XA0300582 - IAEA/NSNI LIMITED DISTRIBUTION
INIS-XA-583
WVORKINGMATERIAL
NOTES ON THE
REGIONAL WORKSHOP ON MODELLING OF EXTERNAL HAZARD S TC PROJECT: RER/9/046
6-10 November 2000 Sofia, Bulgaria
Reproduced by the IAEA Vienna, Austria, 2000
VOLUME I
NOTE
The material n this document has been supplied by the authors and has not been edited by the IAEA. The views expressed remain the responsibility of the named authors and do not necessarily reflect those of the governments) of the designating Member State(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this document. XA0300583
Fire Hazard Assessment
by: M. Roewekamp FIRE PSA METHODOLOGY AND APPLICATIONS
M. Roewekamp Gesellschaft f~ir Anlagen und Reaktorsicherheit (ORS) mbH
Workshop on Modeling of External Hazards in PSA Sofia, Bulgaria, 6 - 10 November 2000 Contents
o:Pa-rt 1: Screening for Fire PSA -GRS Methodology c Part 2: Fire Simulation with the GRS-Code COCOSYS
cPart 3: PSA Study for an Exemplary Plant Location of a German PWR Built to Earlier Standards o Part 4- Results of recent sensitivity and uncertainty analyses Screening for Fire PSA GRS Methodology Major Steps of a Fire PSA
• Selection of relevant plant areas
* Detailed fire specific analyses for selected plant areas
* Implementation of fire specific results in existing PSA
Screening Steps
• Data collection Data collection of all information for each compartment with relevance to fire
* Screening Estimation of those rooms (including ranking) with a necessity of fire specific detailed probabilistic analyses Steps of the Screening (1)
Fire specific compartment data
- Compartment identification * Building name * Compartment number • Building level
- Adjacent compartments and connections • Openings (to adjacent compartments) • Doors, type of doors (to adjacent compartments) * Openings for monitoring devices * Fire and smoke dampers * Fire barriers (rating) Steps of the Screening (2)
Fire specific compartment data
- Fire specific data • Fire load density (in MJ/M 2)
• Combustibles (PVC, FRNC, oil, wood ... ) • Fire detectors (automatic, press button) • Fire extinguishing systems
- Compartment inventory (safety related equipment incl. cable) Steps of the Screening (3) Screening procedure
- Excluding criteria • Fire load density < 90 MJ/M2 * Inventory criteria (compartments not containing safety related equipment without openings to adjacent ones with such equipment are screened out)
- Ranking criteria * Fire load • Fire occurrence frequency • Fire spreading
- Expert discussion • Combination of compartments * Representative plant areas Data Collection - - Example 0208*1 2 8 T Antrebe D-Forderpumpen K 02 084 T
PVC. 012 A~~~ V ertefliung n
02082 KLB1SAAO1 1
lBrandlast PVC und 01 fr die Antriebe der 3 H D-Forderpumpen, Brandlast wesentlich groBer 90 (nach Begehung 17.01.99);
Datensatz: 211-62 II von 271 04 03 04 0 1 4 ~~~~~~~~~~0417 04 11 )4 Data Analysis (1) II0 1 1.MMM MM0 (slelolla) ) OK BK BK BHK BK *BK
0301803002 ~~~~~030811
03 019 ~~~~~~~~~03003 0377 0=70 0~~~~~~~~~~=43 Ef
03 000 ~~~~~~~~~~~~~~~~~~03072 B~~144 0=~~B144
M A
B=20 028 H27 211 T 90 02071A 02
R ~~~~~~~~~~~~~~~02066 R B-~~~~~~~~~~~~~~~~~~032
02 074
02 010 7 A~~~~~~~~~~~~2
1M 0 o o~~~~~~~~~~~~~~~~0
01
00 _ 01 02 07 I1 Data Analysis (2)
Building Number of compartments Fire load group No. of compartments UBA switchgear building 201 No fire load 59 UBP emergency power supply building 54 Fire load" 90 MJ/m2 72 UJA containment 136 Fire load > 90 MJ/m2 140 UJB3 reactor building annulus 271 Fire load density (> 90 MVIJ/2), UKA nuclear auxiliary building 331 groups of compartments and connected ULB emergency feedwater building 94 compartments UMA turbine building 67 2870031005 U1QB cooling water pump station 9 03 034 03 034 03 058 03 058 Level of UJB1 No. of compartments 03 071 03 071 01 77 1900 02 39 01 024 01 024 02 024 03 36 01 025 01 025 02 025 04 34 01 027 01 027 02 027 05 17 01 028 01 028 02 028 06 28 1365 07 16 03 025 03 025 04 025 08 13 03 028 03 028 04 028 09 11 1150 03 024 03 024 04 024.. Estimation of the F 0 t-
9 Occurrence Frequency I.,-$TTTI4 1'9
Example (1)ttemr _ ____
------
C ______
1~~~~&~~~Ftt*~~~~irr \ a~Kwn Estimation of the Occurrence Frequency Example (2)
Inside building UJ8 a conditional fire occurrence frequency was estimated for 139 of 271 compartments, minimum value 6,61E-04 average value 7,19E-03 maximum value 1,82E-02 1,82E-02 01 004 1,62E-02 01 035 01 043 01 062 01 086 02 083 1,57E-02 01 055 1,37E-02 05 033 05 058 05 071 1,31E-02 05 014 A 1,28E-02 01 006 01 007 1,26E-02 01 005 01 022 01 033 01 044 01 045 01 056 01 057 01 063 01 072 01 087 01 088 01 089 02 016 02 037 03 016 03 035 03 072 1,22E-02 06 018 B 06 035 06 036 06 037 A 06 037 B 08 060 B 09 033 9,94E-03 06 033 8,92E-03 01 091 01 092 02 081 04 003 04 014 04 035 04 057 04 072 8,01E-03 01 009 01 017 01 036 01 046 01 060 01 073 02 009 02 017 02 036 02 060 02 073 02 084 03 017 03 036 03 060 03 073 04 042 04 060 04 061 04 064 04 066 04 073 04 079 6,48E-03 03 009 04 074 08 015 B 08 033 08 060 A 08 061 08 073 09 034 09 060 etc. Fire Specific Event Tree Example (1)
Fire detection: * Presence of humans * Availability of automatic fire detectors in the compartment and in the adjacent compartment Fire extinguishing (early): * Presence of humans * Availability of manual fire extinguishing and fire extinguishing systems Fire containment: * Compartment open * Compartment not open (fire doors, fire dampers) Fire Specific Event Tree Example (2)
N - S~~~~~~~~~~~~~~~~~~~~~~~~~~rn Bxn, Bad-B &'
UJB~~~~~~~~~~~~~~~~~,E Selection of Fire Zones *Ranking
- Fire load
- Occurrence frequency (ntroduction of a cut-off value)
- Fire effects / consequences
- Combined ranking *Expert selection
- Discussion basis: original raw data, processed data due to the above mentioned ranking
- Considering of further data: plant operational and safety related inventory
- Combination of compartments
- Representative compartments XA0300584
Plant Walkdown
by: M. Kostov REGIONAL WORKSHOP ON EXTERNAL EVENTS PSA 6-1 0 November 2000, Sofia
Plant Walkdown Presented by Dr. M.Kostov Risk Engineering Ltd
Content:
1. Preparatory steps for performing plant walk-down 2. The objectives of the (first) plant walkdown 3. Plant walk-down procedures 4. Earthquake screening evaluation 5. Walk-down documentation 6. Second Plant Walkdown 7. Conclusions
Based on: 1. IAEA TECDOC-724, Probabilistic Safety Assessment for Seismic Events 2. NUREG/CR-4482, Recommendations to the NRC on Trial Guidelines for Seismic Margin Review of NPPs 3. UCRL-ID3-115714Rev2, Walkthrough Screening Evaluation Field Guide 1. Preparatory steps for performing plant walk-down a The site seismic hazard (or RLE) is known. m The system modeling (event tree) is preliminary elaborated. m The systems, structures and components are preliminary grouped
and categorized. a The groups for generic fragility are preliminary set 2. The objectives of the (first) plant walkdown are: 1. To mark which component could be modeled by generic data and which have to be address on plant-specific basis. To confirm that no weaknesses exist in the plant structures and equipment that would make their HCLPF lower than the generic values and to look for signs of abnormal aging or poor maintenance that would invalidate the use of generic values.
2. To confirm the accuracy of system descriptions found in plant design documents (e.g., FSAR, general arrangement drawings, piping and instrumentation diagrams, and line diagrams for electrical equipment).
3. To identify any system interactions, system dependencies and plant unique features not already identified.
4. To gather information on certain potentially weak components for further HCLPF (fragility) calculations Walk-down team:
THE SYSTEM ANALYST In defining these items, the system analyst should consider: • those components that comprise safety related systems. • potential systems interactions, i.e., component and system failures of non-safety related items that can lead to failure of systems performing safety functions
He should mark these items on general arrangement drawings, piping and instrument diagrams and line diagrams, for electrical equipment.
THE FRAGILITY ANALYST: By reviewing such marked up drawings should be able to identify the plant areas that require inspection. The fragility analyst should consider: • The element seismic capacity • The potential system interaction
The team members should have: • Detailed knowledge on the analyzed facility (NPIP) • Experience in earthquake engineering • Studied other (similar) PSA projects • Knowledge on earthquake fragility data (experimental or real) 3. Plant walk-down procedures
• After performing the review by the systems and fragility analysts to become familiar with the plant systems, structures, components they identify areas of the plant requiring physical walk-down.
• It is expected that the system analyst will review all areas of the plant and provide advice to the fragility analyst on the function of any component in the safety systems and the consequences of its failure.
The steps in performing the first physical walk-down are described below: 1. A pre-plant visit meeting between system and fragility analysts to plan the plant walk-down and discuss areas to concentrate on should take place. 2. Necessary arrangements must be made with the plant management regarding radiation protection and scheduling the walk-down activities to create minimal conflict with normal plant operations. The walk-down team should either include or have access to the following: • A reactor operator or utility engineer familiar with the plant systems, • An electrical technician capable of de-energizing and opening electrical cabinets for anchorage inspection. 3. An orientation meeting among the systems analyst, the fragility analyst, and the plant operating personnel should take place for exchange of general and plant-specific information.
4. The walk-down by the systems analyst is expected to: • Verify correctness of system configuration models • Look for potential systems interactions * Verify locations of each piece of equipment • Identify unusual features in the plant • Advise on the significance of different failure modes of components identified by the fragility analyst.
For example, the following failure may or may not be significant to core melt: • Objects (e.g., roof slabs, adjacent unreinforced masonry walls) falling on the component or on the electrical cabinets supplying power and control to the component. • Tank failures flooding the vicinity of the component. • Direct seismic failure (e.g., anchorage failure).
Certain valve failure modes may or may not be important. For example, valve stem sticking may or may not be important depending on whether the valve has to move or not during an earthquake. Identification of potential failure modes of a given component will assist in decisions on the significance of different systems interactions. During the plant walk-down, the systems analysts should record the above information directly on the simplified schematics that were developed in the preparation phase. The arrangement of the equipment being reviewed within rooms or vaults should be recorded along with the location of any other items whose failure may possibly affect the equipment or components. These items can be drawn directly on the schematic. The room or vault identification and elevation should be recorded. The location of electrical equipment, and if appropriate, its attachment location should be shown. The location of piping penetrations through walls should also be indicated. Any overhanging piping, dueling, and/or equipment, such as HVAC, should be indicated along with the existence of tanks and floor drains within the equipment rooms. like this one. o- 4
Whileispectin aCparicua icgfeqimn ftefrgltnls veiieha o eknseseisthsreut hul e oedad a alsobe ndicatedongesmlfedshmtc 5. One objective of the walk-down by the fragility analyst is to confirm that no weaknesses exist in the components judged to have generically high capacities and to uncover any systems interactions. In the (first) walk-down, the fragility analyst will concentrate on safety related components and systems interaction that may affect them. The output of this effort will be to assign components into categories: a Most rugged component, usually screened out (not considered in the further analysis) a Components for which the generic fragility data may be used a Components for which specific analyses are needed a Weak components (directly affecting the plant risk assessment): • repair • specific analysis • other consideration.
By a review of design documents the fragility analyst will have by this stage identified components to focus on during plant walk-down. He should have become familiar with the plant arrangement, and details of equipment supports and anchorage. The items to be inspected in this first walk-down are a function of the earthquake review level. The number of items increases with the earthquake review level. 4. Earthquake screening evaluation
The evaluation should include:
m Anchorage
a Load path
a Structural integrity
m Operability (including relay chatter) a System interaction a Building adequacy RELATIVE CAPACITY LEVELS FOR EQUIPMENT Relative Equipment category capacity level High Piping, ducting, cable trays and electrical conduits, valves (except small motor operated valves) Medium-High Small vessels and heat exchangers, horizontal pumps, compressors and turbines, fans and air conditioning units, diesel generators, reactor coolant loop components Medium Small motor operated valves, large vessels and heat exchangers, batteries and racks, vertical pumps, reactor internals and control drive mechanism Low-medium Motor-control centers, switchgear, control panels, control panels, instrument rack, non-seismically qualified components (e.g. offsite power system) Most Common Failure Modes due to Seismic Loading Component category Failure modes Reactor coolan loop 1.Supports components Piping and ducting 1. Supports 2. Excess deformations Vessels ad Heat Exchangers 1.Supports 2. Vessel nozzles Pumps and Fans 1.Supports 2. Nozzles 3. Shaft deflections Electrical cabinets 1.Chatter 2. Trip 3. Structure (anchorage) Cable trays and electric 1.Supports conduits 2. Tray or conduit run Valves 1.Operator support (Yoke) ______2. Operator 4. Earthquake Screening Evaluation
The items to be addressed for walkthrough screening evaluations for earthquake effects include anchorage, load path, structural integrity, operability (including relay chatter), systems interaction, and building adequacy. This section provides many example illustrations for the items discussed for screening evaluations. Some of the figures present good seismic engineering details and others show configurations that may result in poor performance and should be screened for .* further evaluation. The common stashed-circle .no" symbol is drawn over the illustrations with poor seismic details.
Figure 1: Shell-type anchor
4. 1. Anchorage
Adequatc z.-z:"rage is-alc l~ways ecential to the survivability of any item. Anchorage should always receive special attention during walkthrough screening evaluations. Anchorage integrity requires adequate installation and capacity. Example concerns for each of the major anchorage types (i.e., expansion anchors, cast-in-place anchors, and welded anchors) follow.
a E42ansion Anchors. The shell- type or displacement controlled (see Figure 1) and wedge-type or torque controlled (see Figure ...... 2) expansion anchors have . been widely tested and have reasonably consistent capacity...... when properly installed in sound concrete. Some effort should be spent to determine the type of anchors used (e.g., look at Fgr 2 eg-yeaco abandoned anchors, interviewFgue2Wde-peacr construction or maintenance personnel, or review installation specifications). Other types of nonexpanding anchors such as lead cinch anchors (see Figure 3), plastic inserts, and lag screw shields are not as reliable and should be screened as a potential concern, especially if seismic demand loads are judged to be close to the capacity of the anchor. Items to consider when reviewing expansion anchor bolts include: Bolt embedment length may not be adequate if ~ ~ ~ part of the shell is exposed or if there is a Figure 3: Lead cnch anchors have lower capacity
long stud protruding ______above the concrete surface. nadequate M embedment also may result from use of shims , or tall grout pads as shown in Figure 4. Taoul-2'1 Pan'~ About 10 bolt diameters' spacing aid edge distance are required to gain full capacity (see Figure 5). Cracks in the concrete may reduce the bolt
capacity. ______Figure 4: Tall grout pad resultszn minimal anchor Anchors in damp areas or embedment harsh environments should be checked for corrosion deterioration if EDGEDIST =¶0 D SPACING =10 D heavy surface rust is observed. Expansion anchors may have low resistance to imposed bolt bending moment (may result from
gaps between base and 2- floor).
Loose nuts may indicate .DAEE0
inadequate anchor set. 2
Figure 5: Edge distance and spacing should be 1 0 D minimum '44.
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Figue8 Unerstdasen igrc)hnorblt r deirae anchrorsesinc hefilurenoes dcilepit Figure 10: Raised pad with no connection to Figure 11 Welded anchor concrete floor Items to consider when reviewing ClIP bolts include: Raised pads should be checked since embedment may not extend down to the floor slab as shown in Figure 10 (adesign drawing review is usually the only way to check embedment length). About 10 bolt diameters'spacing and edge distance are required to gain full capacity.
CIP anchors in badly cracked concrete may have significantly reduced capacity. Anchors in damp areas or harsh environments should be checked for corrosion deterioration if heavy surface rust is observed. a Welded Anchors. Well-designed and detailed welded connections A to embedded plates (Figure 11) or structural steel provide high- capacity anchorage. Items to consider when reviewing welded anchors include., Presence of weld bum- through at light-gage steel indicates weak connection. Line welds have minimal resistance to bending
moments applied about the ______axis of the weld--this maly Figure 12: Plug welds have low capacity to resist occur when there is weld only uplift on one side of a flange.
Puddle.welds and plug welds used to fill bolt holes in equipment bases (see Figure 12) have relatively little capacity for applied tensile loads. p~~~~ ~ W
PN N~*~'.~. 'O' 0,~
clips provide no positive"N Figureshims 13: mayWelds haveacross very low Figure 14: Friction
capacity anchorage~~~~~~~~~~~~~~~~~~~~~~~~~~~N.~
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very small effective throat area and thus low capacity (see Figure 13). harsh environments should be checked for - ~ded anchors in damp areas or corrosion deterioration.
