,50E-04
Other External Events 1,26E-05 1,26E-05 1,26E-05
Internal Fire 9,78E-05 9,29E-05 1,25E-05
Seismic Events 6,03E-05 5,68E-05 5,62E-05
Internal Flood 4,62E-06 4,55E-06 4,36E-06
Internal Events 5,44E-05 5,34E-05 3,09E-05
1992 - "NEK16B" "NEK98" "NEKC17"
1992 1998 2000
IPE (RS) ISA Stage 1Post-modern.
update
Total CDF /yr 2,3E-04 2,2E-04 1,2E-04
Large Release
Fraction5% 4,5% 3,1%
15
Background Documents - Strategies
Purposes were:
• Identify if all generic strategies are
applicable to NEK - can successfully be
applied;
Accident Management measures
or strategies may be
PREVENTIVE (delay or prevent
core damage) or MITIGATIVE
(mitigate core damage and protect
fission product boundaries) or
BOTH
• Verify if plant specific IPE insights are
adequately addressed in generic
strategies;
• Identify the plant specific capabilities
(equipment that will be used), action to be
taken to mitigate the challenge
Necessary Severe Accident t-h calculation and evaluation: • Sequence sensitivity
analyses and time evaluation • Available time for mitigate
actions • Success criteria • Plant staff and
equipment availabity
16
Krsko NPP Containment Geometry/Wet Cavity Modification
RWST
RWST + RCS + 2 ACCU
(1) (2) (3) (4) (5) (6) (7) (8) (9) (10) (11) (12)
PDS PDS Sequence Sequence Time Time Time Time Time Time Time Win Time WinFrequency Frequency (sec) (sec) (sec) (sec) (sec) (sec) (sec) (sec)
Reactor Core T > 650 K Onset of Support RV RV External RV Cavity
Scram Uncovery Core Plate Failure Failure Flooding Flooding
Melting (9)-(7) (10)-(7)
TEHAYN 1.282E-05 TRA12_1 3.80E-06 9 4171 4725 6126 6640 8440 1915 3715TEHANN 7.217E-06 TRA12_33 5.40E-08 9 4175 4734 6136 6530 8330 1796 3596ALLBYN 6.450E-06 LLO2_1 1.30E-06 0 2750 3301 4012 6698 14679 3397 11378
No CMT Heat Removal LLO2_1A N/A 0 4014 4635 5444 8909 15358 4274 10723
WUUUUB 4.424E-06 SGR9_1 3.00E-06 306 59661 62206 65039 68059 74381 5853 12175
SEHAYN 4.350E-06 SBO20_1 2.60E-06 0 2566 3148 4235 9636 11436 6488 8288
Increase Debris Mass Ejected SBO20_1A N/A 0 2566 3148 4235 9636 11436 6488 8288
SLLBYN 2.015E-06 SLO3_1 1.50E-06 558 61498 64944 122947 146458 181236 81514 116292
FR-C.1 Depressurization Fails SLO3_1A N/A 558 56027 60059 65320 69737 82423 9678 22364
VXXXXB 2.003E-06 ISL1_1 2.00E-06 9 20378 21327 22360 25621 32983 4294 11656
TEHNNN 1.819E-06 SBO63_37 3.90E-08 0 6004 6641 8217 10147 11947 3506 5306
Reduced Debris/Coolant CHF SBO63_37A N/A 0 6004 6641 8217 10147 11947 3506 5306
Reduced Spreadhout Area SBO63_37B N/A 0 6004 6641 8217 10147 11947 3506 5306
SLNNN 1.250E-06 INA3_1 6.80E-07 0 2638 3206 4307 15361 45670 12155 42464
One Fan Cooler Running INA3_1A N/A 0 2634 3212 4304 14710 39693 11498 36481
ALLBYI 9.330E-07 LLO2_2 1.60E-07 0 2630 3187 3890 6559 15390 3372 12203
TERAYN 4.177E-07 LSP6_1 4.10E-07 0 6200 6939 8594 25533 1.E+10 18594 1.E+10
N/A N/A CREEP6 N/A 9 4174 4732 6133 6467 1.E+10 1735 1.E+10
Minimum Time Window (sec) for RV External Flooding 1735Minimum Time Window (min) for RV External Flooding 29
Minimum Time Window (sec) for RV Cavity Flooding 3596Minimum Time Window (min) for RV Cavity Flooding 60
Note: The 1E+10 value is to represent a large time for the calculation since there is no RV failure.
