practical examples of samg

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


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


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.

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 – 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


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


• 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?


Technical Support Centre (TSC)


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


Emergency Operations

Facility Technical Support

Centre (TSC)


Main Control Room (MCR)

Off-site Support Organisations


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

IB4 - 3EDG.pdf

IB4 - 3EDG.pdf

• 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

NEK_hisa_xvid.avi

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

NEK_EQ_report-TR06_07-0.pdf



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.),….

NRC_BU_2011-01.pdf

NRC_BU_2011-01.pdf

NRC_BU_2011-01.pdf

NRC_BU_2011-01.pdf

NRC_BU_2011-01.pdf

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

STORE_short.ppt

STORE_short.ppt

STORE_short.ppt

STORE_short.ppt

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)

Evaluation of Margins (Earthquake)_ib_20110614.pdf

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.

http://www.ursjv.gov/