4.2. Load Path from the item to its In addition to anchorage, an adequate structural-load-resistiflg path is needed point. During screening evaluations, the anchorage inspection should also anchorage attachment stiffness, and review the load path through the item to its anchorage for adequate strength, carefully Example anchorage ductility. The review must check the connections as well as the support members. load path features to review include: ______
Friction Connectors. Friction connections such as holddown clips (see Figure .14) often pry off or completely slip out of place during seismic loading and become completely ineffective. Adequate anchorage requires positive connection.
m Vibration Isolators. Vibration isolators such as that shown in Figure 15 must be retrofitted with bumpers that provide a load path in lateral and vertical poorly directions. Figure 15: Vibration isolators perform *U Thin Framing and Cipo Angles. Thin framing members and clip angles may lack the strength and stiffness required to transfer loads to anchor bolts. Stiff load paths with little eccentricity are preferable for anchorage. Figure 6 shows an example of a good anchorage load path for an electrical cabinet. Sheet Metal Enclosures. Thin
cabinet bases should be - reinforced with steel angle framing so that seismic loads Figure 16: Preferred stiff anchor
may be transferred to anchor ______points. In addition, oversized washers are required when anchors are bolted directly through thin sheet metal bases (see Figure 17). WeaK-way Bending. Heavy components that are mounted on upright channel sections ~ may need to rely on weak- way bending of the channel to transfer shear loads to the anchorage (see Figure 18)......
Unstifened, light-gage . channels may not have eoug strsengt tohnlhs Figure 17: Oversized washer reinforces sheet metal load transfer. ~~loadpath Tie Downs to Supports. Distribution systems or equipment components such as tanks should be positively attached to their support bracket or saddle. Examples of concerns include: Distribution systems on supports that allow for sliding should have end restraint to limit movement.
Figure 18: Weak-way bending provides poor load path ~~\'.'*'' ~ ~ ~ W.
Figure ------&~~~~~~---
Fiue19: Pipe shoe without lateral restraint may Figure 20: Tanks not tied to support may slide fall off end of bracket
Figure 19 shows an example pipe support that lacks guides. Unsecured tanks (lack of tie down) such as shown in Figure 20 are free to slide or rock, which can cause attached piping to become damaged. Battery racks must have wrap-around restraint to prevent sliding and toppling of batteries.
Beam ClampS. Beam clamps should not be oriented so that loads are resisted by fictional clamping (see Figure 21). Eccentricities in the load path cause prying and are also a concern.
4.3. Structural Integrity
Structural integrity is important. Components required to function after an earthquake must ~~% remain intact as well as remain in place. As with // ' anchorage load path, this requires adequate strength, stiffness, and ductility. Examples of specific structural integrity features to be alert for include:
m Connections. The adequacy of the connections in the structural system's load-resisting path is a key item for structural integrity. i 4 Example features to check for include:
Missing bolts in some Figure 21: Clamps resisting loads by friction may holes of connected parts. slip off Structural members S with oversized holes or Fiueflame-cut holes (see
Nuts without full thread engagement. Weld discontinuities (such as cracks, undercut, rollover, undersized leg, and
inadequate effective '" thoa) (e Fgr Figure 22: Oversizedflame-cur holes result in poor connections
Cotter pins not installed ______properly (ears not bent over).
Fasteners missing\ positive locking devices for vibrating equipment. Nonductile Materials Nonductile materials in the load- resisting path should be screened out and evaluated in more detail. Example nonductile materials include ceramics (used as insulators), cast iron (used in piping joints, see Figure 24), and plastic (used in restrainers and clips). Figure 23: Example of poorfillet weld profiles
Cut-outs and Coping. Cut-outs an in side panels of electrical cabinets or HVAC ducting h should be reinforced.
Figure 24: Bell-and-spigot piping joints of cast iron are brittle '\n~~~~~~~
Buckat latch
Figure 25: Structural members coped out (to give Figure 26: Doors and latches on cabinets should be clearancefor piping) have reduced strength secured Unreinforced cut-outs reduce the shear-load-resisting strength of sheet metal. Structural members wfth excessive coping (see Figure 25, for example) of flanges and/or webs may have significantly reduced section properties (and strength). m Door Latches. Doors of electrical cabinets should be positively latched or screwed shut to maintain continuity of the cabinet shear panel resistance. Internal assemblies such as motor controller devices (see Figure 26) should also be atched to prevent contactor damage. a Subcomponents and In-line Components. Heavy internal components and in-line components should be located near anchor points and positively restrained. Lack of restraint may result in significant damage to the parent component or distnibution system. a Rod Hangers. Rod hanger supported items behave well, provided attached branch lines can accommodate seismic-induced displacements (see Figure 27). Short, isolated rod hangers with fixed-end connection details (see Figure 28) and heavy supported loads may experience a fatigue failure and loss of structural integrity.
Figure 27: Branches are vulnerable to header Figure -28: Fvxed-end rods are subject to fatigue seismic anchor motion damage a Flexible Joints. Flexible joints such as bellows or Dresser K2 couplings (see Figure 29) n cannot transfer loads from their attached distribution systems. If An there is no independent restraint M to handle these loads, structural integrity may be lost. Rack Framing. The framing system for racks such as battery racks and instrument racks should be capable of resisting lateral loads. Longitudinal cross bracing is preferable for racks supporting station batteries. Figure 29: Flexible couplings without pipe Within the context of structural integrity, manyretanscnohndeipedielas types of mechanical equipment components are inherently very rugged and need not be subject to more than a brief review of anchorage and load path. Examples include compressors, pumps, motors, engines, and generators.
4.4. Operability
Component operability needs to be maintained for systems that must function during or immediately after a seismic event. A relay chatter evaluation should be performed for electrical components whose operability must be maintained during an earthquake. In addition, this often requires a detailed review of differential displacement considerations and subcomponent response. Examples of features to be alert for during walkthrough screening evaluations for operability include:
a Failure Position. The failure state or failure position of certain components, even possibly those that are not required to operate, should be reviewed to check if the failure state may render a system inoperable. For example, ventilation dampers may not be required to operate but may restrict flow or adversely affect vital system performance if they fail closed. m Relays. A systems review is generally needed to identify relays whose chatter or inadvertent change of state during seismic motion may result in loss of function of vital equipment. This generally requires more effort and time than is available for a walkdown screening evaluation. However, during walkthrough screening evaluations, equipment should be reviewed for the presence of trip- sensitive (or 'low ruggedness') devices. Examples of these include: Mercury switches (see Figure: 30). Sudden pressure switches (such as Buhnoltz relays on switchyard transformers). General Electric relay types CFD, CF`VB, CEH, CPB, IJD, HGA, PVD1 1 and 12, RAVII, and HFA65. Westinghouse relay types HLF, HU, ITH, ARMLA, PMVQ SG, SV, SC, SSC, and COM-5. vibrauon sensitive switches attached conduit~~~~~~~~~Cndui
vibTabinn-and CondusitcSackeAtacedtiand conduit ne ohv nuhfeiiiyt
accommodate seismic and thermal motion of components. Watch for conduit attached to in-ine components offie~xjiepiping systems (see Figure .31). Transformer oils. nternal coils of dry-type transformers need to berestrained to prevent short-out by contact with their sheet metal enclosure (see Figure 32). This internal inspection requires partial disassembly of the transformer cabinet. Similarly, busbars must have adequate clearance where they penetrate cabinet sidewalls to avoid short-out during seismic motion. a Electrical Euipment. Subcomponents of electrical equipment with electrical contacts (such as switchgear breakers with secondary contacts) need to have adequate support and stiffness so that motion does not cause contact damage or misalignment. Figure 33 shows a favorable restraining bracket on a low- voltage breaker unit.
Figure 32: Core coil bolts should be in place Figure 33: Restraining bracket prevents damage of cotcson breaker Batteries more than about 10 years old may become brittle and be vulnerable to seismic loading. Batteries should also have spacers between adjacent cell jars to prevent jar impact and transfer of loads through busbars and terminal posts.
a Mechanical Equipment. The motor driver and equipment item (such as pump, fan, or generator) should be attached to a common rigid skid (see Figure 34) to
prevent shaft binding, which leads ______to inoperability. Figure 34: Common rigid skid is favorable mounting 4.5. System nteraction
Potential seismic systems interaction hazards need to be evaluated. Seismic systems interaction is the physical interaction (bumping, falling) of items close to one another. Vital components with fragile appurtenances (such as instrumentation tubing, air lines, and glass site tubes) are most prone to damage by ltr~cicns. Th tpDes of seismic systems interactions that need to be reviewed include:
M Structural Failure and Faling. Inadequately anchored and unanchored components may slide or topple and fall, causing damage to vital components (see Figure 35). Plant operations, safety, and maintenance equipment as well as facility architectural features are commonly overlooked in seismic design programs and present sources of seismic interaction concerns. Examples of potential seismic interaction failure and falling sources include the following:
Partition walls and unreinforced masonry block walls. Ceiling tiles on unrestrained T-bar grid systems. Overhead walkway platfor m grating lacking___ tie-downs. Suspended light fixtures and fluorescent tubes. Storage cabinets, files, and bookcases. Tool carts on wheels and tool chests.
Ladders and scaffolding. Figure 3: Failureand falling interaction hazards ISLEPIP
Figure 6: Upper and lower restraintsare required Figure 37: Proximity and impact interaction hazard for gas bottles
Portable testing equipment. Unrestrained gas bottles (see Figure 36) and fire extinguishers. Unrestrained equipment on wall-mounted supports. Proximity and Impact. Adjacent components may impact each other and become damaged if there is not adequate clearance between them to accommodate seismic- induced deflections (see Figure 37). This is especially a concemn if one of the items is fragile or if one of the items has sufficient mass, hardness, and response (energy) to damage the other. Examples of impact interaction sources to investigate for include: Flexibly supported piping, ducting, and raceways (consider also thermal motion of piping) in proximity to vital equipment. Flexible electrical panels in proximrity to walls and columns. Suspended equipment components such as room heaters and air W conditioning units. a Differential Displacement. Distribution systems that span between different structural systems need to have sufficient flexibility to accommodate the ~~. differential motion of the supporting structures (see Figure 38). Piping may be vulnerable where it Figure 38: Differential displacement interaction interfaces with a building structure ~r foundation. hzr K'~~~~~~ E
SPRiNKH.DWW 500 i-ll ' ' '~ .Dw8- ?I
Figure. 39: Fusible link sprinkler heads are sensitive Figure. 40: Pipe break potentialfor unancho red to impact tanks
a Spray/Flood/Fire. Potential seismic-induced spray, flood, and fire interaction sources should be evaluated. Example sources to watch for include:
- Hazardous/flammable material stored in unanchored drums. * Nonductile fluid-carrying pipe (such as cast-iron or PVC pipe).
- Fire protection piping with inadequate clearance around fusible link sprinkler heads (see Figure 39).
- Unanchored and poorly anchored tanks with nonflexible attached piping (see Figure 40).
4.6. Building Structure Review
The walkdown screening evaluation should also include facility building structures. This is especially true when the building structure forms a confinement banier for hazardous materials. For complex building structures, htis advantageous for a separate walkdown team to concentrate solely on the building structure. As a minimum, the following should be reviewed:
a Lateral-force-resisting System. The lateral-force-resisting system of the building structure (including roof system and floor diaphragms) should be identified and reviewed for strength, stiffness, and ductility. It is important to concentrate on the connection details between structural members (e.g., beam-to-column and roof diaphragm-to-wall connections). For the overall structure, any irregularities in stiffness or mass that could result in excessive torsional loading should be examined. Irregularities in the load- resisting path such as a weak first story or diaphragm or shear wall discontinuities should be evaluated.
a Foundation. The type of foundation and the anchorage of the building to the foundation should be evaluated. Inadequate foundation and anchorage can result in the loss of a facility. Exterior Walls. The type of exterior cladding and its attachment to the building structure should be identified and evaluated. Use of unreinforced masonry, especially if in the lateral-load-resisting path, should be reviewed in detail by methods other than a walkthrough screening evaluation. Soil Conditions. Soil conditions that may affect the integrity of the facility should be identified for further review. Example conditions. to check for include liquefiable soils, unstable slopes, and possible uneven settling. If geotechnical reports are not available, a soils evaluation specialist should be consulted. 5. Walk-down documentation
The systems analysts should document the following:
• The accuracy of the design documents and the systems descriptions for the various safety systems and their support systems. • Any systems interactions or unique-plant global features uncovered during the walk-down.
In addition, the systems analyst should have a complete set of annotated schematics of the safety and support systems along with any additional documents, analyses, and results of discussions with plant personnel on the configuration and operation of the plant under review. The fragility analyst should document the following:
• For each class of components judged to possess high HCLIPF values (larger than the earthquake review level), • state that anchorage and supports are adequate and similar to the generic component • provide example photographs showing the overall view and details of support and anchorage systems of components screened out. • state that no weak spots were observed in the local (supporting) structures that may make the component vulnerable to earthquakes • state that there are no seismically weak objects (e.g., block walls) near the equipment, and that there is no potential for flooding caused by tank failure or fire caused by obvious electrical problems. • Any observed systems interactions and their possible effects on the safety systems and components. • Information needs for components requiring detailed review. Simplified Analysis
Following the first plant walkdown, the fragility analyst may screen out the potentially low capacity components using the data collected in the design review and walk-down. This is accomplished by performing some simplified analysis.
The first step in this analysis is to estimate the response that the component experiences at the earthquake review level. Two candidate approaches-conservative deterministic failure margin method and fragility analysis method, are proposed to estimate the HCLPF values of components. In both methods, some reanalysis of structures and equipment may be needed to calculate the seismic responses. Such reanalysis (of structures at least) should be done before the second plant walk-down is conducted. If the structural responses (e.g., floor spectra) calculated for the plant SSE can be scaled to RILE responses, the effort of seismic margin studies would be considerably reduced. For this purpose, the fragility analyst should review the structural models to confirm the adequacy of the models and the appropriateness of scaling the responses. Next, he will make analyses to judge if the components' seismic capacity exceeds the earthquake review level.
The screening can be aided by using tables and rules. The fragility analyst is expected to develop such rules of thumb to assist in screening of components. Such screening tables can also be used in the plant walk-down to minimize inspection time by focusing the data collection (e.g., dimensions, number and size of bolts, etc.) efforts. A peer review team is expected to critically review these screening tables and approve their use. 6. Second Plant Walkdown
After the first plant walk-down, structural analysis is performed either by doing a full analysis or by scaling of design calculated responses (forces, moments, spectra, etc.). Some equipment response analyses might also have been conducted before this plant visit. Certain safety related components suspected to have generically low capacities might have been screened out using simplified analyses.
The second plant walk-down is primarily carried out by the fragility analyst, taking into account the results of the first walk-down, preliminary analysis and the systems analysis results obtained so far. This walk-down will emphasize actual physical study of plant components requiring detailed fragility analysis. Systems analysis input will be needed but in a supporting capacity.
The objectives of this walk-down are: • To obtain additional specific information (i.e., dimensions, number and size of anchor bolts, support details, estimate of weight, etc.) for evaluating the HCLPIF values of screened in components, and • To verify the systems models and collect any additional needed information. Plant (second) Walkdown Procedures By the second walk-down the fragility analyst has a good idea of the components that require margin evaluation. This second walk-down is used to determine whether additional components may be screened out using plant-specific data.
Before conducting the second physical walk-down of the plant, the fragility analyst should study the design details of each screened in component (e.g., design criteria used, details of supports and anchorage systems, qualification method) using stress reports, equipment qualification reports, and other sources. He should examine the photographs of the component taken during the first walk-down to assess the as-built conditions. He should also identify the potential failure modes of the component so that he can concentrate on those elements (e.g., anchorage) during the walk- down.
Since the event- and fault-trees are not finalized at this stage, the systems analyst can make preliminary recommendations that certain components be more thoroughly studied during the walk-down and that others may not require extensive analysis because of their function in the plant systems.
During the second plant walk-down, the fragility analyst will determine details of components; note the type and size of the anchorage and its arrangement around the base of the equipment; determine the size of piping and overhang of motor actuator, and estimate the actuator weight for motor-operated valves on small pipes. He will make simple analyses to judge, for example, valve item binding or the adequacy of anchorage for the earthquake review level. He will use the screening tables developed earlier to screen out additional components.
For the remaining components, detailed measurements (additional to those shown in design drawings) are taken such that their HCLPF can be estimated. In practice, this is the last chance for the analysts to inspect and obtain field information on the screened in components. 7. Conclusions:
At the conclusion of the plant walk-down(s), the following objectives will have been achieved:
• The plant system configuration is verified in order to proceed
with event tree and fault tree analyses.
• Systems interactions, other types of dependencies or plant
unique features are identified.
• The safety related components that are judged to generically
possess high capacities (i.e., larger than the earthquake review
level) have been verified to contain no weaknesses.
• Further analyses needed to establish the capacities of
remaining safety-related components are identified and
necessary field data are obtained.
• Information on components is obtained to assist in HCLIP1
(fragility) evaluation and peer review of the seismic margin
study. XA0 300585
Seismic Fragility Analyses
by: M. Kostov REGIONAL WORKSHOP ON EXTERNAL EVENTS PSA 6-10 NOVEMBER 2000, SOFIA
SEISMIC FRAGILITY ANALYSES (Case Study) Marin Kostov
1. Introduction
In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, B~E, determined by the relation:
f=- [d[,j3(x)]/dx]~(fi x) dx (I)
I-x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(f x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard B(x) and the fragility P(fI x).
The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined probabilistically and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency.
2. Basic Formulation of Fragility Curve Model
The fragility of a structures is defined as the conditional probability of failure at a given value of seismic response parameter as maximum acceleration, velocity displacement, spectral acceleration, effective acceleration Arias intensity, etc. Generally there are two ways of defining seismic fragilities, i.e. in terms of global ground motion parameter or in terms of local response parameter.
Most frequently the objective of the fragility evaluation is to estimate the peak ground motion acceleration value for which the seismic response of a structure (system, component) exceeds the capacity resulting in failure. The estimation of the ground acceleration value could be performed on the base of calculations or based on experience data (the later could be from real earthquakes or dynamic tests). Because there are many sources of variability the structure (component) fragility is expressed usually by family of curves. A probability value is assigned to each curve to reflect the uncertainty in the fragility estimation (fig.2. 1)
The first step in eneration fragility curve is a clear definition of what constitutes failure for the analyzed object. The failure definition may differ significantly depending of the goals of the analysis, e.g. failure could be any loss of function, strength, integrity, value, etc. One and the same failure may happen in different failure modes, each of them have to be clearly identified and addressed. A post office may fail for instance due to structural failure, failure in the electrical supply, failure of the road system, failure of the communication equipment, failure of the auxiliary facilities. etc. Another example of failure mode differentiation is the ductile or the brittle mode of failure. If there is clear definition for the possible failure modes, fragility has to be developed for the mode which is most likely to occur. otherwise fragilities have to be developed for each identified mode.