Background Documents - Strategies
COMPONENT NAME
TAG NUMBER
COMPONENT
CHARASTERISTICS
(Nominal flow, shutoff
head, etc)
SUPPORT SYSTEMS
Instrument air Cooling AC BUS/MCC DC BUS/BRKR
PUMPS
Motor driven auxiliary pump 1A,
1B
AF102PMP-01A
AF102PMP-02B
Rated capacity 84.14 m3/h at
104.9 kp/cm2 (1022.3m); Shutoff
head 129.5 kp/cm2 (1264.9m);
required NPSH 5.8m
just for AF control
valves
CC train A
and B
EE105SWGMD1/3
EE105SWGMD2/3
DC101PNLK101/4
DC101PNLK301/4
Turbine driven auxiliary pump 1C
AF101PMP-03C
Rated capacity 184 m3/h at 106.2
kp/cm2 (1035.7m); Shutoff head
127.8 kp/cm2 (1249m); required
NPSH 6.1m
just for AF control
valves
N/A
N/A
(steam pressure must be
greater than 5 kp/cm for
pump operation)
N/A
Main feedwater pumps (1A, 2B,
3A(B)-powered from M1 or M2
bus)
FW 105 PMP 001
FW 105 PMP 002
FW 105 PMP 003
Rated capacity 2339.6 m3/h at
65.9 kp/cm2 (642.5m); Shutoff
head 78.8 kp/cm2 (768 m);
required NPSH 33.5m
just for MFW control
valves
N/A
EE105SWGM1/6
EE105SWGM2/9
EE105SWGM1/7 or
EE105SWGM2/8
DC101PNLG701/17
DC101PNLG701/2
DC101PNLG710/17
DC101PNLG710/2 Condensate pumps
CY 100 PMP 001
CY 100 PMP 002
CY 100 PMP 003
Rated capacity 1362 m3/h at 28.6
kp/cm2 (279 m); Shutoff head
33.5 kp/cm2 (326m); required
NPSH 1.1m
N/A
N/A
EE105SWGM1/10
EE105SWGM2/5
EE105SWGM2/6
DC101PNLG701/1
DC101PNLG701/18
DC101PNLG701/18
Condensate transfer pump
CY 110 PMP
Rated capacity 37.5 m3/h at 6.7
kp/cm2 (65.5m); shutoff head
8.11kp/cm2; required NPSH
2.13m
N/A
N/A
EE103MCC111/6C
N/A
Demineralized water transfer
pumps(2)
WT114PMP001
WT114PMP002
57 m3/h each at 6.1 kp/cm2
N/A
N/A
EE103MCC111/7A
EE103MCC212/10E
N/A
Background Documents - Strategies
Background Documents – Setpoints, CA examples
Determination of NEK Specific Value:
The available NPSHs from the VCT calculated for charging pumps, exceed the required NPSHs
independently to the water level in the VCT (Document Id:630-7).
Based on discussion for L04, vortexing formation is limiting for the CVCS pumps. Water velocity in
VCT outlet nozzle (4-CS-151R, sch. 40 pipe) due one centrifugal pump running is 1.22908 m/s (based
on 160 gpm flowrates [36.34m3 /h] - USAR Table 9.3-2, outlet nozzle cross section area of 8.213E-3
m2). Based on curve of Hydraulic Standard required relative submerge is 0.632m.
(1) Centerline of pipe = 116.786m (dwg. E-304-680)
(2) Centerline of level tap = 117.355m (dwg. B-814-670, sh. 19)
(3) Radius of outlet pipe = 0.05113m
(4) Relative submerge = 0.632m
(5) (2)-(1)=0.569m
(6) Relative submerge toward level tap = (4)+(3)-(5)=0.114m
(7) 100% of level span is 1.906m
(8) Relative submerge toward level tap in % = (6)/(7)*100= 5.988%
NEK Specific value for L06 = 6.0 % of VCT level
Figure 3-1a
Potential for Hydrogen Combustion Based on Wet
Hydrogen Measurement
0
5
10
15
20
0 0.5 1 1.5 2 2.5 3 3.5 4Containm ent Pressure [kp/cm ^2]
Co
nta
inm
en
t H
yd
rog
en
[%
]H2 BURN
HYDROGEN
SEVERE
CHALLENGE
75% ZIRC REACTION
50% ZIRC REACTION
25% ZIRC REACTION
NOT
FLAMMABLEEach computational aid has a “background
document”
that provides the methodology and necessary input
to generate a plant-specific version.