One simple but effective fragility model supposes that the entire family of curves representing a particular failure mode can be expressed by median ground acceleration Am and two random variables F R and u thus the ground acceleration capacity A is given by
*Associate Professor, Dr., Head of Department "Seismic Mechanics", CLSMEE, Bulgarian Academy of Sciences Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
---- 00 --
0.
0 I / ~~~~~~~~~~~~~0.05mn .95 n0,edf~
Fig.1. Seismic fragility family curves
A= AmSE R. U
3 ER and u are log-normally distributed with unit medians and standard deviations PR and jPu respectively. They represent the inherent randomness about the median and the uncertainty of the median value respectively. In some cases the composite variability c is used, defined by:
3 1 P3C=( fR 2±OU') '
The use of P~c and A,, provides a single best estimate fragility curve which does not explicitly separate randomness from uncertainty.
In estimating the fragility parameters it is convinient to use an intermediate random variable, called factor of safety, F. The factor of safety is defined as: F=(Actual seismic capacity)/(Actual response due to DE) DE is the design earthquake Further on the factor of safety can be expressed by:
F=FS.FO.FRS
Fs is called stress factor, representing the ratio of the ultimate strength to the stress, calculated for DE. FP3 is the inelastic energy absorption factor, which depends on the available ductility and reflects the ability of the structures to withstand seismic loads beyond yield without loss of function. FRS is the structural response factor that recognizes the in the design the structural response have been computed using specific (some times conservative) response parameters. The response factor is modeled as a product of factors influencing the response variability, e.g. spectral shape factor (representing the variability of the ground motion), damping factor (representing the variability of response due to difference of the actual damping and design damping), modeling factor accounting for the uncertainties due to modeling assumptions, mode combination factor, earthquake component combination factor, factor to reflect the reduction of the seismic motion with depth, factor to account for soil-structure interaction effects, etc.
The median and logarithmic standard deviation of the safety factor F are expressed as: mF=mFs.mFpmFRS and
3 PF ( W+2 p2 f RS2 )/2
2 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
The logarithmic standard deviation could be further divided into random variability and uncertainty. The median factor of safety multiplies the design ground acceleration to obtain the median ground capacity.
3. Case 1. Concrete Gravity Dam
The probability of seismically induced failure of large concrete dams is of special importance because of the potential flood due to the released water from the lake. The need of such assessment arises moreover in the case where the design values of existing large dams differ from the design values specified by new standards. The basic steps of the procedure are: assessment of the seismic hazard and uncertainties; statistical formulation of material properties and loading; assessment of statistics of the response; definition of the failure criteria; evaluation of the probability of failure. The case presented hereafter is on Antonivanovzi dam which is located in the sout-west part Bulgaria. The dam safety is not related to the nuclear facilities in Bulgaria and is presented only as an example.
3.1 Seismic hazard analysis of the dam site
The seismic hazard curves result from the application of probabilistic models of the site region defined on the basis of complex analyses including description of regional tectonic, review of historic seismicity, identification of seismic source zones, development of earthquake recurrence relationships. The models incorporate the following main characteriastics: there are no contemporary active faults which pass through the dam and the dam reservoir. Four potential foci zones have been identified in the near field (30km) zone around the dam site which might generate earthquakes with maximum magnitude from 5 to 7. The events with magnitude 7 are generated in a depth 10km to 20km. Earthquakes with magnitudes greater than 7 have occurred at a distance of about 60km from the site 6 and events with MŽ6 - over 40km. At shorter distances the events occurred are with Mmax< . In the region there are five epicentral zones i the territory of Bulgaria and one in the North part of Greece. The strongest seismic event is realized in the Marica zone on the territory of Bulgaria. The possible effects from Vrancea zone situated at a distance of more than 400km from the dam site are also studied. The ground motion attenuation relationships used for the models are based on the analysis of strong motion data records from earthquakes in the Balkan region countries, Italy, USA.
--Median
- Mean 1.5 *. perendes Median 10-2 ...... 15 and 85 1.0 ...... perceotiles
'X 1E'3 1E-2
31 ~~~~~~~~~~~~2.0 2.0
10.5 ., 1.5
0.0 0.5 1.0 10.2 1E-2 0.1 1 Peak acceleration() Period()
Fig. 3.1 Mean. Median 15 and 85 percentiles Fig. 3.2 Hazard Response spectra, 5% damping, Hazard curves anual probability of exceedance: a) 10-4; b)105
3 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
The following uncertainties in the mathematical model are considered: configuration of the seismic sources, uncertainty in spatial distribution of seismicity (for the Marica epicentral zone), uncertainty of focal depth, uncertainty in maximum expected magnitude, different alternatives of the acceleration attenuation law, and uncertainty of law dispersion. As a result for the Antonivanovtsi dam site 72 hazard curves are obtained.. In Figure 3.1 are presented the mean, median, 15th-percentile and 85th-percentile hazard curves obtained fom the calculated total hazard assuming lognormal distribution of the peak acceleration at a given annual probability of exceedance.
In a similar way the equal hazard response spectra for four hazard levels A, B, C and D, with annual probability of exceedance 0.01, 0.001, 0.0001, 0.00001, respectively, are obtained.. In Figure 3.2 are shown the equal hazard response spectra for levels C and D.
3.2 Statisticalformulation of material properties and loading
Strength and elastic properties
The materials of the dam structure are identified into 8 types: 5 of them are for the concrete of the dam body and 3 - for the rock foundation. For each type of material the mean value and the variation coefficient of the material characteristics (static and dynamic compression, tensile and shear strength, and the elastic module) are determined based on in situ and lab tests. Thermal loads The thermal loads are represented by sets of nodal temperature differences. For this 2D linear transient heat transfer analysis) is performed with input data obtained on the basis of statistical meteorological observations.
The values of the hydrostatic. hydrodynamic and filtration pressure are function of the water level in the lake. The maximum working water level of Antonivanovtsi dam is 535.8m and the minimum level is 505m. An uniform distribution is assumed for the water levels between 535.8m and 505m.
The seismic load is the most important for the seismic risk analysis. For each seismic hazard level (A, B, C and D) the seismic loading is presented by a set of acceleration response spectra and the corresponding acceleration time histories. Those spectra are generated on the base of the statistics of the equal hazard spectra obtained by the seismic hazard analysis. Each one of the generated spectra is used as a target spectrum for generation of acceleration time histories (three statistically independent generations representing three components - two horizontal HI and H2 and one vertical V). The maximum accelerations of the vertical components are obtained from the horizontal ones by scaling with random numbers with mean value of 0.5 and standard deviation of 0.3.
3.3. Assessment of the response statistics
Finite Element Model A 2 finite element model of the highest block of the dam structure is used in the analysis. A plane strain condition is assumed. The rock foundation and the concrete dam body3 are modelled. The model length is 450m, the height of the rock foundation is 206m and the total model height is 22
340m. Along the rock base boundaries the model is fixed. - The rock foundation is assumed massless in the analyses. Fig. 3.3 Location of sections
4 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
Four cross sections are investigated in details. Their location is shown in Figure 3.3. The sections are as follows: base joint - section 1; injection gallery - section 2; control gallery - section 3; crest zone - section 4.
Computational Procedure The deterministic analyses are carried out according to the computational procedure based on an advanced Monte Carlo method (Latin Hypercube Experimental Design, LI-CED) for simulation. The main steps of the computation are as follows: I. Preparation of input variable samples by Latin Hypercube Experimental Design procedure. 2. Computation of stresses due to static loads. 3. Computation of eigen values and modes of vibrations. 4. Evaluation of stresses due to seismic loads. 5.Stress superposition. 6. Computation of maximum stresses. 7. Statistics of the results. Applying the LHFCED procedure (Iman & Conover., 1981) sets of the input variables contributing to the response of the dam structure are prepared. In the case of the Antonivanovtsi dam for each variable a stratified sample of size 10 is prepared. The input parameters that are varied in this study are the strength and elastic properties of the dam structure and foundation rock; thermal loading; hydrostatic, hydrodynamic loading, filtration pressure - correlated with the water level in the reservoir; material damping; natural frequency of the structure vibrations;' seismic excitation.
The seismic loading is represented by 10 acceleration response spectra and 10 accelerograms for each seismic level A B C and D. The spectra are generated on the base of the mean equal hazard spectra obtained by the seismic hazard analysis. Lognormal distribution of the spectral values is assumed. The response spectra of the generated acceleration time histories are computed and statistically processed for each safety level. Their mean values and standard deviations are compared with the mean spectrum of equal hazard and the respective standard deviation. As an illustration a comparison for seismic level C is shown in Figure 3.4.
2.0 . . 2.0 ~~~~~~~~~~~~~~~~~~~~~~~2.0.. Damping 0.05 Damping 0.05 Damlping 0.05 mean Tean mean - '.5 ~Signa '.5 sigina- 1.5 slgm
,UU U
0.0 .10
Fic. 0.0 1 tO .0 01 t 1 Natwral period (s) Natural period (s) Natural period (s)
Fi.3.4 Equal hazard spectra and spectra of generated accelerograms, annual probability of exceedance 10-4, components: HI:;-12; V
The procedure is applied for each of the accepted four levels with annual probability of exceedance 10-2 (level A), 10-3 (level B), 10-4 (level C), and 10-5 (level D). The results for each safety level are processed separately. The response parameters studied in the probabilistic analysis are the normal Svy stress, the tangential Sxy stress and the tensile zone length.
5 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
Statistics of results For the statistical processing of the response a normal distribution of response parameters is assumed. In . Figure 3.5 is shown the cumulative distribution functions (CDF) of Syy in nodal point 73 (the corner of the base joint-upstream face) for safety level B. The stars in this figure present the computed values and the *0 line shows the theoretical normal CD3F. The computed0. values fit well the accepted normal CDF and allow to assess the statistical character of the investigated04 parameters. The mean values and the standard deviation of the generated response quantities are estimated for each0. nodal point of the investigoated sections. For illustration in Figure 3.6 are shown the mean value and the 0..b- -. .... standard deviation of Syy for section 2. The mean w a values and the respective standard deviation of the Fig. 3.5 Comulative distribution function tensile zone lengths of the four sections for all seismic SYY, NP 73, level B loading levels are computed as well.
10.00 5.00 ...... Level A ... Level A LeveiB -~-Level B ---Level C ---- Level C Level D ___Level D 0.00 3.00
0.0C 1.00
4 . .0 ~~~~~~~~0.201.4 06 08 I Relative length Ll1 13.7 m Relative length L=1 13.7 m
Fig. 3.6 Statistics of stresses, section 2: a) mean SYY, b) standart deviation SYY
Scenarios offailure due to seismic excitation To assess the probability of failure the following scenarios for failure are considered: 1. Exceedance of the concrete tensile strength. This is the most common case for failure of concrete dam structures. In the base joint (section 1) two additional cases are investigated. For this section the scenarios are: a) Probability for exceedance of the concrete tensile strength (above the foundation joint), b) Probability for exceedance of the rock base tensile strength (below the foundation joint), 2. Exceedance of the allowable length of tensile zone. This scenario is defined because of the requirements of the norms for design of concrete gravity dam. It is related to the possibility of failure due to large tensile stresses as well as to the control of pressure stresses. The mean value of the allowable tensile zone is accepted conservatively as follows: 10% of the length of base joint and 20% of the length of all other sections, and the variation coefficient is 20%/ Norms, Concrete and reinforced dams, 1986). 3. Probability of failure from simultaneous exceedance of the tensile strength and the allowable length of tensile zone. For the base joint two additional cases are investigated: a) Probability of failure from simultaneous exceedance of the concrete tensile strength and the allowable length of tensile zone, b) Probability of failure from simultaneous exceedance of the rock tension (below the base joint) and the allowable length of tensile zone, 4. Probability for loss of stability due to sliding in the concrete-rock contact. The moost important scenarios from engineering point of view are scenarios 3 and 4.
6 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
For each scenario two alternatives are considered. The first alternative is when the maximum stresses are used in the analysis. That implies full correlation of damages in adjacent nodes, i.e. if there is a damage in one node all other nodes would be also damaged. This assumption is very conservative. The second alternative is when averaged stresses are used in the analysis. This means the tensile stresses to be determined in a small zone, i.e. some correlation between the behaviour of adjacent elements to be accepted.
Conditionalprobability offailure due to seismic events The probability of failure expressed by the probability of exccedance of the tensile strength in each node of the FE mesh within the investigated section is computed under the assumptions that the stresses and the resistance (strength) are normally distributed. The conditional probability of failure is computed by the expression: Pf= FR(x)fL(x)dx (2) where FR(x) is the distribution function of the resistance and fL(x) is the density function of the seismic loading distribution.
3.4. Risk assessment
The total probability of failure is determined by integration of the seismic hazard curves together with the fragility curves. The LHCED procedure is applied for the integration.. The fragility curves are obtained from the computed discrete values of the conditional probabilities for each seismic level. For this purpose the function that passes through those discrete values is approximated with cumulative log- normal distribution functions. The approximation is done by the least square method. The procedure is performed for each seismic level and each section (for 50% and 85% confidence level). For example, in Figure 3.7 the curves are shown for scenario 1, first alternative (maximum stresses), section 1 (base joint). The solid line represents the curve of the mean values of the conditional probability, the stars are the discrete values. On the basis of the approximation function characteristics 10 fragility curves for each scenario and section are generated with mean value and standard deviation corresponding to the values obtained by the numerical experiment.. In Figure 3.8 are shown for scenario 3b the distributions of the total probability of failure in dependence on the acceleration. The figure illustrates also the way of integration. Each ordinate gives the contribution of the respective acceleration to the total probability. The total probability distribution shows that the seismic loading levels for which the analysis is performed are selected correctly. In section 1 accelerations higher than 0.4g have small contribution.
85% S~~~~~~~~~6OE-005
4.OE-005 -- Man~l
3.0E-005 Mal
~~~.2.0E-~~~~~~~~~~~~.E-0 iOE-001~ ~ ~ ~ ~ ~ ~ ~ 2t 0
"C 'C~~~~~~~~.c005
0.OE.000 0 . *'0l 00 da 08 Fic.3.7 ~~Amv(j) A g Fg37Approximation of the numerical generation Fig. 3.8 Distribution of the total probability Results for level A. B. C. D with log-normnal CDF7 of failure depending on Amax; scenario 3b,
The highest probability, of failure is in the base joint resulting from simultaneous exceedance of the rock tensile strength and the allowable tensile zone length. From all considered sections the most affected is the section 1 - base
7 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
joint and section 2 - injection gallery, n these section the total probability of failure for a period of 50 years is defined and the values are given in Table 3.1.
Table 3.1 Total probability of failure for 50 years Scenario Probability of non-exceedance 15% 50% 85% confidence confidence confidence level level level Failure from simultaneous exceedance of the tensile strength in 3.49E-03 7.69E-03 1.69E-02 the rock and the allowable length of tensile zone Failure from simultaneous exceedance of the concrete tensile 4. 1OE-04 L.OOE-03 2.45E-03 strength and the allowable length of tensile zone - section 2 ______
Sensitivity analysis The sensitivity of the results when the input parameters (strength characteristics of materials, variation of seismic hazard, of stresses, etc.) are varied and the influence of the size of parameter samples on the probability are investigated. Thus the probability of failure in the base joint can be reduced by decreasing the tensile stresses, by increasing the tensile strength in the rock or by decreasing the central seismic hazard values. These three factors can be used also in combination with the purpose of reducing the seismic risk. By 50% variation of each one of those factors the most considerable effect on the seismic risk reduction is achieved. By the reduction of the tensile stresses the probability of failure decreases 29 times. By reduction of the central seismic hazard values the risk reduces 6.27 times and by the increase of the tensile strength of the rock the risk reduces 1.79 times.
4. Case 2. Steel frame industrial building
The steel frame industrial building analysis presented is part of a probabilistic safety analyses of a nuclear power plant. The building is an existing one, it is not safety relevant, but is important for electricity production. The outcomes of the probabilistic analysis have been the bases for a consequent seismic upgrading project. The probability of failure before and after upgrading respectively is used as measure for the upgrade effectiveness.
4. 1. Structure description
The structure is a 2 bay steel frame. The first span is 45m and with a height of 38m and the second span is 12 m with a height 42m. In longitudinal direction the steel columns are placed in 3 rows (A, B, C) and 12 transversal axes (2m between two adjacent axes). They are connected by longitudinal hinged beams. Diagonal bracing provides the longitudinal stiffness. The foundations are single footings. The roof structure is steel truss.
Next to the steel building there is a reinforced concrete building. It is also frame structure (one bay, 12 m span). Because of the small gap between the two buildings as well as due to the common foundation used in row C the two buildings are analyzed together in order to account for the structure to structure interaction.
4.2. Modeling and Analyses
A comprehensive 3-dimensional FE model has been developed for all essential structural elements. The soil-structure interaction is accounted by equivalent spring and dashpots, connected to each footing. The seismic excitation is defined by a maximum acceleration for 0-' annual probability of ecceedance and a site specific broad band design spectrum. Modal analysis and time history integration are used for the dynamic analyses. Some of the important mode shapes are presented in fig. 4.1I. Static analyses for all design load combinations are performed also. Most unfavorable loading condition is found out and capacity estimation for all bearing elements is performed respectively. The results of the analyses are showing that:
8 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
1. All columns have sufficient bearing capacity and can withstand the design seismic motion. 2. The column-roof connection has insufficient capacity and bad detailing. Failure of the roof structure is a possible failure scenario. 3. The longitudinal beams and X-brancings are very slender and failure due to overloading is possible. 4. The safety of the foundation for sliding and overturning is guaranteed.
Fig. 4.1 Important mode shapes
Those conclusions mean that there are at least 4 scenarios with considerable influence on the overall structural safety that have to be analyzed. For each scenario a fragility curve is developed following the methodology described in section 2 above.
In order to improve the safety an upgrading concept has been proposed. It consist of improving of the column-roof connection by using a better detailing and reducing of the slenderness of the longitudinal beams and x-bracing by
9 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia increasing the sections of the members or additional support points (shortening of buckling length). Fragility analyses are performed also for the upgraded structure and results are compared with the existing structure.
4.3. Fragility curves, Failureprobability assessment
As explained in the beginning the fragility curve is assumed log-normal. The mean failure acceleration is determined by scaling the design acceleration by the factor of safety. As also explained the factor of safety can be represented as product of different partial safety factors. For the present analysis the following partial safety factors are used:
F=Fs.FO.FRS
Where FRs=Fm*Fc FM is safety factor representing the conservatism of the modeling. Fc is safety factor representing the conservatism of the member force combination procedure.
Similarly the variations O3Rand 13u are determined for each partial safety factor. As an example the values for the determina ion of the fragility curve for roof failure are presented below:
Table 4. 1. Partial safety factors and corresponding uncertainties Safety Mean P3 R 13U Factor Fs 1.050 0.100 0.150 FP3 1.400 0.050 0.100 F, 1.100 0.050 0.050 FC 1.100 0.050 0.050 F 1.791 0.132 0.193
The mean maximal acceleration that will cause failure is determined as Am= .1*1.791=0.179g. The corresponding fragility curve is shown in fig. 4.2.
1.00......
A=0. 179 U'0. 193 0.80
0.60
.40
.0
.0
0. ~~ ~ ~ .0
0.0900 0.20 ...... 0.40 F~ragility Parameter (PGA g)
Fig. 4.2 Fragility curves
10 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
A summary for all considered failure modes before and after upgrading are presented in table 4.2.
Table 4.2. Fragility curve characteristics
______BeforeUpgrading AfterUpgrading 3 Failure Mode Am PR - ____ Am PR - 1U
Roof failure -0.179 0.132 0.193 -0.356 0.132 0.193
Bracing failure- row A 0.125 -0.132 0.193 0.404 0.132 -0.226
Bracing failure- row B 0.147 0.132 0.193 -0.390 0.132 -0.226
Bracing failure- row C -0.154 0.132 0.193 0.396 -0.132 -0.226
For the site of that building there are seismic hazard curves for maximal acceleration developed. The methodology uses is the same as described in case . The fragility curves and the hazard curve are integrated together and the annual probability of failure is determined for each failure scenario. The results are presented in table 4.3 for 85% confidence limit.
Table 4.3. Probability of failure (85% confidence level)
______Before Upgrading - After Upgrading Failure Mode Probability of failure Probability of Failure Roof failure 1.242E-3 2.356E-5 Bracing failure- row A 7.872E-3 1.51 8E-5 Bracing, failure- row B 3.548E-3 1.886E-5 Bracing failure- row C 2.836E-3 1.71 7E-5
5. Case 3. Soil Liquefaction
There are big, number of liquefaction analysis methods. Most of them are based on statistical observations. Unfortunately there are rear confidence limits and uncertainties reported for the respective method. In the presented case analysis a site was initially assessed deterministically using several methods. Some of the methods were assessing the site as potentially liquefable, some were saying that the site is safe. The characteristics of the site are presented in table 5.1I. For example the simplified Seed method (1971) is saying that the site will liquefy. The Seed method based on SPT values (1 979) is saying that the site is almost on the boundary, i.e the safety factor is generally more than 1, but is not definitely safe. The Seed method (1979) based on SPT and experimentally determined shear stress ratios (3 axial tests) is saying that the site is safe. The pore pressure generation analysis (Seed 1976) is also saying that the site is safe.
Table 5.1 Site characteristics Layer FMaterial FThickness Weight Relative SPT (N) Permeabilit No __ _dniy dniyy
______F loes m t/m3 % N cm/s 1 Sandy, lo 8.6-10.5 1.9-2.1 2 Loes-clay 1.7-3.5 1.9-2.1 - - 3 Fine sand 0.5-3.0 1.6-1.7 53 9/21 4* 0-4
1 *)There is an ovebudof24 0 ka onthat layer. ~~~ 492 *0.
For that site probabilistic analysis has been performed. The analysis scheme is similar to the one presented in case - dam analysis. There are only different methods used to assess the safety, i.e. for the dam assessment multiple stress analyses have been performed, for the liquefaction assessment multiple analyses based on the Seed (1979) method are performed. Regional Workshop on External Events P~SA, 6-10 Novemiber 2000, Sofia
The safety factor for liquefaction is expressed according to Seed (1979) as:
FS=SRJL
The SR is the cyclic stress ratio for equivalent number of cycles Ne (soil resistance) and L is the stress ratio caused by the equivalent number of cycles by an earthquake (seismic load). The experimentally determined cyclic stress ratio to generate liquefaction for the site is presented in fig. 5. 1.
D-06-
a52 -
022 -
0.36-~~ ~ ~ ~~~~~*
V0~~ ~ ~ ~~~~~~~
012-
Fig. 5.1 Experimentally determined cyclic stress ratio
Because of the importance of the equivalent number of cycles of the earthquake excitations an additional simplified statistical analyses was performed to assess the site-specific excitation cycles. The site is located in the northern part of Bulgaria. The seismic hazard is governed by two type of seismic sources: local shallow sources with maximum magnitude up to 5.5 and fare field intermediate depth Vrancea source with maximum magnitude up to 7.5. The shallow source excitations were represented by 43 three component accelerograms recorded in Bulgaria, Italy, Turkey and Japan. The Vrancea excitations are represented by 9 three component accelerograms recorded in Bulgaria and Rumania during the 1977, 1986 and 1990 Vrancea earthquakes. The equivalent number of cycles is determined according to he Seed's procedure (1975). The results are summarized below:
Local sources Fare field sources Equivalent number of cycles Equivalent number of cycles Mean value: 5.778 Mean value: 10.34 Standard deviation 3.006 Standard deviation 6.08 Minimum value: 1.57 Minimum value: 2.82 Maximum value: 16.63 Maximum value: 30.62
The multiple deterministic analysis is based on the LH-CED procedure. There are 10 samples generated and analyzed for 3 seismic levels (annual probability of exceedance 10-, 0-4 and 10-i). The generation is done on the following key parameters (table 5.2):
For each set of variables the Seed's method for liquefaction analysis is applied deterministically for three levels of seismic excitation. The seismic levels correspond to annual probability of exceedance IO-, l0C and 0-'. The seismic hazard curve for maximum acceleration is the one presented for case 2 (fig. 5.2). In order to achieve better accuracy the layer 3 and 4 (where liquefaction is expected) are further subdivided in thinner layers (usually 0.5m thick). For each layer the central factor of safety and the corresponding standard deviation is computed for each seismic level. The conditional probability of failure is computed assuming that the resistance and loading are log-normally distributed. The computed conditional probabilities are presented in table 5.3.
12 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
Table 5.2. LHCED generated parameters Parameter /Generation No 1 2 3 14 5 6 7 8 9 10 Thickness of layer 1 8.98 10.03 9.76 8.69 10.43 9.39 9.1 9.15 9.66 9.31 Thickness of layer 2 2.89 2.68 3.12 2.96 3.25 3.09 3.15 3.31 3.03 3.60 Thickness of layer 3 1.34 3.00 2.36 1.77 1.93 1.65 0.90 2.29 1.57 1.30 Thickness of laver4 2.02 3.45 1.66 2.38 2.51 2.78 1.97 2.94 5.05 3.29 Weight density of layer 1 2.02 2.07 2.06 2.06 2.02 2.04 2.00 2.05 2.04 2.03 Weight density of layer 2 2.09 2.12 2.11 2.11 2.09 2.10 2.08 2.10 2.10 2.10 Weight density of layer 3 1.64 1.62 1.64 1.66 1.63 1.65 1.67 1.69 1.66 1.65 Weight density of layer 4 1.60 1.67 1.65 1.71 1.67 1.68 1.59 1.63 1.66 1.63 Underground water level` 28.95 28.60 29.81 29.14 29.21 29.34 29.00 28.82 29.57 30.02 Equivalent cycles (10`') 2.44 3.58 3.25 2.18 4.10 2.85 2.98 2.60 3.15 2.77 Equivalent cycles (10") 7.48 5.59 9.99 8.25 11.69 9.68 10.38 12.52 8.95 17.3 Equivalent cycles (10`) 10.75 16.79 8.67 12.46 14.63 23.24 20.49 19.15 13.40 27.63 Max. acceleration (1 0`) 0.10 0.12 0.10 0.11 0.12 0.14 0.13 0.13 0.11 0.15 Max. acceleration (10-4) 0.15 0.19 0.14 0.16 0.17 0.21 0.20 0.23 0.24 0.17 Max. acceleration (10--) 0.20 0.27 0.22 0.18 0.26 0.33 0.28 0.35 0.31 0.25 Relative density of layer 3 42.02154.22 39.05 46.59 49.92 60.22 57.04 55.62 70.85 47.69 Relative density of layer 4 32.92 44.00 36.37 28.92 41.80 54.77 45.47 58.20 50.45 40.32 Experimental 3-axial sill 0.39 r0.45 0.49 0.36 0.51 0.42 0.43 0.40 0.45 0.41 resistance: eq. Y=Ax B
B =-0. 16: A = ...... ______
*)Underground water level is relative to sea level, the site level is +35.00m.
Table 5.3 Conditional probability of failure due to liquefaction of layers (sub-layers) 3 and 4 Laver No Level 1 (10`) Level 2 (10`) Level3 1) 3.1 1.78E-10 5.07E-02 5.96E-0l1 3.2 2.24E-10 5.33E-02 6.05E-01 3.3 2.06E-10 5. 1E-02 6.11IE-01 3.4 2.34E-10 5.7 1E-02 6.17E-0 1 4.1 2.62E-10 5.8 1E-02 6.2IE-01 4.2 2.59E-10 5.69E02 6.23E-01
Fig. 5.2 Hazard curves, mean, median, 15 and 85 percentiles
10-IO ------
C.) ~ ~ r------.r------r----
0 ~ ~ ~ ~ ~ ~ - L- - - - -L------L - - - -
0*
------ed ------
0. 00 0_.16 0.32 0.46 V064 060- -09 6 Peak acceleration () 1 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
CE-01 - __ 8.0E-001 C5M ~~~~~85j
6.oE-001 6CE-001B
4.OE-001 4.CE-O01
L.OE-001 2.0E-001
O.CE+000 0.0E+000 1...... 0.00 0.8 0.18 0.24 0.32 0.00 0.08 C 16 0.24 0.32 Amex () A-ax(g)
I .OE+00050 i.0c+000 0
/ I A6.0E-CO1 G.OE-001
2 .OE- 0014.E0
2.OE-DO1 ~~~~~~~~~~~~~~~2E-00
0.0E+000 C.0E+000 0.00 0.08 0.16 024 0.32 0.00 0.08 0.16 0.24 0.32 Amax () Am ()
1 OE+000 1.CE+000 50. 50.
8.CE-001 B.CE-001
6. I..0
2 .OE-001 .BE-C 1
4E00 0.t 4.0E'32001 0
Amex () A- () WU 7 rui
Fig. 5.3 Fragility curve generated by interpolation and extrapolation of the results for each layer
14 Regional Workshop on External Events PSA, 6-10 November 2000, Sofia
Based on those conditional probabilities a fragility curve is generated by interpolation and extrapolation of the results for each layer (sub-layer), fig. 5.3. The fragility than is convoluted by the hazard curve and the probability of failure due to liquefaction for each layer (sub-layer) is determined.
Table 5.4. Probability of failure due to liquefaction of layers 3 and 4
Layer No Probability 3.1 6.71IE-05 3.2 6.89E-05 3.3 6.98E-05 3.4 7.07E-05 4.1 7.18E-05 4.2 7.19E-05
6. Conclusions
The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinarx structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper.
After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The first group comprises the methods for monitoring and control. Generally their aim is to collect additional information and based on that to improve assessments. The second group of measures is the engineering. The engineering includes the repair, strengthening and upgrading of the investigated systems. In all cases the risk assessment is a power tool for decision taking.
7. References
Ang. A. & Tang, W. 1984. Probability concepts in engineering planning and design. Vol. 2. Decision, risk and reliability. John Wiley & Sons. Bolotin, V. 1979. Random vibrations of elastic systems. Moscow (in Russian). Borges. J.F. & Kastaneta, M. 1971. Structural safety. Lisbon. LNEC. Cornell. C. 1968. Engineering seismic risk analysis. BSSA. Vol.5 Franzini. ., McCann. M. & Shah, H. 1984. Application of probabilistic risk analysis to the safety of dams. 8WCEE. Vol. 7. San Francisco. Iman. R.L. & Conover, W.J. 1981. Latin Hypercube sampling program. Short course on sensitivity analysis techniques. Texas Technical University. Kostov. M. et al. 1994. Report: Probabilistic assessment of the seismic safety and risk of Chaira dam structure. Vol.2..NEK.Sofia. Kostov. M. et al. 1995. Seismic PSA of Kozloduy 3 NPP, Report, Vol1 to.3. Risk Engineering, Sofia. Kostov. M. et al. 1997. Seismic PSA of Kozloduy 2 Report: Vol.3. Risk Engineering, Sofia. Lomnnitz, C. & Rosenblueth. E. 1976. Seismic risk and engineering decisions. Elsevier. Murzeweski. . 1974. Sicherheit der baukonstruktionen. VEB f. Bauwesen. Berlin. NISA 1I. 1992. Computer code. User manuals. Engineering Mechanics Research Corporation. Michigan. Norms for design of buildings and structures in earthquake regions. 1987. Sofia (in Bulgarian). Norms (SNiP 2.06.85). 1986. Concrete and reinforced concrete dams. Moscow (in Russian) NUREG/CR-2300. 1983. A guide for the performance of the probabilistic risk assessment for nuclear power plants.
15 XA0300586
Recent Development in the External Hazard Risk Assessment in Ukraine by: Ukrainian Representatives
1 Ukrainian State Scientific and Technical Centre on Nuclear and Radiation Safety jointly with Ukrainian Ministry of Energy
RECENT DEVELOPMENTS IN THE EXTERNAL HAZARD RISK ASSESSMENT IN UKRAINE
Presented at the IAEA Workshop on Modeling External Hazards in PSA
Sofia, Bulgaria, November 2000 Recent Developments in the External Hazard Risk Assessment in Ukraine
PRESENTATION OUTLINE
INTRODUCTION: PSA AS PART OF SAFETY ANALYSIS REPORTS FOR UKRAINIAN WVERS
* CURRENT STATUS WITH PSA DEVELOPMENT
(INTERNAL) AND EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE NPP UNIT 1 General Scope and Approach
Selected Methodological Aspects
Basic Results
ON-GOING AND EXPECTED FUTURE ACTIVITIES
IAEA Regional Workshop on Modeling External Hazards in PSA2 Recent Developments in the External Hazard Risk Assessment in Ukraine
INTRODUCTION: PSA AS PART OF SAFETY ANALYSIS REPORTS FOR UKRAINIAN VVERS
• Ukrainian legislation and international commitments prescribe SAR development for all operating and future nuclear power plants
• SARs for operating Units shall include
- Main (Summary) Report
- Five Major Appendices
• Safety Analysis Supplement ("DMAB")
• Design Basis Accident Analysis
• Beyond Design basis Accident Analysis
• Probabilistic Safety Assessment
• Technical Substantiation of Safety (TOB) IAEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
INTRODUCTION: PSA AS PART OF SAFETY ANALYSIS REPORTS FOR UKRAINIAN WVERS (CONT'D) Regulatory Requirements to PSA contents: Criteria: * Core Damage Frequency * Large Radioactive Release Frequency Radioactive release sources: • Nuclear core * Spent fuel storage pool • Other Initiating events: * Internal lEs • Internal Hazards • External Hazards Operation State: • Full Power Operation * Low Power Operation * Shutdown
IAEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
INTRODUCTION: PSA AS PART OF SAFETY ANALYSIS REPORTS FOR UKRAINIAN WVERS (CONT'D)
Internal Hazards to be Considered External Hazards
- internal fires - seismic impacts
- internal floods - external fires - etc. - external floods
- extreme ambient temperatures
- aircraft crashes
- etc. Assessment of the internal/external hazard contribution to the plant risk is to be predicated on the results of deterministic analyses included in other SAR parts (DMVAB, TOB, DBA analysis)
IAEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
CURRENT STATUS WITH PSA DEVELOPMENT
Three pilot Projects: • WWER-440 Model 213: Rivne NIPP Unit 1 • WWER- 1000 Model 320: Zaporizhzhya NPP Unit 5 • WWER-1000 Model 302: South Ukraine NPP Unit 1
Results of the pilot projects will be adapted to other reactors of similar type: • Rivne Unit 2 (WWER-440 Model 213) • South Ukraine Unit 2 (WWER- 1000 Model 302) • Zaporizhzhya Units 1-4, 6; Rivne Unit 3, South Ukraine Unit 3, Khmelnitsky Unit 1 (WWER-1000 Model 320)
Two WWER- 1 000 Model 320 reactors are under construction completion - Khmelnitsky-2 and Rivne-4. Licensing and SAR development for these Units will be performed under separate requirements.