Simple, non-computerized, graphical tools
(computational aids) were developed to
provide information required for decision
making process.
20
Procedures
Writing Team established from:
• Operations personnel (SE and SS to assure proper
linkage to EOP and up front familiarization);
• Engineering personnel (people included into
preparation of SAMG background documentation -
Setpoint, Plant Capabilities, CA);
Procedures and background documentation were
reviewed internally (NEK TSC members) and externally
(Westinghouse - WENX-00-05 and IAEA RAMP mission ,
IAEA-TCR-00959)
21
Procedures - Attachments
Generic WOG SAMG does not deal with possibility of fast conection and injection with mobile equipment
22
Review of NEK E- plan
• Transition from ERGs to SAMG
• Termination of SAMG
• Identification of personnel to evaluate SAM actions
• Special approval for intentional fission product releases
Performance of SAMGs – responsibility?
Emergency Response Organisation (ERO)
Technical Support Centre (TSC)
Operational Support Centre (OSC)
Emergency Operations Facility (EOF)
ERO Training
ERO Drills and Exercises
References
Krško NPP’s Emergency Response Organization (ERO) from Special Safety Review Final Report, Rev.0, October 2011
Review of NEK E- plan
Krsko NPP Proactive Measures before Fukushima
• Described in Special Safety Review
Final Report Available together with EU
EC Peer Review at.
http://www.ursjv.gov.si/en/info/reports/
EU response on Fukushima Event - Stress Tests
• 11 March 2011: Fukushima accident occurs
• 24 – 25 March: European Council Requests
– Stress tests to be developed by European Nuclear
Safety Regulators Group (ENSREG), the
Commission and WENRA
– Safety of all EU plants should be reviewed
– Scope of review developed in light of lessons
learned from Japan
– Assessments conducted by national Authorities
– Assessments completed by a peer review
Development of Stress Test Methodology
• Methodology drafted by WENRA in April
• Agreed to by ENSREG in May
• On 25 May 2011 ENSREG including the
European Commission published the ENSREG
declaration that described EU Stress Tests
methodology
Peer Review Process
• The Peer Review Board
• Three topical reviews in parallel, January and February 2012 – Initiating Events
– Loss of Safety Functions
– Severe Accident Management
• 17 country visits in 6 parallel groups, March 2012
• About 80 experts involved
• ENSREG Report + 17 Country Reports
General conclusion over Europe
• Significant steps taken in all countries to
improve safety of plants
• Varying degrees of practical
implementation
– Regulatory systems
– Extent of programs
Consistency of approaches in European countries
• Global consistency over Europe in
identification of:
– Strong features
– Weaknesses
– Measures to increase robustness
Measures to increase robustness of plants
• Significant measures to increase robustness already decided or considered, such as: – Additional mobile equipment
– Hardened fixed equipment
– Improved severe accident management with appropriate staff training
• Details available in Country Reports and Main Report
• Typical measures: – Bunkered equipment including
instrumentation and communication means
– Mobile equipment protected against extreme natural hazards
– Emergency response centers protected against extreme natural hazards and radioactive releases
– Rescue teams and equipment rapidly available to support local operators
Prevention of accidents resulting from natural hazards and limiting their consequences
32
Introduction of Krsko AMP
• Plant specific analysis (IPE – Individual Plant
Examination & IPEEE for external events) - plant
response on Severe accident
– Hazard Analysis (voulnerabilties identification)
– PSA Level 1
– PSA Level 2 including Containment fragility
analysis, deterministic analysis (MAAP),..
• NPP Krško specific full scope simulator with capability
to simulate severe accidents (based on MAAP models)
• Plant specific SAMG based on WOG generic
procedures
Concept of Krsko AMP
35
WOG Generic SAMG Implementation
• Review of WOG Generic SAMG applicability; • Development of plant-specific strategies
• Development of plant-specific SAMG setpoint;
• Development of plant-specific computational aids;
• Review of NEK EOPs to incorporate transitions to
SAMG;
• Writing of plant-specific control room SACRGs;
• Writing of plant-specific TSC guidance, including SAGs,
SCGs, DFC, SCST, and SAEGs;
candidate high level actions (CHLA) strategies and mitigate system/structure/component (SSCs) (based on OECD, IAEA and EPRI Severe Accident Management Guidance Technical Basis Reports (TBR) in comparison with NPP design, available SSCs and its applicability – NOT DIRECTLY APPLICABLE!!!