IAEA Regional Workshop on Modeling External Hazards in PSA 6 Recent Developments in the External Hazard Risk Assessment in Ukraine
CURRENT STATUS WITH PSA DEVELOPMENT (CONT'D)
*Similar two-phase approach is applied in all pilot projects: - to start from Level 1 PSA for internal IE;
- to complement Level 1 work with internal/external hazard analyses and Level 2 PSA *Level 1 PSA for internal initiating events has been completed for South Ukraine NPP Unit 1 and submitted to Nuclear Regulatory Administration. Regulatory peer review in SSTC NRS is on-going • Zaporizhzhya NPP has completed Level 1 PSA for internal initiating events for Unit 5. Final report will be submitted to the NRA by the end of 2000
• Rivne NPP1 Unit 1: Level 1 PSA for internal Es to be completed by the end of 2000 • Khmelnitsky NP3P started adaptation of the results of the pilot project for its WWER- 1000 Model 320 Unit 1
IAEA Regional Workshop on Modeling External Hazards in PSA 7 Recent Developments in the External Hazard Risk Assessment in Ukraine
CURRENT STATUS WITH PSA DEVELOPMENT (CONT'D) Internal and External Hazard PSA a In all three pilot projects evaluation of the internal and external hazard risks is to be conducted in two phases
Phase : * data collection and analysis; * screening of not important hazards, plant locations and SSCs; • identification and assessment of plant vulnerabilities; • plan for further detailed analysis Phase El: to proceed with detailed analysis of the internal/external hazards associated with major vulnerabilities to the plant safety, and evaluate hazard contribution to the core damage frequency. a Phase * is completed for South Ukraine Unit 1 * is started for Rivne Unit 1 * is scheduled to start by the end of the year for Zaporizhzhya Unit 5
[AEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
(INTERNAL) AND EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE UNIT 1 Internal Hazard analysis: internal fires internal floods Data collection and vulnerability analysis study involved: - extensive search and collection of the generic and plant-specific data;
- plant walkdowns intended to identify vulnerabilities and verify validity of the collected data;
- screening of the not vulnerable plant locations, fires zones and flood areas;
- qualitative and limited quantitative vulnerability assessment Main results: - collection of available data and identification of additional data needs;
- qualitative and quantitative screening;
- plan for detailed analyses Problems: - lack of exact information on the power and I&C cable tracing;
- lack of information for estimate of the internal flood frequency
IAEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D)
General Scope and Methodology a Phase ivolved collection of data and vulnerability analysis for both man-made (MMH) and natural phenomena hazards (NPH) m Works started with extensive list of external hazards, predicated on NUREG/CR-5042 "Evaluation of External Hazards to Nuclear Power Plants in the United States" and NUREG-1407 "Procedural and Submitted Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities"
IAEA Regional Workshop on Modeling External Hazards in PSA 10 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D) General Scope and Methodology (cont'd) After initial data collection and initial screening the following hazards have remained in the list MMH NPH - aircraft crashes - intense precipitation - ground transportation accidents * extreme rain * explosions * hail • releases of chemicals or toxic substances * severe snowfalls - industrial or military facility accidents - extreme winds and tornado - pipeline accidents * extreme winds * tornado * squalls - extreme temperature * high summer temperatures * low winter temperatures - ice cover - meteorite strikes - seismic activity IAEA Regional Workshop on Modeling External Hazards in PSA Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D) General Scope and Methodology (cont'd) * Basic steps:
- Compiling of initial list of hazards - Collection and evaluation of the plant- and site-specific data, as well as applicable generic data necessary to support subsequent quantitative analysis • plant characteristics • data describing the external hazards
- Plant walkdowns, initial identification of vulnerabilities
- Qualitative or quantitative vulnerability analysis
- Development of the plan for detailed external event risk evaluation
* Screening at each step
IAEA Regional Workshop on Modeling External Hazards in PSA 12 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D)
General Scope and Methodology (cont'd)
The following screening criteria were applied
1. the frequency of an end state initiated by the external hazard is more than a factor of 100 less than the frequency for the same end state quantified in the Level 1 PSA of internal initiators; or
2. the initiating event frequency is below 10O9 per year plus screening based on the Screening Distance Value approach for selection of the potentially hazardous objects, as proposed in the IAEA Safety Guide No. 50-SG-S5 (External Man-induced Events in Relation to Nuclear Power Plant Siting)
IAEA Regional Workshop on Modeling External Hazards in PSA 3 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT' D) Selected Methodological Aspects Aircraft Crashes
• Known data: - Air traffic in the SUNPP vicinity for different aircraft classes - Location of the nearest airports - Aircraft performances - Site characteristics (in fact - containment geometry and structural properties) - Crash frequencies (generic) for aircraft of a given category
• Crashes due to take-off/landing operations were screened out by the SDV criterion (Noperations< 1 000d2 IAEA SG)
• Annual aircraft crash frequency is obtained as F=1 Nj,k Xj~k f (Rk, hk KAj j,k (k - denotes corridor and j - denotes aircraft type) k= 1.74 x 1 0` per km (IAEA Workshop, Moskow, 1995)
IAEA Regional Workshop on Modeling External Hazards in PSA 14 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT' D) Selected Methodological Aspects (cont'd) Aircraft Crashes (cont'd)
• Assessment is focused on containment
• Simple estimate of the potential for containment damage due to aircraft crash, as proposed in DOE-STD-3014-96
td 1. 2-~ MV_2 < > 11 - 1.2 m
If analysis for containment shows that neither of the screening criteria is satisfied, we do have plant vulnerability, and detailed analysis is necessary in Phase 2.
IAEA Regional Workshop on Modeling External Hazards in PSA 15 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT' D) Selected Methodological Aspects (cont'd) Tornado
• South Ukraine plant is located in the area of the most frequent for Ukraine repetition of tornado-dangerous situations (one occurrence in 5 years in the area)
• Estimated annual probability of tornado passing over the SUNPP site is -3x1 Q6 per year
• Potential vulnerabilities:
- loss of house loads power supply - loss of circulation water - loss of service water, and - loss of power supply combined with loss of ESWS and NESWS (-2x1 0-)
IAEA Regional Workshop on Modeling External Hazards in PSA 16 lE (TI) Reactivity Emergency Power Condition: Sccondary Condition: Condition: Secondary Secondary Reactivity Control Primary Hleat Loss of control Supply "Powcr Hleat Removal `NO Hligh "Power Pressure Control Pressure Control (Boron Injection) Removal Offsite Restorahic Temperature Restorable within Power within 1 Ii' Environment' 8 Ii, Supply AZ-I1 DG's SG feeding by SDV operation in SDV operation in TJ13/33D01 or LPIS operation in FW pumps cooldown mode. P2=const mode. (TiJ l0130DO01 and RIHIR mode Start no later then Not stuck in open EGE) or I ,5ti after IEF posit ion ('1K and TID(TM ))
Automati - Auitomaticall y 1/3 114 3RZU-A (44RU-K or 4/4 1/3 or 1/3 or (1/3 1/3, Operator cally 113 Operator Operator FASIV) and and 1/3012)) 4/4BRU-A Operator Vt7 Al S Q2El Q' E3 E2 1B2133 F
No
High Temperature
More then 8 h
SBO e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ______
S f 0 ______~~~~~~~~~~~~~~~~~~~~ -0-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Figure 5. Event tree for Initiating Event Group 1'- 1. Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D)
Basic Results
Man-Made Hazards
- Explosions at the railroad (3 km distance from the site) do not impose hazard to the plant. Explosions at the motor road (700 m from the site) potentially can be dangerous for the plant in specific cases;
- Industrial or Military facility accidents have been screened out;
- Pipeline accidents do not impose danger for the plant (as has been determined from estimate of the consequences of explosion at the nearest high pressure gas pipeline - 8 km/i1000mm diameter)
IAEA Regional Workshop on Modeling External Hazards in PSA 18 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT' D)
Basic Results
Man-Made Hazards (cont'd)
- NPP is vulnerable to aircraft crashes, although the contribution of the aircraft impact to the core damage frequency will be relatively small
- There is potential threat from chemicals/toxic gas releases due to ground transportation accidents (both railroad and motor road)
Additional data are needed for estimate of the probability of the dangerous releases of the toxic substances in the SUNPIP vicinity due to ground transportation accidents.
IAEA Regional Workshop on Modeling External Hazards in PSA 19 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONT'D) Basic Results Natural Phenomena Hazards • The following hazards have been screened out - Extreme rain Qualitative analysis involved assessment of potential for water level increase in the river, in the cooling pond, and for direct water accumulation at the site. Neither of these impacts caused significant vulnerability for the plant - Big hail and severe snowfall - Low winter temperatures - Meteorite strikes • Extreme winds and squalls could be analyzed together with tornado impacts
IAEA Regional Workshop on Modeling External Hazards in PSA 20 Recent Developments in the External Hazard Risk Assessment in Ukraine
EXTERNAL HAZARD RISK ASSESSMENT FOR SOUTH UKRAINE (CONTYD)
Basic Results (cont'd)
Natural Phenomena Hazards (cont'd) • The following hazards require further analyses
- Seismic hazard
- Tornado
- Ice cover Potential for damage of the power transmission lines and further progression into plant transients (loss of power or reactor scram) - High summer temperatures (overheating of the cooling pond)
• The two latter hazards are deemed to be less significant and potentially could be screened out after some additional analysis
IAEA Regional Workshop on Modeling External Hazards in PSA 21 Recent Developments in the External Hazard Risk Assessment in Ukraine
ON-GOING AND EXPECTED FUTURE ACTIVITIES South Ukraine m To conduct Phase 2 - detailed evaluation of the external hazard contribution to the core damage frequency
- Aircraft Crashes and Tornado There is no in significant need in additional data, it is possible to proceed with determining E frequencies, accident sequence modeling and quantif ication
- Chemical Releases under Ground Transportation Accidents * To collect or develop data on intensity of dangerous cargo transportation at the SUNPP3 vicinity (both at railroad and motor road) * To acquire applicable statistics of such accidents * To evaluate CDF
- Collection and evaluation of additional information on ice cover and high summer temperature hazards, and conclusion on the necessity to consider these hazards in risk evaluation
IAEA Regional Workshop on Modeling External Hazards in PSA 22 Recent Developments in the External Hazard Risk Assessment in Ukraine
ON-GOING AND EXPECTED FUTURE ACTIVITIES (CONT'D)
South Ukraine (cont'd)
• The major subject of concern for Phase 2 - earthquakes • No detailed analysis in Phase 1 due to limited time resources and lack of necessary data • No possibility for conducting comprehensive seismic risk assessment in Phase 2 - too expensive • Simplified approach need to be elaborated and applied, which should incorporate the known procedures of seismic margin calculations and simplified fragility assessment development of a method for simplified seismic risk assessment seismic screening and walkdown simplified seismic fragility assessment quantification of seismic risk
• Seismic characteristics have been determined for the Ukrainian NPP sites in the recent study of LLNL--. The report (currently as a Draft) will be made available for Ukrainian analysts by ANL, and could be applied for detailed seismic analyses
IAEA Regional Workshop on Modeling External Hazards in PSA 23 Probabilistic Seismic Hazard Characterization and Design Parameters for the Sites of the Nuclear Power Plants of Ukraine
Prepared by J. B. Savy W. Foxall Hazards Mitigation Center Lawrence Livermore National Laboratory
Prepared for Argonne National Laboratory Recent Developments in the External Hazard Risk Assessment in Ukraine
ON-GOING AND EXPECTED FUTURE ACTIVITIES (CONT'D3)
Rivne Nuclear Power Plant • Phase analyses are initiated for both internal (fires and floods) and external hazards
- The general scope and methodology are/will be essentially the same
- There are site-specific differences that might impact the list of external hazards to be considered (e.g., external fires are to be addressed)
- LLNL- report is available early in the project and more emphasis can be put on the seismic hazard evaluation
• There is more background information for analysis
- Results of TACIS projects on Rivne Units 1,2 and 3 safety evaluation;
- IAEA documents on Bohunice safety analysis
- Some DMAB Chapters are already available Limited internal flood analysis has been conducted in 1994 by Kiev Energoproject
IAEA Regional Workshop on Modeling External Hazards in PSA 25 Recent Developments in the External Hazard Risk Assessment in Ukraine
ON-GOING AND EXPECTED FUTURE ACTIVITIES (CONT'D)
Zaporizhzhya Nuclear Power Plant • Phase is to start by the end of 2000
- The general scope and methodology are/will be essentially the same - There are site-specific differences that might impact the list of external hazards to be considered - LLNL- report is available early in the project and more emphasis can be put on the seismic hazard evaluation • There will be more background information for analysis - On-going DOE Fire Safety Shutdown study will provide a detailed information for internal fire risk assessment, which will be useful as well for internal flood analysis - Relatively recent information characterizing the site neighborhood can be obtained from safety analyses for ZNPP? Dry Spent Fuel Storage Facility
- Some DMAB Chapters are already available
IAEA Regional Workshop on Modeling External Hazards in PSA 26 XA0 300587
External Hazards in PSA Program for VVER 440/213
by: M Patrik IAEA Workshop on Modeling of External Hazards in PSA No~emb-. 6 /0. 2000 S~fa, Bulgana
External Hazards in PSA Program for VVER 440/213
Milan PATIRIK
Nuclear Research Institute Rez Department of Reliability and Risk Assessment Czech Republic
IAEA Workshop on Modeling of External Hazards in PSA N-mb 0 10, 2000,S~,fia, Bulgania
External Hazards in PSA Programn Content
* Living PSA Program for Dukovany NIPP - concept, goals and objectives * Full scope PSA - risk informed regulation
* External Hazards in PSA for Dukovany NPP - objectives, current status * Seismic PSA - current status [AEA Workshop on Modeling of External Hazards in PSA No b 6 10. 000. Sofia. Bulgaria
PSA for NPP Dukovany Background
*NPP Dukovany - VVER 440/213
*A basic PSA study - first step of typical PSA program - completed in 1995 for internal initiators and full power operation
IAEA Workshop on Modeling of External Hazards in PSA N-~e~her 6 - 10. 2000 Sfia. B&garia
PSA for NPP Dukovany Background
*Since that milestone aLiving PSA Program has come into force
2 IAEA Workshop on Modeling of External Hazards in PSA J Norember 6 10.2000, Sofia, Bulgaria
Living PSA for NPP Dukovany
main goal
- to use PSA models and outputs as a support decision tool in a Risk management process
IAEA Workshop on Modeling of External Hazards in PSA V N-me,,,c 6 10.2000. Sofia Bgal-i
Living PSA Programme Concept
*all PSA related activities we have taken under the umbrella called Living PSA
- Support of Risk management (executive task) -Data collection and Information Transfer (data support task)
- Maintenance and Improvement of PSA models (maintenance task)
3 IAEA Workshop on Modeling of External Hazards in PSA No~embe, 60 10,2000 Sofi., Bulg.,ia
Living PSA Programme Concept maintenance task
- updating of models and documentation - PSA scope extension
IAEA Workshop on Modeling of External Hazards in PSA No~'e,Te, 6 10,2000. Sofi0.B.9 i
Living PSA Model Development
skPApmde
1Jpdatin~~ S~fr~xtension [AEA Workshop on Modeling of External Hazards in PSA ~~y~ N-.ivbe, 6- 0. 2000,Sofia, 8ulgaria siutvs
Current Status of Living PSA Dukovany NPP?
Full Power Living PSA - internal initiators including fires and floods - reflects all implemented plant modifications up to June 2000 *Low Power and Shutdown States - internal initiators including fires and floods, heavy load drops - reflects all implemented plant modifications up to December 1998
IAEA Workshop on Modeling of External Hazards in PSA N-eouber 6 - 10.2000. Sofia, Bulgaria
Dukovany CDF Level Development Full power operation
Core damage frequency I/year]
162E-04
4,OE-05
1997 1998 1999 2000 IAEA Workshop on Modeling of External Hazards in PSA 1 UV N'-,r 6 -10 2000.Sofia Bulgaria
Next Target - Dukovany Full Scope PSA
• goal * support regulatory decision making *to use PSA - based information in a risk informed maniner * action * full Scope PSA * make an update of SPSA and internal fires • take a credit for external hazards
IAEA Workshop on Modeling of External Hazards in PSA N--,,b-, 6 -10.2000. Sofia,Blgari
External Hazards in Dukovany NPP PSA
* objectives
- similar to internal hazards * to estimate contribution of external hazards • identify dominant contributors and accident sequences
6 IAEA Workshop on Modeling of External Hazards in PSA ~uJv No~emb~r 6 - 10. 2000. Sofi., B.g i
External Hazards in Dukovany NPP PSA
current status
-Analysis of PSA study completeness is being performed
Identification of site relevant External Hazards to extend a scope of PSA
[AEA Workshop on Modeling of External Hazards in PSA N- b 6 10.2000. Sofi., 8.g i
Identification of Site Relevant External Hazards
*2 groups have been identified
- hazards to be analyzed : initiators supposed to be with non negligible risk contribution • earthquake initiators • aircraft crashes • extreme rainfalls IAEA Workshop on Modeling of External Hazards in PSA No~'ember 6 10 000, Sofi.a 8 gari
Identification of Site Relevant External Hazards
-hazards initiators for which rough estimation of frequencies and safety impact should be performed to decide if to screen out them or not
• high or freezing temperature • storms
LAEA Workshop on Modeling of External Hazards in PSA ;~ V N-meabe 6 0. 2000,Sofia, Bulgaoia