36
– Definition of transition (rules of usage)
– SAMG for MCR (should be similar to FR-C1)
– SAMG for Spent Fuel Pool (not available in generic
SAMG, important issue from Fukushima point of view)
– SAMG for shutdown (e.g. loss of SRH on midloop
operation)
– Alternative means (mobile equipment) usage:
• Different fire protection pumps
• Fast connections to the systems (e.g. injection into SGs)
• Source of waters (e.g. amount for flooding the containment to
protect cavity floor from MCCI OR even flooding the Rx cavity
to the top of acctive fuel to establish external cooling)
Additional Plant Specific Issues
37
Procedures - Attachments
Generic WOG SAMG does not deal with possibility of fast conection and injection with mobile equipment
38
Review of NEK E- plan
• Transition from ERGs to SAMG
• Termination of SAMG
• Identification of personnel to evaluate SAM actions
• Special approval for intentional fission product releases
Performance of SAMGs – responsibility?
Emergency Response Organisation (ERO)
Technical Support Centre (TSC)
Operational Support Centre (OSC)
Emergency Operations Facility (EOF)
ERO Training
ERO Drills and Exercises
References
Krško NPP’s Emergency Response Organization (ERO) from Special Safety Review Final Report, Rev.0, October 2011
Review of NEK E- plan
Emergency Operations
Facility Technical Support
Centre (TSC)
Operational Support Centre (OSC)
Main Control Room (MCR)
Off-site Support Organisations
Emergency Response Organisation (ERO)
Krško NPP’s Emergency Response Organization (ERO) covered by 30 EIP procedures
Krsko NPP Response to NEI 06-12 (B5b)
Emergency Plan Implementing
Procedures (EIP)
Emergency Operating
Procedures (EOP)OP
Severe Accident Management
Guidelines (SAMG)
EOP SAME/FLEX Attachments
Abnormal Operating
Procedures (AOP)
Examples: • PRI-5 Loss of RHR
during shutdown • PRI-6 LOCA during
shutdown
Fire Response Procedure (FRP)
Examples: • FRP-3.9.101 – MCR
Fire – Stabilization of plant in hot shutdown
• FRP-3.9.102 – MCR Fire – Cooldown plant from shutdown panels
MCR evacuated
- fire YES
NO
TSC Operable
Initial Response Judgement
42
Wet Cavity (passive injection) 2001 – IPE insight, preventing the late
containment failure by preventing the MCCI
Periodic safety review conducted each 10 years (1st 2004, 2nd 2014) and
among 128 different actions and modification taken affects plant ERO
(EIPs, AOPs, EOPs, SAMG, etc.)s:
On site AC power source enhancement (new third 1E DG)
As a consequence of new PMF analyses – PMF values increase from
6500 m3/s to 7100m3/s – upstream dike enhancement (increase)
Installation of PARs
Instalation of PCFV system
Major Plant Safety Upgrade: 2000 – nowdays
43
Krsko NPP Containment Geometry/Wet Cavity Modification
RWST
RWST + RCS + 2 ACCU
(1) (2) (3) (4) (5) (6) (7) (8) (9) (10) (11) (12)
PDS PDS Sequence Sequence Time Time Time Time Time Time Time Win Time WinFrequency Frequency (sec) (sec) (sec) (sec) (sec) (sec) (sec) (sec)
Reactor Core T > 650 K Onset of Support RV RV External RV Cavity
Scram Uncovery Core Plate Failure Failure Flooding Flooding
Melting (9)-(7) (10)-(7)
TEHAYN 1.282E-05 TRA12_1 3.80E-06 9 4171 4725 6126 6640 8440 1915 3715TEHANN 7.217E-06 TRA12_33 5.40E-08 9 4175 4734 6136 6530 8330 1796 3596ALLBYN 6.450E-06 LLO2_1 1.30E-06 0 2750 3301 4012 6698 14679 3397 11378