Identification of Site Relevant External Hazards
*external man - induced events
- analyzed in frame of Safety Report
8 IAEA Workshop on Modeling of External Hazards in PSA J N-.~ber 6 - 10. 2000. offi. Bu/g-ri.
Seismic hazard in PSA for Dukovany NPP
current state
- Seismic PSA is not scheduled for a near future
- plant priority to start with design improvements recommended by seismic qualification project
LAEA Workshop on Modeling of External Hazards in PSA J N\ember 6 /0 2000, Sf.,/~ -i
Seismic hazard in PSA for Dukovany NPP
*current state
-nevertheless preparation of seismic PSA has been already started by
-identification of available information useful for PSA
-project for plant seismic qualification *walkdowns performed in frame of this project
9 IAEA Workshop on Modeling of External Hazards in PSA ~u~v~ N'.-b- 6 - 10.2000. Sofia. Bulgaw,
Seismic hazard in Dukovany PSA current state :
- team building up - seismic and fragilities experts - PSA and System analyses people - Operational Staff
- test example
-impact of seismic event to plant components is being now modelled to PSA to make trial evaluation
]0 GUERPINAR, Aybars ______
From: hans. hrusa~tuwien. ac. at[SMVTP:. hans. hrusa~tuwien.ac.at] Sent: Thursday, 16 November 2000 14:03 To: GUERPINAR, Aybars Subject: printd26.-82
E R N NE RU NG Datum: 16/11/2000
DVIR: 0051888 Mine GURPINAR Anisg. 4 A-1220 WIEN
Sehr geehrte Dame/sehr geehrter Herr! Folgende Werke waren oder werden demnachst zur Ruckgabe fliig:
1. Interval mathematics ed. by K. Nickel. - Berlin [u.a.]. - 1975. - (Lecture notes in computer science; 29). - ISBN 3-540-07170-9 .-ISBN 0-387-07170-9 Ausleihdaturn: 19/10/2000 Falligkeitsdatum: 1611/12000 Barcode: +EM22 195306 2. Neumaier, Arnold: Interval methods for systems of equations / Arnold Neumaier. - 1. publ.. - Cambridge [u.a.1. - 1990. - (Encyclopedia of mathematics and its applications ; 37). - ISBN 0-521-33196-X Ausleihdatum:, 19/10/2000 Failigkeitsdatum: 16/11/2000 Barcode:. +EM52806009 3. Moore, Ramon E.: Intervallanalyse Ramon E. Moore. [bers. von Dieter Pfaffenzeller]. - Munchen Wien. - 1969 Ausleihdatum: 19/10/2000 Fligkeitsdatum.. 16/11/2000 Barcode: +EM24991 904 4. Alefeld, Gotz: Introduction to interval computations GOtz Alefeld ; Jrgen Herzberger. - New York [u.a.]. - 1983. - Computer science and applied mathematics). - ISBN 0-1 2-049820-0 Ausleihdaturn: 19/10/2000 Faliigkeitsdatum: 16/11/12000 Barcode:. +EM1 1874305 5. Applications of interval computations ed. by R. Baker Kearfott: and Viadik Kreinovich. - Dordrecht [u.a.]. - 1996. - (Applied optimization ;,3). - ISBN 0-7923-3847-2 Ausleihdatum: 19/10/2000 Fligkeitsdatum: 16/11/2000 Barcode: +EM19761305
Mit freundlichen GruRen
Leihstelle Universitajtsbibliothekdter Technischen UniversitajtWien DVR:0051888 Resselgasse 4 A-i 040 Wien
Page 1 Tel. 01158801-44161 Verlangerungen:. 01/58801-44190 (14-15 Uhr) Diese Erinnerung war kostenlos, Mahnungen kosten Geld: Gebehren: Mahngebahr ATS 21,- Uberziehung ATS 2,- pro Tag und Band Wenn Sie entlehnte Werke nicht fristgerecht zuruckbringen (odler vericangern), verhelfen Sie dem Finanzminister (nicht der Bibliothek) zu Einnahmen, ver~rgern hre Kolleginnen, die auf die Bi~cher warten und verursachen bei den Bibliotheksmitarbeiter nnen zus~~itiche, unproduktive Arbeit. Verlangerungen im Internet: http://aleph.tuwien.ac.at
Page 2 XA0300588
Analysis External Events - Nuclear Power Plant Dukovany
by:M Hladky ______Analysis of External Events -
PSA Leec]I Analysis of External-
Events -fires -exrernal c catis are not includecd yet
7 ~~~~~~~~ShutdownPSA
-floids Rr ~~~~~~~~~~~~~~~~~~~ires -hea, y oa d rap --9 ~W ~ ~~~~~S- ernal cx cris rre not included[ et
Analysis of External Events - 2 Analysis of aircraft drop -
inral Satit .Analysis Repoirt -completed 1997 by Stevenson -emcar tcrd alter I0Ijar., of oprerarion 1995 -97 . -accordling to NTSS 50.~-SG-S5, -50-SG-Df5 -a all fr aperarianal unirs -data about air operation 1994-96. data of airera ft 1,,,- rafi isti c aitaach usedffr an-alsxxi, :ace ide nts Ir nm i st 10 years *.-xnaiisis,.f aicraft rlrolr 2. 1:xrera rrran-induced cx cots Basic: informrations:
-pi-uhibited air area above NPP - cylinder = ki, hl= 1.85- krrr -nearist military airport Njmgt', 8.5 ki
-nearist air. area 3.5 km from NPP
-arrports i distance to 60 km considered
F2 4~Anlysis.,; of arcrat drp-- AnalysisdrAnlyssooparcrft-rop-2 of aircraft 3 groups of air traflic considered Methodology Screening Probability Level (SPL) used j . - puiblic - acceptable risk criterium) < 0-/s-ear - rrilitiiiv - snrall :rircrafts (< 5700 kg)
P = PI P + P Detail analisis of military air traffic in the near ir probability of aircraft drop to main production surroundin,,s wiere performed. unit of NP PI. cau.Sed bs general air. operation Results p causedand h' landingstarting - the risk caused by air traffic p < 10/year p3..causcedl by air traffic in areas ol higher Nei%analsis wvillaby performed next Year. :ace ide nts Analysis of external man-induced events - Analysis of external man-induced events - 2
-completed by Ferjen~il in 2000 Mvethodologs:
-all man-induced events caused outside NPP area 1. Identification of the sources of the risk considered in surroundings < 8.5 kmn(with exeption 2. Preliminary assessment of the risk sources o f a ircra ft d1ro p) - ifSPIL> 1r/year go topoint 3
-nicthodolo,,v according to NUSS 50-SG-SS, I 3. Detail assessment of the risk sources - analyses of 50O-SG-D-5 and Netherlands guides CP11141, consequences (CPRI6E1 used 4. Analysis oftmissiles 7 -acceptance criterium SPL < 1 /-,eari, by SONS acceptable < 1` /year
Analysis of external man-induced events 3 Analysis of external man-induced events - 4
Sourcesof the risk: ~~~~~~~~Induastr%. iriiculture and military buildingos
-iiuustr- aicualture and military uildings- I~stli, -5itchiboaird tations -traffic routes w. t r pio, ei plant
-pipeline routes -stores of clinical products for agriculture -w5ater resersairs
Vl Ipes ofe ensts: . milit.arv hfililings
-fire explosion ~~~~~~~~~~~~Trafficroutes road Il/ 52 detail analysis -cloud oftire mnedium spreading .railway (oniv for NPP)
-toxic cloud] spreading
-toissiles Pipeline routes -only teoreticaly for earth gas
Analysis of external man-induced events -S
Conclusion: The risk caused by external man-induced events is neglitihle
-lowv frequency ot the event or
-ne-litibhic consequences
Analyses viere accepted by SONS.
2 External Hazards Considered for Paks NPP
by: T. Kiss XA0300589
External Hazards onsidered or Paks N\PP
Tibor Kiss Development and Analysis Section Paks NPP
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. 'I Origina Pant Design Soviet Construction Standards for Industrial Facilities were used - Meteorological aspects (Extreme wind, Snow load, Temperature and its possible combinations)
- No available documents for the other external hazards (earthquake, hurricane, flooding e.t.c.)
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. Periodic afety Revie-w *Review of the Meteorological aspects *Floods and low water level (1983) *Other cooling water related events (ice, shells) *Industrial, Military aspects *Road transport
- Poisoning effect
- Explosion
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. illPeriodic afety Revie-w (Continuati on)
* Transport accidents on the Danube (<1 e-7 /year) * Civil and military aircraft accidents (<2e-7 /year) * Earthquake
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. Earthqluakce Main activities:
- Reevaluation of the Plant site (S SE)
- Seismic Safety Technological Concept
- Improving the EQ resistance of the equipment,building
- Installation of the EQ monitoring and protection system
- Seismic PSA
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. II S~eismnic PS-A *Maincontractor:
- EQE International Inc. *Subcontractors:
- Westinghouse Electric Europe
- Siemens AG KWU
- VEIKI (Institute for Electric Power Research)
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. Scope o te eismnic PSA
* The reactor core is considered as the source of potential radioactivity release * Reactor operates at nominal power * The seismic PSA model for unit 3 (1999) * Earthquakes are the only initiators * The seismic PSA relies on the existing level 1 PSA
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. K~j~Transient Identifcation, ET Developmnent
*Review of the initiating events that can be induced by EQ *Examination of the selected transients responses *Identification of transient initiating failures that can be induced by an EQ, but not included in the PSA for the internal initiators due to the low frequency 'Development of functional ETs for single transient initiating failures *Development of a generic event tree for modeling plant responses to an EQ with combinations of single and multiple initiating failures
Workshop on External Hazards PSA, Sofia, 6-10. November 2000. SystemiAnalysis, Fault Tree Developmient
*Addition of seismic-induced causes for component failure modes that are included in the PSA models for internal initiators *Modeling of containment systems for an interface to level 2 PSA, and other systems important for the SSTC *Modeling of dependent failures *Modeling of seismic failures of structures and failures from spatial system interactions
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. ~iI~fHumnan Reliabilty Analysis
* Possible modeling places:
- Contributor to an initiating event
- Basic event at fault tree level
- Basic event at event tree level * Human-system interactions:
-Pre-initiator (type A)
-Initiator (type B)
-Post initiator (type C)
Workshop on External Hazards PSA, Sofia, 6-1 0. November 2000. Internal Flooding Analyses Results of Slovak NPPs
by: V Sopira XA0300590
Internal Flooding Analyses Results of Slovak NPPs
Vladimir Sopira
IAEA Regional Workshop on Modelling of External Hazards November 6-10, 2000, Sofia BULGARIA
RELKO Ld. p. 1 of 14
CONTENT
• Introduction • Overview of used methodology • Results of analyses a Conclusions
RELKO Ltd. p.2 of 14 SLOVAKIA inhobitans: 5.4 mil. Area: 49 000 sq. km NPP Bohunice: V - 2x44 OMW PWR type V230 FP PSA pre-reconstruction (incl.flooding fire) FP PSA post-reconstruction (mcl. limited fire seizrnic) S-PSA (9/2000 - 12/2001) V2 - 2x440M WPWR type V213 FP PSA, -PSA (ci. fire flooding) NPP Mochovce 2 x 44OM WPWR type V213 1st unit in operation since 998, 2nd unit - 1999 (units 3 and 4 in ~gstr~ucion)- FP PSA (cl. ire, flood , external azardsi ext erne weather, sizrnic, indu S-PSA -start /12001
p.3 of 14 ~~~~~~~~~~RELKOLtd.
RELKO Ltd. Engineering Et Consulting Services Private Company founded in February 1993 from the Nuclear Safety and Reliability Department of VUPEX Institute Main activities: Experience : • Safety analyses a Probabilistic safetyassessment of the • Evaluation of operational experience Bohunice V1 NPP (level ) • Probabilistic risk evaluation W `Living PSA - time dependent model PSA • Internal fire and flooding analyses level of the Bohun ice VI NPP • evelopmenFurther of dama* Probabilistic safety assessment of the plFurnher deeomn fmge Bohun ice VI NPP (level 2) • Drafts of safety requirements, a .Lhunic NPPstd frUit3o standards and instructionsJBouceNPV a Low power and shutdown PSA far Unit 3 Cooperation with: of J.Bohun ice NPP V2 • VUJE (Slovakia) a Level 1 PSA study for unit 1 of Mochovce • Electrowatt Engineering Services (UK) NPP • AEA Technology (UK) a Real time risk man itoring system (NPP a Nuclear Electric (UK) J.Bohunice V2) • SAIC (USA) Cretatvte SIEMENS (Germany) Cretatvte VEIKI (Hungary) m Optimization of technical specifications j JV (Czech Republic) a Level 2 PSA project for unit 3 of a ENCONET Consulting (Austria) J.Bohun ice V2 NPP * KEMA (Holland) a S-PSA J.Bohunice VI NPP * Westinghouse (USA) m S-PSA Mochovce NPP • FRAMATOME (France) • NUS-Corporation (USA), etc.
p.4of 4 RELKO Ltd.
2 Introduction
The internal flooding analyses receive less attention as potential cause of reactor accidents in comparison to other hazards, like fire and earthquakes. As a consequence, there are no well (international) established methods for the analysis of internal floods of NPPs.
Why floods have to be analysed: Internal floods may have a relatively greater potential to cause an accident with non-negligible risk Internal floods may be an important cause of multiple dependent failures Flooding is one of the failure mechanisms associated with CCF
Possible sources of equipment ooding or spraying: Cracked piping, flanges or welding Overfilled tanks or containers due to an operator failure Damaging of pump bushings Incorrect maintenance intervention
Critical equipment : Electrical safety related equipment like Instrumentation and control devices Motors of valves and pumps Non-hermetic cable junctions After a rise of an internal flood the reactor is shut down mostly manually by operator or rarely automatically exceeding the technological limits. A core melting will not occur if the systems necessary to shut down and for cooling down the unit are available.
RELKO Ltd. p. 5 of 14 Engineering &Consulting Services Overview of Used Methodology
The objectives of the internal flooding analysis are to evaluate the flood induced core damage frequency and to develop insights into the main contributors to the flood risk.
Major Tasks in a Internal Flooding PSA: 2
btaJb~mpact
E imation
E
In ~ ~ ~ ~ ~ ~ '
f nter$4[~ scenario
L) Data collectionand assessmentn 6ano s U orinternai~~~~~~~~I 1uue d Io
-~~~~ U7
Unc~R%1R~ and i analysi U rtffifgw nd manalysisaysi analsis nd mpoanaclansyss
RELKO Ltd. p. 6 of 14 Engineering & Consulting Services Overview of Used Methodology (cont.) Analyses Assumptions A potential risk as to a core melting is represented by internal floods that could influence safety systems of the unit and imply a manual shutdown of the reactor Only a single flood is assumed to occur in any of plant lo cat ion Spreading of flood to adjacent areas is taken into consideration unless it can be justified that the flood is contained in the original flooding location Natural phenomena, such as earthquakes, is not assumed to occur concurrently with a flood Loss of coolant accident is not examined as internal flooding source unless it is a consequence of an internal flooding event Floods induced by other initiating events are not considered Pipe rupture in general was considered in all water carrying systems with potential for damage to safety related systems and components Passive safety related components like pipes, cables, manual valves, check valves are assumed not to be affected by flood inR The internl flooding analysis relies on the plant response model developed for internal initiating events. Essential is the availability of PSA level model. Expanding an internal events PSA to internal flooding PSA requires a considerable amount of plant specific data.
RELKO Ltd. p.Z7 of 14 Engineering & Consulting Services Overview of Used Methodology (cont.)
Definition of flooding areas Every NPP? contains many potential sources of flooding and flood locations. In order to make the analysis tractable, a process was defined. The process consist of the following steps: Identification of safety related systems and components which availability has to be assured for the following safety objectives: »> Safe shutdown of the reactor from all operational con dit ion s > Assurance of residual heat removal and of maintaining the reactor in a safe condition >> Limitation of radiological consequences on environment Localisation of safety related systems and components Determination of possible flooding sources, barriers and routes of flooding medium » Identification of water- and steam-carrying systems >> Possibility of flooding from the neighboring rooms > Determination of operating parameters »> Determination of the covering leak >>Analysis of the maximum released water flow and volume This ifo plant doueW ,r~~tlekVFM~eh plant personnel and extensive walk down (e.g., Fig.]1.2, Tab.]1.38).
RELKO Ltd. p. 8 of 14 Engineering & Consulting Services Overview of Used Methodology (cont.)
Determination of components vulnerabilities and effects of flooding
Active safety related components are assumed to be affected by flooding if the following levels are reached:
• Motors of operated pumps and valves: lowest edge of the motor • Transducers, terminal boxes and connection boxes: lowest edge
Analysis of component vulnerabilities and effects of flooding is based on:
• Real elevation of the active safety related components • Component damage thresholds • Permissible loads on the building structure • Way of water level detection • Means of stopping the flow • Time available for operator action • Water depth in the room (calculated with the known flow rate, room volume, capability of water removal) • Time to damage of components
9 of 14 ~~~~~~~RELKO Ltd. p. o 14Engineering &Consulting Services Overview of Used Methodology (cont.)
Definition of initiating events
The Es that can arise can generally be categorised as one of the following:
. Events leading to controlled reactor shutdown . A reactor trip initiated by the operator . Transients leading to automatic scram . LOCA from the primary circuit
Screening by impact: An internal flooding area may be eliminated from the analysis if a flood source can not damage the components of interest.
The flood screening technique uses conservative assumptions. If any of boundary failures are likely to occur the scenario is retained for further analysis.
(e.g., low resistance of flooding barriers, doors left open, seal failures, the existence of other openings, failures of drainage valves, etc.)
RELKO Ltd. p. 10 of 14 Engineering & Consulting Services Overview of Used Methodology (cont.) Estimation of Internal flooding occurrence frequencies
. Internal flooding occurrence frequencies are established for each of the flooding locations . Where reliable flooding frequency data are available this is strongly preferred to generic data
The internal flood frequencies in our analyses were based on the general operating experience data related to the size and type of the flooding source. Mainly the IAEA data were used. Quantitative screening analysis Rooms with a flooding frequency less than 1 OE-06 per year were screened out. Screening for similarity of likely consequences Identified comparable internal flooding scenarios can be grouped for further analysis. The assigned frequency is that of the summation of the individual frequencies. Frecguency based screening The flooding scenarious that cannot be screened out in the preceding task are subjected to a frequency based screening. The flooding area can be screened out from the further analysis if the resulting core damage frequency is less than 1 OE-06 per year.
Detailed analysis The purpose of this is to reduce the level of conservatism in those damage situations that were not screened out in the preceding steps and to obtain a realistic estimation of the flooding risk.
of 14 ~~~~~~~RELKO Ltd. p. 11 o 14Engineering & Consulting Seixices Results of Internal Flooding Analyses for Slovak NPPs The Internal flooding analyses were performed for: • Bohunice NPP Vi - VVER440 type V230 (1993) • Bohunice NPP V2 Unit 3 - VVER440 type V21 3 (1994) • Mochovce NPP Unit 1 - VVER440 type V21 3 (1997, 1999)
Most comprehensive analysis was performed for NPP Mochovce. Scope of the analysis: 20 systems 826 active safety related components 6 buildings 87 rooms and flooding areas
Reactor Building • 1 9 safety important rooms can be endangered by internal flooding • Operator action to mitigate extent of flooding by stopping the outflow from the break and opening drainage was considered 30 minutes after detection of the break. • Assuming successful operator action the safety related system components will retain their availability. • If the room drainage valve will not be opened for some special cases there is the possibility to damage the safety related components. • Flooding in the rooms of Emergency Core Cooling Systems can lead to the unavailability of 1 out of 3 redundancies of ECCS. • Flooding in the room of Intermediate cooling systems (ICS) can cause after long time the failure of ICS MCP and ICS CRIDM which can lead after undefined time to unavailability of all MCPs (Fd 6.OE-5/year, CDF = 7.8E-1 2/year).
RELKO Ltd. p. 12 of 14 Engineering & Consulting Services Results of Internal Flooding Analyses for Slovak NPPs (Cont.) Turbine Building • The covering break is a circumferential break of a Circulating Cooling Water line of the Main Condensers. • The lowest safety related components cannot be endangered by this covering break. Middle Building and Control Room Building • Only one room can be endangered. The safety important SG safety valves, valves of the steam dump station to the atmosphere, quick acting valves on the steam lines, pressure and flow sensors and emergency feedwater system valves cannot be endangered. • The other safety important rooms of mainly electrical equipment cannot be flooded. Diesel Generator Building • There are 3 redundancies of DGs of unit 1 and 3 redundancies of unit 2 located in separate sections. • No flooding source can influence more than one redundancy of DG's. Essential Service Water Pumping Station • There are 3 redundancies of Service water system installed in separate sections. • No flooding source can endanger active SW components. Emergency Feedwater Building • The EFWS is strictly subdivided in 3 redundant channels whose pump units are arranged in 3 separate sections. • No flooding source can influence more than one redundancy of EFWS.
RELKO Ltd. p. 13 of 14 Engineering &Consulting Servi ces Conclusions
• The assessment of the flood risk was the objective of the internal flooding analysis for NIPPs Bohunice V], V2 and Mochovce. • All important flooding sources were identified. • The rooms containing safety important components were analyzed from the point of view of - Integrity of flood boundaries - Capability for drainage - Flood signalisation - Flood localization and liquidation - Vulnerability of safety system component • The redundancies of safety systems are located mostly separately and no flood can endanger more than single train. • Itcan be concluded that NPPs with VVER440 are very safe against the flooding initiating event.
of ~~~~~~~~RELKO Ltd. p. 14 of14 Engineering & Consulting Services Seismic Characterization of the NPP Krsko Site
by:J. Obreza XA030059 1
REGIONAL WORKSHOP ON MODELLING OF EXTERNAL HAZARDS IN PSA Sofia, November 6.- 10., 2000
SNSA PRESENTATION J Obreza Division for Nuclear Safety
TIME: 16:00 - 16:15 (Thursday, November 9)
TOPICS: *Seismic characterization of the NPP Krdko site
CONTENT OF THE PRESENTATION
*SNSA -intro *NPP Krgko Probabylistic Safety Assessment *INTRODUCTION (seismics) *NPP KR~KO SEISMIC DESIGN BASES *SEISMIC PSA ANALYSIS (IPEEE) - Steps in process - Fragility analysis - results - Risk quantification - results - Conclusions of seismic PSA *ONGOING SITE GEOLOGICAL AND GEOPHYSICAL INVESTIGATION *CONCLUSIONS
0~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ SNSA Organization
NPP Krgko PSA NPP Kr~ko Level Initial Project Model Risk Profile over Initiators
• Internal Events Result CDF7 - Total of 16 initiating events categories 5,44E-05/yr (23,7%) • External Events - Seismic events analysis 6,03E-05/yr (26,3%) - Internal flooding analysis 4,62E-06/yr (2%) - Internal Fire analysis 9,78E-05/yr (42,5%) - Other external events analysis I1,26E-05/yr (5,5%)
NPP KRSKO CDF 2,30E-04Iyr (100%)
4
2 NPP` Krgko PSA NPP Krgko PSA Project Update •Goal - Inclusion of plant changes (i.e. configuration/operational related) through the period January 1, 1993 till the OUTAGE99 (April 1999)into the integrated Internal/Extemnal Level /Level 2 NPP Kr~ko PSA RISK SPECTRUM model • Achievements - Reviewed and assessed changes and implemented the changes to the NPP Kr~ko PSA model where it impacts the model - Resulting in new RISC SPECTRUM based NPP Kr~ko PSA model "NEK 98", thet represents the initial project model plus implemented changes on the plant till end OUTAGE99, as well as plant OE till then - Documented update of the PSA model, and re-quantified results (ISA NPP Kr~ko Stage 1 Phase) S
INTRODUCTION
NPP Kr~ko is located on seismnotectonic plate Highest earthquake was recorded in 1917 with magnitude 5.8 at a distance of 7-9 km. Site (founded) on Pliocene sediments which are as deep as several hundred meters. No surface faulting at the Krgko site has been observed and thus it is not to be expected NPP Kr~ko is equipped with seismic instrumentation, which allows it to complete OBE (SSE) analysis within 5 minutes.
6
- ~~~~~~~~~~~~~~~3 NPP Kr~ko seismic design bases
• Long term investigation (domestic and international experts involved -, started 1964 intensive from 1971 - 1975 - Geophysical, geomechanical, hydrogeological and engineering seismological investigations) Suggested SSE - 0.22 g, at the end conservatively adopted SSE=0.3 g, OBE= 0. 15; • Free field response spectrum according to RG 1.60; • Soil structure interaction analysis - SAP IV (6 blocks, soil springs and dumping values developed by - D' Apolonia Consulting Engineers); • All components and systems qualified based on US standards and regulations
Seismic PSA - PEEE • Per IPE (GL 88-20) and SNSA licensing amendment NEK decided for seismic PSA to evaluate plant vulnerabilities on seismic events • Methodology developed in USNRC sponsored Seismic Safety Margin Research Program (SSMIRP) • First required step: - Development of free-field ground motions - Probabylistic hazard curves and uniform hazard response spectra - Work started in 1991 with all available domestic earth science experts under leadership of FAGG Ljubljana) - Work was reviewed by IAEA international expert teams (IPERS) - Results were published in 1994 - local earthquakes addressed separately (PSHA analyses was performed by FAGG Ljubljana)
* ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~8
4 Seismic PSA - IPEEE Steps in process are: PSHA analysis - Probabylistic hazard curves and uniform hazard response spectra Perform probabylistic response analysis (development of structural and soil model, foundation impedance and foundation input motion) Seismic walkdown(s) (equipment categorized into generic versus specific designs for subsequent fragility analysis) Fragility analysis Risk quantification (event and fault trees modified from internal events IPE) • Relay chatter evaluation (deterministic) • Local earthquakes treated separately - deterministic approach
9
Seismic PSA - Fragility analysis - results Fragility analysis was performed for 3 structures and 37 equipement items Screening level adopted based on negligible effect on CD3F (A=2.Og PGA, HCLPF=0.74 PGA) - majority of structures and equipment were screened out Fixes recommended based on fragility analysis - main control room ceiling support - leg supported tank (CCW surge tank) - cabinets poorly welded through shims - ALL IMPLEMENTED)
I0
5 Seismic PSA - Conclusions of seismic PSA * Plant structures and equipment are generally very rugged * Fixes defined based on walkdown and fragility analysis - all of them were performed * CDF dominated by LOOP and random failures (DG) * Seismic contribution to CDF7 dominated by few "low-fragility" (lack of documentation to higher qualification level for electrical components) * The seismic PSA successffilly demonstrated high seismic margin at NEK
Seismic PSA - Risk quantification - results
Results: Hazard Conditional Core Melt Frequency Core Melt Frequency Probability lot. 2,54E-3 3,31 1 E-3 8,433E-6
lot. 2 7,3 2E-4 1. 100E-2 8, 1 1E-6
Int. 3 3,1 1-S5 3.133E-2 9,766E-6
Int. 4 8,205-5 1,5885-1 1,3005-5
lot. 5 2.0015-5 5.5335-1 1, 10OE-5
mnt.6 3,605-6 8,855E- 1 3,188E-6
> 6 1,4053-6 1,00015+0 1,400E-6
Total 5,5005-5 12 Ongoing site seismic investigation
• During PSHA experts pointed out that due to lack of knowledge - very conservative estimations have been given. • IAEA review recommended additional investigation • Phase 1 - 1994-95 domestic institutions - additional geological and seismological investigation - geophysical investigation - first additional seismic profile measured
• Work reviewed by IAEA expert team; • Conclusion: additional 4 seismic lines shall be measured (square around NPP Krtko)
13
Ongoing site seismic investigation
In phase 2 (1997 - 2000) additional investigation: Continuation of seismological and geological investigation of Kr~ko valley conducted by domestic institutions (geological map, ...) Geophysical investigation founded through PHARE program - Field measurements and data collection (additional 45 km of profiles to be recorded with sismnic reflection method) - Data processing and data interpretation
14
7 Conclusions . NPP Ki-~ko seismic design based on US regulations and standards . Seismic PSA successfully demonstrated high seismic margin (analysis done for 2 time SSE value) - risk comparable with internal initiators;
S
15
8 XA0 300592
Safety of Ignalina NPP in Relation to External Events
by: S. Svirmickas Geoglraphy andl Demography
j~ LATYRI Nea rest cities
~~.1.4:~I;1J,:T.4I- ~Vilnius 130 km over 600 000 Daugavpils 30 km 126 000
I: ~ ~~~.IUAN uIJTr.7R Visaginas 6 km 33 200 1.JA I'll N 11A L 11.4
- V1LN[U~La kes
77 / ~~LUBLA Visaginas A ~~~~~~K ~~Druksiai
u ~~2*~~' u - ~~r.~~~ r~~a4 Apyvarde 8 km
11-1r..IAlksnas r 13 km
.313rhf ~River Daugava 30 km Population Distribution andl Near y Indlustrial Re ions
Population 30 km radius is about 85 000 (excluding Daugavpils) 15 km radius - 16.1 people/km2 (excluding Visaginas) 15 km radius - 63.1 people/km2 (including Visaginas)
Highway Ignalina - Dukstas 12km
Railroad Vilnius - St. Petersburg 9 km No chemical or oil process industries Population Distribution. and Near y Indlustrial Regions
/~~~JF rPanorama of auxiliary services:
.1 1 - NPP site, 2 - open distributive system, / j 3 - construction base, 4 - purification constructions,
/ ~~. 5 artisan well site, 6 -supply base, 7 -7motor transport department, 8 - car service station, 9 - industrial construction base, 10 - construction base, 11 - military base, 12 - health clinic, 13 - city of Visaginas, 14 - railway station, 15 - city transformer, 16 -the NPP transformer, 17 - recreational area. Meteorology
The Ignalina NPP is situated in the temperate climate zone The average annual air temperature is 5.50Oc 170 atmospheric fronts passes per year The average annual precipitation is 638 mm The average annual relative humidity of air reaches 8/ Hydirologlic Englineering Lake Drksiai: area - 49.32 k 2 (including 9 islands) maximum depth - 33.3 m average depth - 7.6 m predominate depth - 12 m length - 14.3 km width - 5.3 km perimeter - 60.5 km drainage area - 613 k 2 normal affluent level - 141.6 m minimum permissible level - 140.7 m Anthropogenic factors: control of discharge water circulation in the lake Geological and Seismological Effects
East-European platform The Baltic sineclize and the Mazur-Belarus antiklize junction The crystal foundation and sediment case are separated by a series of tectonic breaks Geological andl Seismologlical Effects
* '~ ~ *~K ~.) Glacial accretions of the ground-
P 5; ~~cover:
~~<~'K~~,~ / > 1 -swampy (peat, slimy sand); ~~~iwy; 2- alluvial (sand, gravel, pebble, 3 ~ ~ P sandy soil); ii '7~~~~ 3 - limno-glacial (clay, alevrit,
// ~sand); 4 - fluvi-glacial (sand, gravel, pebble);
/ 1 5 - water - glacial (sand, gravel, E~~h~~~ ~ Li j pebble, sandy soil); 6 - glucagon (sandy soil, sandy loam) of late (a) and early (b) stage of last glacier Probabilstic Assessment Projects at Ignalina NPP
1991 - PSA in conjunction with Barselina Project 1994 - full scope PSA Level 1
1996 - PSA Level 1 1997 - hazards evaluation for Service Water System 1998 - Control Protection System Safety Case 1998 - Ignalina NPP-1 gas gap closure probabilistic estimation Probabilistic Assessment Projects at Ignalina NPP
1999 - PSA Level 2 2000 - Risk Based Inspection Pilot Study In preparation - Living PSA Implementation In preparation - Safety Analysis Report 2 with Further Developing of PSA In future - PSA use in regulation and risk management In future - PSA applications in optimisation of technical specifications Hazardis Evaluation for the INPP Service Water System
Screening criteria NUREG/CR-2300,1983) 1. Damage potential 2. Mean frequency and consequences 3. Distance 4. Definition 5. Timing Hazards Evaluation or the
- -INPP Service Water System
The screening criteria can be applied to the number of external hazards Biological events 1 Coastal erosion 3 Drought 1 Fog 1 Meteorite 2 Pipeline accident 3 Transportation accidents 3 GlB103BI~059311XAI 1
Summary Review of PSA Topics in Connection with Romanian Regulatory Body Activities
by: S. Stoian Regional Workshop on M~odelling of External Hazards n PSA Sofia, Bulgaria, 6 -10 November 2000
Alexandru Stoian N ational1 Comm iss ion f or N ucl1ear Activities Co ntrol1 Roman ia *Romanian Regulatory Body Guide m PSA Level Requirements a PSA Level Project Status m PSA Level 2 Support Activities a Scenario for PSA activities a External and Internal Cooperation Guide for seismic PSA and seismic
kt ~safety margins *CNCAN contracted a guide for seismic PSA and seismic safety margins inline with the newest actual requirements in 5 USA a For evaluating the guide, CNCAN has requested an IAEA expert mission to be * held up to the end of this year pS level Requirem-ents m The PSA Project is developed by type "A"procedures - plant procedures approved by ONCAN, directly refferenced by the Operating License to be fulfilled - are under approval phase - These procedures are equivalent with QAM for PSA from NUREG 2300 and IAEA requirements PSA Level 1 Requirements of CNCAN
a First issued in 1991- Based on NUREG 2300&1 150 a Reiterated in OL for ernavoda NPP * Unit m Reiterated in Strategic Policy for
<~ Cernavoda NPP Unit 2
.1 ~~~~~~~~~~~~~4-1 PSA eve 1 Requirem-ents *I.AEA Requirements for PSA Level as for example
-IAEA - TEODOC - 1135 "Regulatory Review of PSA Level 1"
-IAEA - TEODOC - 1144 "PSAs of NPPs for Low Power and Shutdown Modes"5
-Safety Series No. 50-P-4 "Procedures for conducting PSAs of NPPs Level 1" PS-A Leve 1 Project tatus
• The PSA level project for Cernavoda NIPP Unit has already begun
• The actual status is reviewing of initiating events and fault trees
6 PS;.A Leve 2 upport Activities
mAnalysis Program for containment respond based on the SIDMV for containment m Research programs for PSA level 2 phenomenology j PSA Leve 2 upport Activities
a Training courses for external hazard events..
mThere is a SIDMV and deterministic studies related to plant behavior for external events.
-Q' "
,,,''I'M~~~~~~~~~~~~~~~~~~~~~~ Scenario or PSA activities
.Finalizing PSA level in parallel with * assessing seismic safety margins, followed by starting seismic PSA level U mDeterministic analyses cover generally DBAs (there is not expected that major modifications will result, taking into account previous outcomes) External and Internal Cooperation a Cernavoda NPP Unit has a seismic monitoring system (now under review for operational aspects) mCNCAN uses as consultant Stevenson&Associates Company mThe paper "Reglementation Strategy for Seismic Safety Margins and Seismic PSA" has been finished already
10 External and Internal Cooperation mFrom CNCAN will be allocated 2 coordinators: Mr. Petrescu and Mr. Stoian, following a training programme m Partime are 5 people m Regulatory body and Cernavoda NPP PSA level team interact continuously during the project development phases by exchanging information External and Internal Cooperation a CNCAN imposes specific requirements at each phase level, as well a For independent verification of PSA level 1 CNCAN has internal procedure based on AEA-TECDOC presented in
K 998 at the joint IAEA-NEA TOM held in Paris (CNCAN guide under preparation Is similar to Switzerland reglementation mode and complies with the IAEA/NEA TEODOC) ~~~~~~~~~~~~~ ~~~~~~~~~~~~12 Probabilistic Safety Assessment (Cernavoda)
b: Ml. Mircea XA0300594
PROD CERNAVODA iNrr UNITI
PR OBA B IL IS TIC SA FETY A SEMN
Experience and Strategies
presented by: MarianiaMircea PSA - PROD
ERNAODA xperenceand Strategies
*An ABA project named "Supportfor PSA related activitiesfor Cernavoda NPP" was agreed at the beginning of 2000.
*Obj ectives
- upgrading of capability and framework to perform deterministic analyses as support for PSA (accident analyses and severe accident analyses);
- upgrading of capability and framework to extend the scope of PSA model for Cernavoda NPP to include internal and external hazards (nternal fire, internal flooding, earthquake);
- upgrading of capability and framework to perform the Level 2 PSA for Cernavoda NPP. PRODP S -,%
CERNAVODAI Experience and StrategiesL
*Existing development and strategy:
- The Cernavoda Probabilistic Safety Evaluation (CPSE) Level 1 PSA Project was initially developed for the early design evaluation of Cernavoda NPP Unit 1.
- The PSA level 1 is requested as a condition for maintaining the NPP operating licence and there is a plant management commitment to develop full scope PSA Level 1 and also PSA applications for different operational purposes (risk monitoring, test and maintenance optimization, emergency procedures optimization, etc.) PRODPS -[1 Experience and Strategies
Existing development and strategy:
- In present CNE-PROD Cernavoda staff is working on site, with external support (Data Systems and Solutions) in order to develop the as-operated level 1 PSA model. This model will be finalized at the end of 2001 , then a consolidation stage (performance of supplementary deterministic analyses) will follow. The scope of this activity will be extended in the future to include internal and external hazards and level 2 PSA.
- Deterministic safety analyses for Cernavoda NPP Unit 1 include design basis accidents. It is intended to strengthen the existing capabilities for reviewing the adequacy of existing accident analyses and to perform new accident analyses in order to support the PSA modeling. In the same time it is intended to perform severe accident analyses. (.N.F~~P A
PROD P S A I
CERNVODAExperience and Strategies
*Status of Internal and External Hazards:
- Evaluation was done for the status of the development of the seismic PSA, fire PSA and flooding PSA.