No CMT Heat Removal LLO2_1A N/A 0 4014 4635 5444 8909 15358 4274 10723
WUUUUB 4.424E-06 SGR9_1 3.00E-06 306 59661 62206 65039 68059 74381 5853 12175
SEHAYN 4.350E-06 SBO20_1 2.60E-06 0 2566 3148 4235 9636 11436 6488 8288
Increase Debris Mass Ejected SBO20_1A N/A 0 2566 3148 4235 9636 11436 6488 8288
SLLBYN 2.015E-06 SLO3_1 1.50E-06 558 61498 64944 122947 146458 181236 81514 116292
FR-C.1 Depressurization Fails SLO3_1A N/A 558 56027 60059 65320 69737 82423 9678 22364
VXXXXB 2.003E-06 ISL1_1 2.00E-06 9 20378 21327 22360 25621 32983 4294 11656
TEHNNN 1.819E-06 SBO63_37 3.90E-08 0 6004 6641 8217 10147 11947 3506 5306
Reduced Debris/Coolant CHF SBO63_37A N/A 0 6004 6641 8217 10147 11947 3506 5306
Reduced Spreadhout Area SBO63_37B N/A 0 6004 6641 8217 10147 11947 3506 5306
SLNNN 1.250E-06 INA3_1 6.80E-07 0 2638 3206 4307 15361 45670 12155 42464
One Fan Cooler Running INA3_1A N/A 0 2634 3212 4304 14710 39693 11498 36481
ALLBYI 9.330E-07 LLO2_2 1.60E-07 0 2630 3187 3890 6559 15390 3372 12203
TERAYN 4.177E-07 LSP6_1 4.10E-07 0 6200 6939 8594 25533 1.E+10 18594 1.E+10
N/A N/A CREEP6 N/A 9 4174 4732 6133 6467 1.E+10 1735 1.E+10
Minimum Time Window (sec) for RV External Flooding 1735Minimum Time Window (min) for RV External Flooding 29
Minimum Time Window (sec) for RV Cavity Flooding 3596Minimum Time Window (min) for RV Cavity Flooding 60
Note: The 1E+10 value is to represent a large time for the calculation since there is no RV failure.
3rd EDG – Design Modification Package 2010
• Enhancement of Emergency Power Supply at NEK is identified as
the most significantly contributing solution for reducing core damage
probability due to a seismically induced SBO event.
– Based on the latest updated NEK PSA model, the contribution of Station
Blackout sequences to CDF from internal events are around 30%.
– The sensitivity study results for installation of the a third full size 6.3kV
diesel generator are as follows:
• The resulting seismic CDF for both cases is 1.25E-05, which is a
reduction of 51.9% from the base case seismic CDF of 2.6E-05.
• The resulting internal events CDF is 1.91 E-05, which is a reduction
of 39.7% from the base internal events CDF of 3.17E-05.
• The resulting total CDF is 6.11E-05, which is a reduction of 29.9%
from the base internal events and seismic CDF of 8.72E-05.
New 3rd EDG building
New 3rd EDG fuel reservoir
Cable traces
3rd EDG – Design Modification Package 2010
• The plant flood protection system should be
designed to withstand at least 10000-year
flood flow with no inundation.
– New preliminary estimation:
• Q100 = 3290 m3/s
• Q1000 = 4040 m3/s
• Q10000 = 4790 m3/s
– With right-bank inundation, the left
bank should be protected from at least
MPF flow (currently estimated as 6500
m3/sec), i.e. right dike is lower than left
dike, as postulated in USAR.
Protection of MPF with increased left
level of dyke to 158.95m long 620m
(+1 m) and road to plant long 500m.
• Measures against Beyond Current Design
Flood (e.g. 10000m3/h): big area cannot be
protected, but ITS building on plant location
can be protected with removal shields and
improved doors.
Flooding Protection Re-evaluation 2005
47
Harsh Environment - equipment surveviability during DEC conditions?
Figure 1 3D Containment model for gamma calculation in AB/IB
Figure 1 Principle leakage scheme for dose and TH evaluation
100
101
102
103
104
105
Time [s]
Pre
ss
ure
(k
Pa
)
150
200
250
300
350
400
450
500
550
600
N EK ES
LOCA DBA
MSLB DBA
SBO limiting
envelope
Figure 1 NEK ES DBA and DEC RB pressure envelopes (log time scale)
10-2
10-1
100
101
102
103
104
105
Time [s]
Te
mp
era
ture
(C
)
40
60
80
100
120
140
160
180
200
220
N EK ES
LOCA DBA
MSLB DBA
SBO limiting
envelope
Figure 1 NEK ES DBA and DEC RB temperature envelopes (log time scale)
48
Example: HELB Influence of EQ
Temperature contours at z=1m, time 14.655s, scenario B
10-2
10-1
100
101
102
103
T ime [ s]
Tem
per
atu
re [
C]
30
40
50
60
70
80
90
100
N EK B low dow n break
PL
OT
FE
R V
1W
17
:50
:42
, 27
/04
/02
ib11_09_1x2be
ib12_09_1x2be
ib14_09_1x2be
ib15_09_1x2be
10-2
10-1
100
101
102
103
T ime [ s]
Tem
per
atu
re [
C]
36
38
40
42
44
46
48
50
N EK blow dow n break
PL
OT
FE
R V
1W
01
:31
:43
, 13
/05
/04
tv11_09fia
tv11_09fic
tv11_09fie
tv11_10fia
tv11_10fic
tv11_10fie
Without isolation
With isolation
Modification to protect the ITS SSC against loss of NNS systems
49
SAME (Severe Accident Measurement Equipment) modification
performed to extend existed mobile equipment and satisfy NRC B.5.b
(NEI 06-12) measures and requirements for all NPPs:
ensure equipment and personnel to manage serious fires and
ensure mobile equipment for:
Core cooling and Containment cooling,
Spent Fuel Pit (SFP) cooling.
In such manner the emergency such as a commercial aircraft crash on the plant can be managed.
Major Plant Safety Upgrade: 2000 – nowdays
50
Preliminary Post Fukushima Actions
• Response on STORE (Safety Terms of Reference) including NRC
Bulletin 2011-01, and WENRA stress report:
– May 2011, preparation phase of DMP
– June 15th, 2011, presentation of results and proposed changes to the KSC
(including SES 10CFR50.59, UCP and DMP)
– July 1st , 2011 presentation to the SNSA, approval
– July – August 2011, implementation (OL25)
– September 30th, 2011, testing and notification of new configuration
• Mitigative actions need to take into account the following scenarious:
– Loss of SBO and UHS without any off-site support 72 hours,
– Time windows > 7days, core damage postulated,
– Extreme external events (seismic, flooding, storms, etc.),….
51
Preliminary Post Fukushima Actions
Covering:
• Design Modification (establishing a new system “AE – Severe
Accident Management Equipment” covering the hardware connections
to available systems (AFW, MFW, CI, CS, IA, VA, etc.), local control of
SG PORVs and purchase the mobile DGs, mobile injection and flood
pumps, etc.)
• Software Changes (Associated SEOPs/SAMGs and Emergency
Program changes and purchase the personal protection equipment)
• Safety Function establishing:
– Alternative Residual Heat Removal through SGs (alternative feed/bleed
means) • E.g Rosenbauer pump FOX III ((60m3/hr, 15 bar), OR HS60 (60m3/hr, 11 bar)), fix connection downstream
AFW pumps/upstream FW cont. isolation
– Alternative Residual Heat Removal through RCS (PORvs with alternative
feed/bleed means)
– Alternative SFP makeup and cooling
Krsko NPP Response to NEI 06-12
EOP SAME/FLEX Attachment Example: • Format of standard EOP is used
• Detailed instruction for local operators and firemen
• System flow diagrams were corrected to clearly evidence possible SAME fact connections of alternate equipment
• SAME equipment on-site is tested periodically
• Pumps head/flowrate • DGs capacity and initiating • Pipe hose status, spare nozzles
and connections, etc. • Training and drill is annually performed
taking into account realistic scenario driven on simulator and including MCR, TSC, OPC and locall fire brigade
BDBA Earthquake Evaluation
Earthquake Design Basis
– OBE 0.15g
– SSE 0.3g
Probabilistic safety concept is
based on the assumption that
there is no completely safe
structures.
Any structure or structural
element has a probability of
failure load.
The calculation takes all the
variables which are statistically
processed and uses them in the
form of the distribution function
of a certain probability.
BDBA Earthquake Evaluation
Methodology:
• Identification of ETs
Success Paths
• Mapping to critical SSCs
• Determination of failure
mode with corresponding
HCLPF identified for each
critical safety function
and initiator / damage
state
• Evaluation of Plant Level
Seismic Margin (PDS
status, cliff edge,
remaining success path)
Implementation of NEI 12-06 (FLEX)
Added as EOPs Attachments (37 !!!) which are referenced to SAMGs if needed Revision of SAMGs
56
Alternative Means (FLEX)? • Discussion about possible solutions...
– Availability of FLEX fast conections?
– Availability of people to do fast conections?
– Available time window?
– Expanded SEOPs and SAMG for FLEX
Implementation of NEI 12-06 (FLEX)
Modifications to implement the
fast connection to DHR
systems (RCS, containment,
SFP) close to the containment
barrier
Verification and validation of
SAME:
• Scenarios (time
sequences) calculated by
MAAP + exercised by
simulator
• Drill of local operator and
plant fire brigade staff
Concept of Safety Upgrade 2013-2016
Revision of SAMGs due to implementation of BB2 (alternate UHS, SI, SFP etc.) new ECR - schedule 2016
EU Peer review conclusions
A comprehensive description of the behaviour of the plant, as well as the various ways of ensuring fulfilment of the main safety functions during the scenarios of LOOP and loss of UHS, is provided in the report.
According to the information provided in the SI-NR, the present accident management organisation appears to be well structured and adequate to cope with different levels of severity in the event of an accident, including severe core damage.
WENRA safety reference levels have been implemented in the Slovenian regulatory process. Moreover, the Slovenian regulatory assessment also applies selected USNRC regulations, standards, guides and good practices.
An especially noteworthy characteristic of the Slovenian SAM organisation is the validation of the SAMG using a full-scope simulator.
Several provisions are already in place to support SAM with the use of mobile equipment. It is important that upgrading measures identified to improve SAM capabilities (e.g. installation of PARs, filtered venting, new emergency control room, third engineered safety features train) are implemented as planned.
References
[1] "Krško Source Term Analysis"; paper presented at the 2nd Regional Meeting
"Nuclear Energy in Central Europe"; Portorož, Slovenia, September 11-14,1995.
I. Basic, B. Krajnc (NEK);
[2] “Methodology and Results of the Krško Level 2 PSA”; paper presented at the
International Conference on Nuclear Containment”; Robinson College University
of Cambridge, England, September 23-25, 1996., R.P Prior (W), M-T.Longton
(W), R.Schene (W), B.Krajnc (NEK), J. Spiler (NEK), I. Basic (NEK);
[3] "Reanalysis of some key transients with MAAP code for NPP Krško after SG
replacement and power uprate"; paper presented at the International Conference
“Nuclear Energy in Central Europe"; Portorož, Slovenia, September 6-11,1999. I.
Basic, B. Krajnc, B. Glaser, M. Novsak, J. Spiler (NEK);
[4] "NPP Krško Severe Accident Management Guidelines Implementation”; paper
presented at the international conference “Nuclear Option in Countries with Small
and Medium Electricity Grids 2002"; Dubrovnik, Croatia, June 17-20,2002., I.
Basic, J. Spiler, B. Krajnc, T. Bilic-Zabric (NEK);
References
[5] “Prioritization Of The Recovery Actions In The Krško NPP SAMGs”, IAEA-
NUPEC Technical Meeting on Severe Accident and Accident Management,
Toranomon Pastoral, Minato-ku, Tokyo, Japan, 14-16.03.2006I. Basic, I.
Vrbanićem (APoSS), T. Bilić-Zabric (NEK)
[6] “Upgrade of Krško Level 2 PSA Model for Regulatory Activities”, “International
Conference Nuclear Energy for New Europe 2008”; Portorož, Slovenia, 8-11.09.
2008.; I. Vrbanić, I Basic (APOSS), S. Cimeša (SNSA);
[7] NPP Krsko Special Safety Final Report, Rev 0, October 2011 (from
www.ursjv.gov – Slovenian Regulatory Body)
[8] IAEA RAMP report IAEA-TCR-00959 is available on
http://www.ujv.gov.si/fileadmin/ujv.gov.si/pageuploads/si/Porocila/PorocilaEU/ra
mp_krsko_final.pdf.