- For seismic PSA it was concluded by IAEA experts that this work needs adequate human and financial resources. Decision was taken to coordinate this project from Cemnavoda but using specialists from external institutions.
- A Fire Hazard Assessment -HA is in progress for U 1. First stage, regarding the methodology, was reviewed by IAEA experts in November 1999. In present, work is done for Reactor and Service Buildings.
- Work on flooding PSA was not started yet. PROD P A
CERNVODAExperience and Strategies
*Status of Internal and External Hazards: For the extension of PSA scope:
- Capability will be extended to develop the seismic PSA, fire PSA, flooding PSA (procurement of supplementary computer codes and specialist training).
- The extension of PSA scope to include internal and external hazards will continue after the completion of deterministic studies and is expected that the effective inclusion in the PSA model will start at the end of 2002. VUJE Trnava Inc., Enginheerng. Design and ReerhPgnzto Division of Nuclear Safety Jeatmn
Noemtber 6-0 00
Extera azarsAnalsi ofMchoLe NPP
Prepared and presented by. Tibor Stoika VUJE Trnava Inc. Slovak Republic VUJE Trnava Inc.,Engineering, Design and Research Organization
Analyses of external events had been first time performed at the design stage of the Mochovce NPP showing sufficiently low contribution of external hazards to core damage frequency.
BUT Based on IAEA document Safety problems of WWER-440/213 NPPs and the categorization" (IAEA-EBP-WWER-03, 1996) the need of new reassessment arised due to discrepancy of some origin recommendations incompare with present IAEA ones.
Internal Mochovce NPP Nuclear Safety Improvements Program elaborated at the same time included the IAEA recommendations and following improvements were proposed to perform incontext of external events:
1. Seismic pro-ject and new locality seismic evaluation This safety improvement includes also some "on site" technical improvements in seismic stability of structures and equipments. 2. Unit specific analyses of extreme meteorologic conditions This safety improvement focuses on impact of feasible extreme conditions on NPP systems caused by: - rain, snow and hail storms, frost - winds - low and high temperatures 3. Analyses of external hazards caused by human In this safety improvement were specified: - feasible sources of explosions - analyses of hydrogen, gas and propane-calor gas depots - air crash risk
The results of these implemented safety improvements were considered inthe PSA study. VIJE Trnava Inc.,Engineering, Design and Research Organization
The External hazards analysis is also a part of Level 1 PSA Mach ovce NPP performed by PSA Department of VUJE Trnava Inc., Engineering, Design and Research Organization, Slovakia
Some partial analyses are performed in cooperation with following companies DS&S - SAIC, USA Geophysical Institute Academy of Science, Slovakia Relko, Slovakia
Basic documents:
NUREG/CR-2300 "PRA Procedures Guide - A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants" and IAEA SS No. 50-P-7 "Treatment of External Hazards in PSA for NPPs
The external hazards analysis consists of following parts: 1. Geography and plant locality 2. Nearby industry 3. Extreme meteorological conditions 4. Aviation 5. Seisrnicity
In some foreign studies an internal fires and floods are included in external hazards analysis. In case of Level 1 PSA study of Mochovce NPP both these events were part of internal hazards analysis. VCIE Trnava Inc,Engineering, Design and Research Organization
1. Geography and Plant Loclt
This initial part represents general and plant specific information essential for performing other analyses.
General information:
* wider area specification (surrounding countries, cities and other agglomerations with their distances from NPP site) * wider area geography (including geologic aspects of country as near mountains and soil characteristics and water reservoires) * climate (data analysis from evaluation of historical hydro- meteorologic records) * essential industrial facilities (list of all industrial facilities accounting area to 50 km around the plant site) Plant specific information:
* plant locality (specific characteristic of plant site) * site geography (including specific geologic characteristics of plant site) * plant site descripin(plant disposal and facilities description the plant consists of including the facilities situated on safety area of plant and which are essential for NPP operation)
TMAOEA CL
'~~~, ~~~~ 3 * ' tz~~~~~~~~~~v~~~o"~~~~~ovace
>1 O K~
- '411~~~~~~~~~~~~~~~~~~~~~~KK I NtC(KQVO,
0.. ,a' ,'3c1~~~~~~~~~~~~~hl .o VUJE Trnava Inc.,Engineering, Design and Research Organization
2. Nearby industry
Analyses of all industrial, transportation and military facilities consider following:
* distance from plant site (3 allocation zones - 0-5 kin, 5-20 km and 20-50 kin)
* kind of industry (activities and products)
* stored dangerous and explosive materials (quantity and kind of materials) * transportation (roads, railways and river traffic including those related to nearby industry)
* pipelines (as and oil lines)
Based on above mentioned aspects evaluation analysis of feasible events (explosions, toxic chemical material releases, fires and floods) was performed.
Analysis result -on surrounding plant territory is not any source of initial events which could menace the NIPP operation and lead to explosions, creation of combustile clouds, toxic materials release, large fires or harmful materials contamination with probability higher than I -6year`. VLYE Trnava Inc,Engineering, Design and Research Organization
3. Extreme meteorological conditions Analyses of meteorological and hydrological data consider following aspects: * local climate evaluation (temperatures, humidity, circulation conditions rainfall and snow, atmospheric pressure and winds; based on historical measurement data) * local meteorology, hydro-mneteorology and hydrology data anayss(average and extreme temperature values, winds, rainfalls, rivers and reservoirs flood historical data) Based on these aspects evaluation following analyses were performed: Analysis of feasible events resulting from winds Analysis of feasible events resulting from extreme cold (calculations of temperature decrease in selected rooms and systems with special emphasis to safety systems) Analysis of feasible events resulting from external floods (maximal flooding calculation)
The evaluation was performed inaccordance with document "Extreme Meteorological Events in NPP Siting Excluding Tropical Cyclones, Safety Series No.50-SG-Sl1A, IAEA, Vienna I' Analysis results:-Considering the location and the climate of' Mochovce NPP site; its facilities are not menaced by external flooding and extreme winds. VCJE Trnava Inc.,Engineering, Design and Research Organization
4. Aviation Analyses of all surrounding aviation and its facilities consider following aspects:
* airport allocations (including distances from NIPP site) * fly frequency (of civil, sport, agriculture, special and military airplanes) Considering of these aspects analysis of feasible events resulting from aviation was performed (air crash frequencies calculation separately for each class using SDV and SPL methods recommended for these purposes by below listed document).
Calculated probabilities:
-civilflights 2,93 x 108'
- ~~-Sport flights ,59x0 - agriculture flights 1,75x10 - military flights 1,36 x 1T
The evaluation was performed inaccordance with document "Extend Man Induced Events in Relation to NPP Siting, A Safety Guide No.50-SG-S5, IAEA, Vienna, 1981".
Analysis result - based on above listed information evaluation the total air crash contribution to overall risk of core damage is 6,63.lO`year-1 i.e. less than recommended value 1.lO7year' VWIE Trnava Inc., Engineering, Design and Research Organization
The seismic hazard curve for Mochovce NPP as the first part of seismic hazard estimation was developed by Geophysical Institute Academy of Science in Bratislava (Slovakia) and consisted of following steps:
* seismic and geological database development (based on historical and real-earthquake measurement data)
* seismotectonic model development (STM) as synthesis of seismic and geophysical data correlation - identification and characterization the seismic sources in vicinity of the site (source zones) - minimal and maximal magnitudo choise
* attenuation characteristics specification for giround-motion - macroseismic intensity attenuation - analogic world region comparation and selection - peak ground acceleration attenuation
* probabilistic calculation and hazard curve production - seismic menace probability calculation - logic tree construction - probability assessment of each logic tree limb - seismic menace calculation for logic tree by SEISRISK Ill code - hazard curve production expressed as return period versus ground motion parameter - peak ground acceleration The origin assessment in Mochovce NPP Pre-operation safety report" was performed for horizontal acceleration value 0,06g for 6'hdegree of MSK-64 intensity scale. Present evaluation uses the lowest PGA value of horizontal acceleration equal to 0,1g and vertical acceleration equal to 0,067g for 7h degree of MSK-64 intensity scale (recommended values). VUJE Trnava Inc,Engineering, Design and Research Organization
The result of the first analysis is site-specific seismic hazard curve that provides the range of ground accelerations that the plant may reasonably be expected to encounter. Seismic hazard curve for Mochovce NPP:
Peak Ground Acceleration
1.00E±4- IH
n~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
r~~r E_ _1.O- - _
-1 *~-----~------Percentile __
o . I~~~~~~~~~~~~~~~ t eretl
L... _ _ _ _ I J~~~~~~~~4tP rc ntl
I _ _ _ __W
r T - 1~~~~ea
the_ seismic 16thtierceniile Theremainingparts of overall seismic hazard estimation~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~. are in proress stil not finihercntl VCJE Trnava Inc,Engineering, Design and Research Organization
5. Seismicity
.Seismic risk assessment addresses potential scenarios that could be initiated by a seismic event and which model the plant response.
The seismic hazard estimation consists of:
* seismic hazard curve generation - based on seismic and geological information * component/structure fragility assessment - developing a list of components and structures susceptible to seismic event which need to be considered in the PSA; divided according to seismic categories - developing the components and structures fragility curves and evaluation of HCLPF (High-Confidance-of Low- -Probability-of-Failure) capacity * accident sequence delineation - adding the seismic failures to the internal fault tree models to address the effects of the seismic event and generating the cut sets using Risk Spectrum code
* system analyses - accident sequences quantification * seismic core damage frequency evaluation - overall seismic risk results
The seismic risk assessment is performed in accordance with document IAEA-TECDOC-724 Probabilistic Risk Assessment for seismic Event" October 1993. Seismic Re-evaluation Process in Medzamor - 2 NPP
by: P. Zadoyan XA00596
~ Seimc Re-iievaluati'on Process in Medzamor-2 NPP
- ~P ZadyanANRA Saey Assvessment Section, State nspector Ph.D.
1. Program Im~plementation Statuis 2 2. Re-Ehvaluation Programi Structure
...... "". 3. Regulatory Procedure and Review Plan
#7~~.... 4. Current Tasks and Practice ...... 5. Regrulatory Aseset anid Researches
'"I"4,~~~~~~~~~~~~~~...... 1. Program Implementation Status
completed T asks Review Level Earthquake determi nation (Armenergo(:seilsmproject); Seismic Re-Evaluation Programn Development (AN.RA-A.NP'P);. Intermediate Walkdown (DOE). Jn Process Tasks Generation of 3 component Seisrnic Input (NIAEP); As-Built and Configuration Data (NIAEP); SSE.L (Gidropress) FRS and Acceleration Analysis; FurtherTasks Soil, Structure, System Capacity Evaluation; Walkdown (Final); Identify Candidates for Upgrading, Design, Implementation 2.The Seismic Safe~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~tyRe-Ev""aluaition Program Structuire
SEISMIC RE-EVALUATION PROGRAM
General Regulaty Procedure Wortkplan List of Thas Review* Plan Time Schedule ...... l; ......
...... 4 c .
...... 3.1 Regulatory Procedure and Review Plan
...... Regulation level laK Item Refer. Submission ion, Info Coord. Appr(v 5 6 7 ...... Selsmic Re-Evaluation 1; 2 3 41 Hard copy Program ......
...... R .L.E ...... p y 2 Generation of 3 com.. ponent ...... 12; 8 91 ...... fi l e s TH Acceleration ......
Collection of as-bullt data. [2) Hard copy configuration property
...... 4 Geotechnical Data 12; 10; 151 Ila ......
5 SS Procedure and Scenario [2] Hardcopy
...... 6, SSEL, ...... [2, 151 H ard copy ......
Modeling of Basement [2; 8 9]
AW S ~~3.1 Reguilatoryv Pr~ocedurte and ev~iew IPIan (oit.)
jMW
I ~~~~~~~~~~3 7
4 A'~dOlh~ fAluhr
-~~~~~ '~~~~~~~ . T o b...... -
,M - 1C R ndc'fraicrs 8; 9 Anlyi fileis
"2' 1 2 ~~~~~~~~~~~~~~~~Structure ~~~ 6 7; 8 -t icdysis file~s
t -______Lva iation A
'7 - 13 Seisitic walkdotWn of Eq. ;( 7 ; ifrno a nd. Ins sse
1)isribtionsy temoil]~& 8; 161
1 6 clerita the u~u lg12; 6; '7. nomto
I LkDsign ai ') ""lnimZnai~-8 19 9; 16j epr
FIIdIItJ~. A &Lcttin 3 AllCŽ\lC m lee ...... -...... n e.Ic lvy I p~a iig Me i; u -*2 P t i'V~iS 9 828> ...... nieI i's1 l h risli 2P - lit~tii n~eiai i f[ie;\N~ ;;i.2' - 97
...... I...... ~iii i~ ~ ~i. 99.. t
....~ ...... ~ ~ ~ ~ ~ ~ ~ '.W. At WF ...... u,. i. n~ ~ 2& l ...... fS nitaf--8.S nateA a s sn al;yR lte een; tt~e t-e.1)2 ...... N RitS ndr e e la o h eie fs e na i eot h P hper St hra :tsis. ~scno
11.iil e13.1 f 2hARA egulN Si atrye- Procnlhta sIt erduiriPeg and111 Rjlev itew Pl8''
1 .3.ly IASeriesI N.S dt No. 50-S(.i-SS. Sn fey Guide ''Safety Aspects O..f...... Fonnd...... a.ti...... o...... f NIP"....1986..
1-tE~lGuide..A Na~y -5(1-I) 15, teismir I )csi en aiM Quali [jeation for Nucleat Power Plains''. 1992
.5 1 11 [no t li r tA. Seiir Sa e~e isti t ANIeii i Reiw of tile'. Seismi7 .12>RevlamnndIgri l~~roginmn~~~~Corde Anaeil an NPP. (88F ai GFv-iia dat reIe.A9N1 Fllo-ipA,,1 SS19)99 I. un 99'
16.-(111>115 A hi;' sri sniie qual i heal~~~~~~~~~~~~~finE1F N neer1. ln E ili 92
17. FJA-ASMEBoi~eand Pessur Vesse ('ode sect on III 1992 Nt ~~~4.Current Tasks and Practice
4, 1 Review LvlEarthqduake:. Free Field Horizontal PGA=0.35g .84thPercentile;, ~~ * ~~~A 0" Response Spectra Shape NUREG/CR-0098 for Roc Site;
Verical Aceleratin Comfponent -2/3 of the Horizontal. 4a Inemeit Walkdwn Obsejrvatis:' Deficiencies in Anchorage of some Components; Quaiyof Originall Istalled Welds; The ateal restrainlts of Pipe and Systemns should be Analyzed and Tested; Prtiorities should be st fort potenitial modifications to systemhs based on safet techil economic and easy o iiplementain ...... 4..3 BaicReuiemnt fr.enraio.o.T .... * Eah alulte.secru.o te.rtfiia.T.soud.nvloe.heDeig .... ReposeSpctum(n.mrethn...oitsan.n.mre10) ... I.heiteete reunc.ane ERI..1.R..S.~I~i t~$~
...... Additional43 BsiRequirements inAC8iseof Aee4-86;o T 7; ~4.4 Essenitial Issues in FRS anid Stress Analysis
Min Building * Soil Proerties Senstivit Aaysis;
*HEydrodynamic Mode of of Spend Fuel Pool; *Crane Bearing Displacement Anialysi1s; *Load Comfbinatli (ith accounting Thermal Effets-~ACI); *Steam Generat and MJV Supor Modeing Desel GeoneratortBii~ng *Foundation Base Mat Accounting; 2 * Diesel Foundation nfluence on an Impedance Funcdtion; *Fouindatin Embedmfent effect;, EvaluationfLocal Strengthening Effect (aidep forestaring) 4.5 Geotechnical characteristics of the ANPP site
Generic properties of layers
______~~~~~~~~ 2.5 150~~~ 1.25~~~~~~1
a ~~ 1.8~2 600 1.45 3 11.7 2300 2.4 4 6.0 650 21.0 5 24 2200 2.4 Geological profile up to 5Dmdet 6 5.0 700 2.1
S h ear w a ve veto c ity v alu e vs So0ilprof ile .3000 (~~~~~Site In v e St ia op ,19 9 5
0
0 ~~~~~~~~20 4 0 6 0 dp t 8 4.6 Generated Acceleration1 TH
......
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...... a...... 3 * H * T ...... p .3 ...... - Rv f1ftA H~~~~~~~~~~~~~~~~~~~ee*13C~~~~~~~~~~~~~~ppm 1p 14a in p 44 43*~~~~~~~~~~~~m
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'3"~ 3, ~ 3"~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~~~~~~~~~~...... U M M ...... 5...... Regulatory ...... - ...... Assessment ...... and ...... Researches
......
......
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......
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...... 5.1 DG Natural Frequencies (Mode ------...... 90) ...... AW
...... : I x N ......
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9,~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~......
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......
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...... 7215
NORM
4- Aws W911.1RI.
......
...... 5.2.Other.Actiities.of.the.ANRA.Safty.Assessment.Sectio ...... D...... l....p....ue...ts ...... TH.Generation....(DYN..E.7); ...... PS...... A nalysis;...... * nal sis (Sup.SA SI) ... .. Anlysisfor.he.Achorae.Upradin.Puroses ...... R eg ula...... o..ns_
...... * Coordin tneo Activities ongthe Loca Stittions;eseteto ...... JAATCPojc.AM9/0.adSSMPrvsin ...... * ... Periodic Reports...Submission; * Permanent work with~~~~~~~~~~~~~~~~~~~~~~~~~~.the..... Laboraory.o...... Ses olg of..Plant...... AHPA PA ""'A'~~~~~~~~~~~~ACHC3P
CpaO8TLIBaHne HeT ocaaoBa CHA3, oCauOR peaKTopa
'PeKO~H~agHHn flapameTpbI 3eMAeTpICeHH5I,
...... A
~~A r~1-petbimeHJne 113. Cpa6aTblBarnxe CI ...... ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .