Preliminary assessment of innovative reactor SVBR … · Preliminary assessment of innovative reactor SVBR-100 in ... compared with LBC cooled ship ... Parameters EP1.1.1.1-EP1.1.1.5

Preliminary assessment of innovative

reactor SVBR-100 in the area of "SAFETY".

Valerii KorobeinikovState Scientific Center Institute of Physics and Power Engineering / Russia

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Consultancy Meeting On the Review of Innovative Reactor Concepts

for Prevention of Severe Accidents

and Mitigation of their Consequences (RISC)

23 to 25 November 2015 Vienna, Austria

Contents

Introduction

1. INPRO approach in the area of safety

2. Basic principle BP1(Defence in depth)

3. INPRO basic principle BP2 (inherent safety)

4. INPRO basic principle BP3 (risk of radiation)

5. INPRO basic principle BP4 (RD&D)

Conclusion

2

Introduction

SVBR-100 design meets the Generation IV innovative nuclear systems key requirements.

Safety and Reliability:

� Inert to water and air lead-bismuth coolant with very high boiling temperature (1670 оC)

� Integral nuclear system design without high pressure in primary circuit

�Passive safety systems

�Any radiological emergency possible for SVBR reactor could not lead to radioactive emissions into the atmosphere

�No hydrogen is released during SVBR operation

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1. INPRO approach in the area of safety

� According to the INPRO methodology, an assessment in the area of safety should be primarily based on a comparison of the innovative reactor design against a reference design, i.e. one that has been put in operation by the same supplier and could be considered as its most advanced design.

� As the innovative reactor was selected the SVBR-100 design.

� The INPRO objective is to confirm the long term sustainability of a NES;

� INPRO methodology is an assessment method:

� INPRO assessment in the area of Safety is not a Safety Assessment using the IAEA Safety Standards (as defined in the IAEA Safety glossary);

� Safety Analyses are necessary prerequisites for the INPRO assessment in the area of Safety.

� The INPRO assessor needs results of safety analyses as input to perform judgments whether INPRO Criteria are met.

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1. INPRO approach in the area of safety (cont.)

5

Basic

Principles-4

User

Requirements-14

Criteria-37; EP-21

(indicators +

acceptance limits)

-goals for development of sustainable NES

-actions to be done to meet the goal

-assessor’s tools to check whether

the actions are done properly

Structure of the INPRO requirements

INPRO approach in the area of safety (cont.)

6

Structure of the INPRO area of Safety

6

The stages of heavy metal

coolant technology

development

APL-705 experimental

(1971)

Test facility (1951)

APL-705 serial

(1976-1996)

SVBR-100

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2. Basic principle BP1(Defence in depth)

Basic principle BP1: Installations of an Innovative Nuclear Energy System shall incorporate enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations.

The basic principle BP1 contains 7 user requirements UR1-UR7

UR1.1 (robustness): Installations of an INS should be more robust relative to existing designs regarding system and component failures as well as operation.

UR1.2 (detection and interception): Installations of an INS should detect and intercept deviations from normal operational states in order to prevent anticipated operational occurrences from escalating to accident conditions.

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2.Basic principle BP1(cont.)

UR1.3 (Design basis accidents): The frequency of occurrence of accidents should be reduced, consistent with the overall safety objectives. If an accident occurs, engineered safety features should be able to restore an installation of an INS to a controlled state, and subsequently (where relevant) to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention should be minimal, and should only be required after some grace period.

UR1.4 (Release into the containment): The frequency of a major release of radioactivity into the containment / confinement of an INS due to internal events should be reduced. Should a release occur, the consequences should be mitigated.

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2. Basic principle BP1(cont.)

UR1.5 (Release to the environment): A major release of radioactivity from an installation of an INS should be prevented for all practical purposes, so that INS installations would not need relocation or evacuation measures outside the plant site, apart from those generic emergency measures developed for any industrial facility used for similar purpose.

UR1.6 (Independence of DID levels): An assessment should be performed for an INS to demonstrate that different levels of defence-in-depth are met and are more independent from each other than for existing systems

UR1.7 (human machine interface): Safe operation of installations of an INS should be supported by an improved Human Machine Interface resulting from systematic application of human factors requirements to the design, construction, operation, and decommissioning.

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Monoblock

vessel

Control rod driver (50)

Core

Steamgenerator

module (12)

Pump (2)

Safeguard

casing

Layout of SVBR-100

11

Reactor operates without partial refuelings.

SNF is unloaded cassette-by-cassette.

Fresh fuel is loaded as a single cartridge.

Fuel element is of a container type.

Wrappless FSA (total number is 61).

Boron carbide is an absorbing material in CPS

rods (enrichment in 10В - 50%).

Lowered power density ( ~ 2 times less) as

compared with LBC cooled ship RFs.

FSA without rods (7)

Control and compensating

rods (12)

Compensating

rods (22)

Automatic

control rod (2)

Additional emergency

protection rods (12)

EP rods (6)

Basic safety systems

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Criterion CR1.1.1 (robustness)

� Indicator IN1.1.1: Robustness of design (simplicity, margins).

�Acceptance limit AL1.1.1: Superior to existing designs in at least some of the aspects discussed in the text.

�Evaluation parameters (EP)

�EP1.1.1.1: Margins of design

�EP1.1.1.2: Simplicity of design

�EP1.1.1.3: Quality of manufacture and construction

�EP1.1.1.4: Quality of materials

�EP1.1.1.5: Redundancy of systems

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Parameters EP1.1.1.1-EP1.1.1.5 estimation

� EP1.1.1.1: Margins of design

� The design of the reactor core and the control rods allow to compensate for the most unfavorable from the point of view of nuclear safety the combination of uncertainties.

� Evaluation parameter EP1.1.1.1 has been met by SVBR-100 design

� EP1.1.1.2: Simplicity of design

� The tract of circulation of the primary coolant SVBR-100 was constructed simpler than for reactors of nuclear submarines. The branched pipelines of the main and auxiliary tracts were excluded for SVBR-100 reactor.


� EP1.1.1.3: Quality of manufacture and construction

� In the design SVBR 100 were taken into account the shortcomings that have been in submarines. Steam heating system design was simplified. The coil system of steam heating is enclosed in a protective casing monoblock. For the connection of the steam generator modules heads with shells was applied welding.


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Parameters EP1.1.1.1-EP1.1.1.5 estimation cont.

� EP1.1.1.4: Quality of materials

� The SVBR- 100 used materials that are approved for use in nuclear power.


� EP1.1.1.5: Redundancy of systems

� Redundancy of safety systems is provided in accordance with the requirements of regulatory documents. In particular, for transition of the reactor into subcritical condition the control rods systems EPR, CR and AEPR are used. Each of these systems may provide a transition of the SVBR-100 in subcritical condition and compensate the power reactivity effect. For passive cooling down was applied 4 independent channels.

� Increased redundancy of systems has been demonstrated in the SVBR-100. Thus, EP1.1.1.5 has been met.

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Final assessment of CR1.1.1 robustness

• Final assessment of criterion CR1.1.1 — robustness of

design

The assessment of the evaluation parameters of the

first criterion CR1.1.1 has confirmed an increased

robustness of the SVBR-100 reactor design in several

aspects compared with reactors of nuclear submarines.

Therefore CR1.1.1 has been met by the SVBR-100

design.

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Criterion CR1.1.2 (operation)

� Indicator IN1.1.2: High quality of operation.

� Acceptance limit AL1.1.2: Superior to existing designs in at least some of the aspects discussed in the text.

� Evaluation parameters (EP)

� EP1.1.2.1: margins of operation.

� EP1.1.2.2: reliability of control systems.

� EP1.1.2.3: impact from incorrect human intervention.

� EP1.1.2.4: quality of documentation.

� EP1.1.2.5: quality of training.

� EP1.1.2.6: organization of plant.

� EP1.1.2.7: availability/capability of plant.

� EP1.1.2.8: use of world wide operating experience.

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Parameters EP1.1.2.1-EP1.1.2.8 estimation

�EP1.1.2.1: margins of operation.

�Operator errors during operation SVBR-100 does not lead to severe accidents with core damage and release of radioactive materials due to advanced self-protection properties

�Evaluation parameter EP1.1.2.1 has been met by SVBR-100 design

�EP1.1.2.2: reliability of control systems.

�Better than a submarine. It will be proved after the development of the project I&C systems.

�Thus, the evaluation parameter EP1.1.2.2 is deemed to be met.

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Parameters EP1.1.2.1-EP1.1.2.8 estimation (cont)

� EP1.1.2.3: impact from incorrect human intervention.

� Operator errors during operation SVBR-100 does not lead to

severe accidents with core damage and release of radioactive

materials due to advanced self-protection properties

� Evaluation parameter EP1.1.2.3 has been met by SVBR-100

design

� EP1.1.2.4: quality of documentation.

� The quality of the documentation meets with the requirements

of regulatory documents


design

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� EP1.1.2.5: quality of training.

� Training will be conducted in accordance with the requirements

of regulatory documents and international experience,

including - using the simulator SVBR 100


design

� EP1.1.2.6: organization of plant.

� Use experience of Russian and foreign nuclear power plants


design

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� EP1.1.2.7: availability/capability of plant.

� High capacity factor is ensured by long-term fuel campaign

(2000 еff days)


design

� EP1.1.2.8: use of world wide operating experience.

� SVBR 100 project was based on the experience gained in the

global nuclear power industry, as well as resulting in the

operation of reactor units submarine project 705 and 705K


design

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Final assessment of criterion CR1.1.2 (operation)

The acceptance limit AL1.1.2 (quality of operation

of INS is superior to existing designs in at least some

of the aspects discussed) of CR1.1.2 is met, if

evidence is available to the INPRO assessor that the

assessment of the above defined evaluation

parameters provides positive answers.

Final assessment of criterion CR1.1.2 — operation

Thus, the SVBR-100 reactor meets requirements of

criterion СR 1.1.2

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Criterion CR1.1.3 (inspection)

� Indicator IN1.1.3: Capability to inspect.

�Acceptance limit AL1.1.3: Superior to existing designsin at least some of the aspects discussed in the text.

The need for inspection SVBR-100 is substantially lower than that of acting NPP due to the high reliability.

The inspection program for SVBR-100 meets regulatory requirements.

Inspections for SVBR are simplified due to the special measures.

Therefore criterion CR1.1.3 has been met by SVBR-100 design.

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Criterion CR1.1.4 (failures and disturbances)

�Indicator IN1.1.4: Expected frequency of failures and disturbances.

�Acceptance limit AL1.1.4: Superior to existing designs in at least some of the aspects discussed in the text.

�For innovative designs the expected frequencies of initiating events (failures and disturbances) should be reduced relative to existing designs. Reduction could be achieved via an increased robustness.

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Final assessment of criterion CR1.1.4 (failures and

disturbances)

� The SVBR-100 technology is based on forty-year experience (80 reactor-years of operating) of development and operation of LBC cooled RFs at nuclear submarines (NS) and ground facilities-prototypes. In the process of mastering this new technology a series of scientific and technical problems has been solved.

� The projected failure rates and irregularities for SVBR-100 are lower than on the nuclear submarines.Thus criterion CR1.1.4 is expected to be met.

Final assessment of user requirement UR1.1 robustness

All criteria were judged to meet their acceptance limit. For criterion CR1.1.2 it is recommended that additional information be requested from the designer.

Thus, user requirement UR1.1 is deemed to have been met by the SVBR-100 design.

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Criterion CR1.2.1 (I&C and inherent characteristics)

� Indicator IN1.2.1: Capability of instrumentation and control system (I&C) and/or inherent characteristics to detect and intercept and/or compensate deviations from normal operational states.

�Acceptance limit AL1.2.1: Key system variables relevant to safety (e.g., flow, pressure, temperature, radiation levels) do not exceed limits acceptable for continued operation (no event reporting necessary).

� INPRO has defined the following evaluation (EP) parameters for CR1.2.1:

� EP1.2.1.1: continuous monitoring of plant health.

� EP1.2.1.2: dynamic plant analysis.

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Final assessment of criterion CR1.2.1

� EP1.2.1.1: continuous monitoring of plant health.

� The system of the control rods SVBR-100 provides control installation options in the volume corresponding to the requirements of regulatory documents. In addition, continuous monitoring of equipment is provided by the system of technical diagnostics (ultrasound diagnostics provided SG leaks and failures MCP). For early detection of anomalies in the temperature field at the outlet of the core is provided by thermal grill.


� EP1.2.1.2: dynamic plant analysis.

� Analysis of the dynamics SVBR- 100 was made in the development of the technical design of the nuclear reactor.


� Both EPs above provides positive results. The reactor SVBR-100 complies with the criterion CR 1.2.1

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Criterion CR1.2.2 (grace period)

� Indicator IN1.2.2: Grace period until human actions are required.


� The acceptance limit AL1.2.2 is met if evidence is available to the INPRO assessor that a grace period of at least 30 minutes is assured by the design of the plant.

� Final assessment of criterion CR1.2.2

� The period of non-interference with the cool down installation due to the natural circulation of primary and secondary circuits, as well as in the passive heat removal through steam generators (SG, PHRS) is 72 hours (the duration of the period provided in the supply of water tanks (PHRS)).

� The reactor SVBR-100 complies with the criterion CR 1.2.2

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Criterion CR1.2.3 (inertia)

� IN1.2.3: Inertia to cope with transients.


� Final assessment of criterion CR1.2.3 (inertia)

� The ratio of power reactor facilities nuclear submarines 705 and 705K and capacity SVBR-100 is 1: 1.8 and volume of primary coolant to SVBR-100 is 30 m3 against 6m3 on submarine reactor units, i.e., 5 times more. Thus, SVBR 100 characterized much greater thermal inertia of the first circuit.

� Thus criterion CR1.2.3 is expected to be met.

� Sufficient grace period and improved inertia of the of the primary and secondary coolant system of SVBR-100 have been demonstrated, i.e. criteria CR1.2.1 , CR1.2.2 and CR1.2.3 are met by the SVBR-100 design. Thus, user requirement UR1.2 is deemed to have been met by the SVBR-100 design.

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Criterion CR1.3.1 (frequency of DBA)

� Indicator IN1.3.1: Calculated frequency of occurrence of design basis accidents. Acceptance limit AL1.3.1: Reduced frequency of accidents that can cause plant damage relative to existing facilities.

� The acceptance limit AL1.3.1 (Reduced frequency of accidents that can cause plant damage relative to existing facilities) of CR1.3.1 is met if evidence is available to the INPRO assessor that the INS shows lower frequencies of accidents with respect to existing comparable designs. For an INS, these frequency values, as well as the consequences of accidents, imply larger uncertainty ranges than those for existing reactors.

� Final assessment of criterion CR1.3.1 (frequency of DBA)

� The frequency of accidents in relation to comparable existing installations is reduced. It will be evaluated in more detail after the completion of the technical design SVBR- 100.

� Thus criterion CR1.3.1 is expected to be met.

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Criterion CR1.3.2 (grace period)

� Indicator IN1.3.2: Grace period until human intervention is necessary.

�The acceptance limit AL1.3.2 (increased grace period relative to existing facilities) of CR1.3.2 is met if evidence is available to the INPRO assessor that the INS in case of DBA has a grace period of at least 8 hours.

�Final assessment of criterion CR1.3.2 (grace period)

�The period of non-interference in the cooling of installation due to the natural circulation of primary and secondary circuits, as well as in the passive heat removal through steam generators (SG PHRS) is 72 hours (the duration of the period provided in the supply of water tanks (PHRS)).

�The reactor SVBR-100 complies with the criterion CR 1.3.2

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Criterion CR1.3.3 (safety features) � Indicator IN1.3.3: Reliability of engineered safety features.

� Acceptance limit AL1.3.2: Equal or superior to existing designs.

� The acceptance limit AL1.3.3 (sufficient reliability of engineered safety features) of CR1.3.3 is met if evidence is available to the INPRO assessor that the INS in case of a DBA shows equal or higher reliability than existing designs.

� Final assessment of criterion CR1.3.3 (safety features)

� Reliable operation of safety systems (CR, ACR, PHRS) is provided due to of their passive operation without operator.

� Requirements of criterion CR1.3.3 (safety features) were fulfilled for reactor SVBR 100

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Criterion CR1.3.4 (barriers)

� Indicator IN1.3.4: Number of confinement barriers maintained.

�Acceptance limit AL1.3.2: At least one.

�Final assessment of criterion CR1.3.4 (barriers)

�Due to the impossibility of boiling coolant and the presence of the guard monoblock excluded loss of coolant accident. Small reactivity margin to burnout and negative feedbacks to power can reduce the risk of an accident with reactivity on prompt neutrons. Availability of passive heat removal system ensures interference period of not less than 72 hours. Thus, the probability of violation of safety barriers - the fuel matrix, fuel cladding and reactor pressure vessel is extremely small.

�Requirements of criterion CR1.3.4 (safety features) were fulfilled for reactor SVBR 100

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Criterion CR1.3.5 (controlled state)

� Indicator IN1.3.5: Capability of the engineered safety features to restore the INS to a controlled state (without operator actions).

� The acceptance limit AL1.3.5 (sufficient capability of engineered safety features)

� Final assessment of criterion CR1.3.5 (controlled state)

� The SVBR 100 has no active safety systems. Theeffectiveness of the emergency protection allows you tobring the SVBR-100 in a subcritical condition, taking intoaccount the temperature effect of reactivity. PHRS allowsremove heat from the core by the natural circulation ofthe coolant in the first and second circuit within 72 hours.

� We can conclude that requirements of criterion CR1.3.5were fulfilled for reactor SVBR 100

34

Criterion CR1.3.6 (sub criticality)

� Indicator IN1.3.6: Sub-criticality margins.

�Acceptance limit AL1.3.6: Sufficient to cover uncertainties and to allow an adequate grace period.

�The subcriticality of the SVBR-100 reactor is ensured in all shutdown states by keeping the control rods inserted into the reactor. The subcriticality of the reactor in shutdown states is monitored with neutron flux detectors.

�The reactor SVBR-100 complies with the criterion CR 1.3.6

�USER REQUIREMENT UR1.3 HAS BEEN MET BY THE SVBR-100 DESIGN.

35

Criterion CR1.4.1 (major release into the

containment)

� Indicator IN1.4.1: Calculated frequency of major release of radioactive materials into the containment / confinement.

� Acceptance limit AL1.4.1: At least an order of magnitude less than existing designs; even lower for installations at urban sites.


� The estimated probability of very severe damage SVBR 100 lower than that of nuclear submarines. More detailed information will be available after the completion of the technical project of RU

� We can suppose that requirements for criterion CR1.4.1 were fulfilled for reactor SVBR 100

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Criterion CR1.4.2 (processes)

� Indicator IN1.4.2: Natural or engineered processes sufficient for controlling relevant system parameters and activity levels in containment/confinement.


� Removal of residual energy is carried out by the natural circulation of the coolant in the first and second circuits and the contours of the PHRS system. In the event of a postulated failure of all four channels PHRS provides the control system beyond design basis accidents, which makes the Gulf water reactor shaft, followed by the removal of heat conduction is performed through the reactor vessel. In case of failure of both systems provided for the reactor shutdown (CR rods and PCR), provides for a system of rods ACR-activated locks after melting temperature at the outlet of the core equal to 700 ° C.

� Requirements for criterion CR1.4.2 were fulfilled for reactor SVBR 100

37

Criterion CR1.4.3 (accident management)

� Indicator IN1.4.3: In-plant severe accident management.

� Acceptance limit AL1.4.3: Procedures, equipment and training sufficient to prevent large release outside containment / confinement and regain control of the facility.

� The acceptance limit AL1.4.3 is met if evidence is available to the INPRO assessor that procedures, equipment and training are available, sufficient to prevent large releases to environment and regain control of the facility.


� Training will be conducted in accordance with the requirements of regulatory documents and international experience, including - using the simulator of SVBR 100

� Requirements for criterion CR1.4.3 were fulfilled for reactor SVBR 100.

� USER REQUIREMENT UR1.4 HAS BEEN MET BY THE SVBR-100 DESIGN.

38

Criterion CR1.5.1 (major release into

environment)

� Indicator IN1.5.1: Calculated frequency of a major release of radioactive materials into the environment.

�Acceptance limit AL1.5.1: Calculated frequency <10-6 per unit-year, or practically excluded by design.

�Final assessment of criterion CR1.5.1

�Project estimated frequency of significant release of radioactive for SVBR-100 significantly lower than on nuclear submarines. More precise information will be obtained after completion of the technical design SVBR-100.

�We can suppose that requirements for criterion CR1.5.1 were fulfilled for reactor SVBR 100

39

Criterion CR1.5.2 (consequences)

40

� Indicator IN1.5.2: Calculated consequences of releases (e.g., dose).

� Acceptance limit AL1.5.2: Consequences sufficiently low to avoid necessity

for evacuation. Appropriate off-site mitigation measures (e.g., temporary food

restrictions) are available.


� Even if the combination of such postulated initiating events as the

destruction of the protective shell, reinforced concrete slab above the

reactor and depressurization of the gas system of the first circuit. with

direct contact "mirror" of coolant in the body monoblock with

atmospheric air, blackout NPP - does not occur reactor runaway or

explosion or fire and the release of radioactivity into the environment

does not reach values that require evacuation of the population outside

the NPS site.

� The requirements for criterion CR1.5.2 were fulfilled for reactor SVBR

100

Criterion CR1.5.3 (risk)

� Indicator IN1.5.3: Calculated individual and collective risk.

�Acceptance limit AL1.5.3: Comparable to facilities26 used for a similar purpose.

� If nuclear energy is to play a major role in future, the calculated nuclear risk has to be comparable to the risk of other facilities used for the similar purpose.

�The acceptance limit AL1.5.3 is met if evidence is available to the INPRO assessor that the INS shows a risk comparable to other facilities used for similar purposes.

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A comparison of health risks of energy sources is difficult because use of fossilenergy or renewable energy sources cause damage mainly locally (within aregion) and for the present generation, while damage from radiation exposure ismore global and also affects future generations. Such a comparison can beperformed in terms of ‘years of life lost’ for a production of 1 TWˑh. Tablecontains these data for different types of facilities.

Table clearly demonstrates that for the same amount of electricity produced, therisk for personnel and population caused by nuclear power is very low.

Final assessment of criterion CR1.5.3 — risk criterion CR1.5.3 has been met by theSVBR-100 design.

TABLE ‘YEARS OF LIFE LOST’ PER TWH OF DIFFERENT ENERGY SOURCES

Final assessment of criterion CR1.5.3 (risk)

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USER REQUIREMENT UR1.5 HAS BEEN MET BY THE SVBR-100 DESIGN.

Criterion CR1.6.1 (independence of DID levels)

� Indicator IN1.6.1: Independence of different levels of DID.

� Acceptance limit AL1.6.1: Adequate independence is demonstrated, e.g., through deterministic and probabilistic means, hazards analysis etc.


� A preliminary study carried out independent levels of protection. The demonstration will be carried out after the completion of the safety analysis SVBR 100.

� We can suppose that requirements for criterion CR1.6.1 were fulfilled for reactor SVBR 100

� USER REQUIREMENT UR1.6 HAS BEEN MET BY THE SVBR-100 DESIGN.

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Criterion CR1.7.1 (human factors)

� Indicator IN1.7.1: Evidence that human factors (HF) are addressedsystematically in the plant life cycle.

� Acceptance limit AL1.7.1: Satisfactory results from assessment.

� The acceptance limit AL1.7.1 is met if evidence is available to theINPRO assessor that human factors are considered during thelifetime of a plant including the planning-, construction-,operating- and decommissioning phase.


� Safety systems are triggered independently of the operator'sactions. Thus, the human factor does not affect the safety SVBR100.

� The requirements for criterion CR1.7.1 were fulfilled for reactorSVBR 100

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Criterion CR1.7.2 (human response model)

� Indicator IN1.7.2: Application of formal human response models from other industries or development of nuclear-specific models.

� Acceptance limit AL1.7.2: Reduced likelihood of human error relative to existing plants, as predicted by HF models; use of artificial intelligence for early diagnosis and real-time operator aids; and less dependence on operator for normal operation and short-term accident management relative to existing plants.

� The requirements for criterion CR1.7.2 were fulfilled for reactor SVBR 100

� User requirement UR1.7 has been met by the SVBR-100 design.

� Final assessment of basic principle BP1 — defence in depth

� Enhanced defense in depth is incorporated into the SVBR-100 design, i.e. user requirements UR1.1 to UR1.7 are met.

Thus, basic principle BP1 is deemed to have been met by the SVBR-100 design.

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Conclusion

� For the SVBR-100 type RF the inherent self-protection properties

assure deterministically elimination of the certain severe

accidents, lack of the accidents, which are within the design basis,

and elimination of the catastrophic consequences of the

postulated accidents, which are beyond the design basis and

require the population evacuation beyond the NPP site.

� The SVBR-100 technology is based on forty-year experience (80

reactor-years of operating) of development and operation of LBC

cooled RFs at nuclear submarines (NS) and ground facilities-

prototypes.

� Within the SVBR-100 reactor design, clearly the defence-in-depth

concept has been enhanced (BP1).

46

2. INPRO basic principle BP2 (inherent safety)

�Basic principle BP2: Installations of an INS shall excel in safety and reliability by incorporating into their designs, when appropriate, increased emphasis on inherently safe characteristics and passive systems as a part of their fundamental safety approach.

�The basic principle BP2 contains 1 user requirement UR21

�UR2.1 (Minimization of hazards): INS should strive for elimination or minimization of some hazards relative to existing plants by incorporating inherently safe characteristics and/or passive systems, when appropriate.

47

Criterion CR2.1.1 (hazards)

� Indicator IN2.1.1: Sample parameters related to hazards: Stored energy, flammability, criticality, inventory of radioactive materials, available excess reactivity, and reactivity feedback.

� Acceptance limit AL2.1.1: Superior to existing designs.

� A list of possible evaluation parameters (EP) for this criterion is given below:

� EP2.1.1.1: Stored energy.

� In terms of stored energy in the primary coolant SVBR 100 vastly superior alternative projects of nuclear reactors, because the use of lead-bismuth coolant, which is under atmospheric pressure eliminates the possibility of steam and chemical explosions, fires and the release of hydrogen.

� EP2.1.1.2: Flammability.

� Used primary coolant and the working medium of the second circuit not burn in air, and do not interact with each other in coolant leakage

� EP2.1.1.3: Inventory of radioactive materials.

� The total amount of radioactive material is expected to less than at the existing nuclear power plants. It should be noted the practical absence of formation of liquid radioactive waste in operation SVBR-100, since the refueling is not required decontamination of the primary circuit.

48

Assessment of evaluation parameters (EP) cont.

� EP2.1.1.4: Criticality.

� Criticality out of the core under transport and technological operations with fresh and spent nuclear fuel is excluded. Including - in the assembly of the reactor - by eliminating flooding of the mine build.

� EP2.1.1.5: Available excess reactivity.

� Reactivity change for the campaign due to the properties of physics of fast reactor much less than VVER. When using MOX fuel reactivity change for the campaign can be further reduced by several times.

� EP2.1.1.6: Reactivity feed back.

� The values of reactivity effects from a change in temperature of the core elements of the SVBRR-100 meet the requirements of regulatory documents (all negative), the efficiency of systems for the reactor shutdown is sufficient to compensate for the effects of the reactivity of the translation of the reactor in a subcritical state. Void reactivityeffect is negative.

49

Final assessment of criterion CR2.1.1 hazards

�The acceptance limit AL2.1.1 is met if evidence is available to the INPRO assessor that the innovative reactor design is superior to existing designs for most of the evaluation parameters listed above.

�The assessment of evaluation parameters for criterion CR2.1.1 were fulfilled for reactor SVBR 100

�Final assessment of criterion CR 2.1.1:

� criterion CR 2.1.1 has been met by the SVBR-100 design.

50

Criterion CR2.1.2 (frequency of AOO and DBA)

� Indicator IN2.1.2: Expected frequency of abnormal operation and accidents.

� Acceptance limit AL2.1.2: Lower frequencies compared to existing facilities.

� The acceptance limit AL2.1.2 is met if there is evidence available to the INPRO assessor that due to the introduction or enhancement of inherent safety characteristics and use of passive safety systems (or components) lower frequencies of occurrence of AOO and DBA can be expected.

� Final assessment of criterion CR 2.1.2

� The frequency of deviations from normal operation is at a low level compared with the existing nuclear power plants thanks to the properties of the internal self-protection and passive principle of operation of safety systems.

� Final assessment of criterion CR 2.1.2:


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Criterion CR2.1.3 (consequences)

� Indicator IN2.1.3: Consequences of abnormal operation and accidents.

� Acceptance limit AL2.1.2: Lower consequences compared to existing facilities.

� The acceptance limit AL2.1.3 is met if there is evidence available to the INPRO assessor that due to the introduction or enhancement of inherent safety characteristics and use of passive safety systems (or components) the consequences of abnormal operation and accidents are lower than in existing designs.


� We can expect less severe effects of deviation from normal operation and design basis accidents due to the lack of steam explosions, exothermic reactions and heat transfer crisis in the first contour RU.



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Criterion CR2.1.4 (confidence in innovation)

� Indicator IN2.1.4: Confidence in innovative components and approaches.

� Acceptance limit AL2.1.4: Validity established.


� Confidence in innovative components was substantiated. The additional confidence will be evaluated in the licensing process of the project.

� Criterion CR 2.1.4 has been met by the SVBR-100 design.

� Final assessment of basic principle BP2 — inherent safety

� Basic principle BP2 requires incorporation of increased emphasis on inherently safe characteristics and passive systems into the reactor design.

� SVBR-100 design clearly meets the goal of BP2.

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4. INPRO basic principle BP3 (risk of radiation)

� Basic principle BP3: Installations of an INS shall ensure that the risk from radiation exposures to workers, the public and the environment during construction/commissioning, operation, and decommissioning, are comparable to the risk from other industrial facilities used for similar purposes.

� The basic principle BP3 contains 2 user requirement UR31-UR32

� UR3.1 (Dose to workers): INS installations should ensure an efficient implementation of the concept of optimization of radiation protection through the use of automation, remote maintenance and operational experience from existing designs.

� UR3.2 (Dose to public): Dose to an individual member of the public from an individual INS installation during normal operation should reflect an efficient implementation of the concept of optimization, and for increased flexibility in siting may be reduced below levels from existing facilities.

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Criterion CR3.1.1 (occupational dose)

� Indicator IN3.1.1: Occupational dose values.

� Acceptance limit AL3.1.1: Less than limits defined by national laws or international standards and so that the health hazard to workers is comparable to that from an industry used for a similar purpose.

� The acceptance limit AL3.1.1 is met if evidence is available to the INPRO assessor that doses to workers in an INS are less than defined by national or international standards and that health hazards to workers in an INS are comparable to that from other energy converting plants.


� Doses for personnel meet the regulations and is expected to be no worse than for operating plants.


� Сriterion CR 3.1.1 has been met by the SVBR-100 design.

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Criterion CR3.2.1 (public dose)

� Indicator IN3.2.1: Public dose values.

� Acceptance limit AL3.2.1: Less than limits defined by national laws or international standards and so that the health hazard to the public is comparable to that from an industry used for a similar purpose.


� Doses for population meet the regulations and is expected to be no worse than for operating plants.



� Final assessment of basic principle BP3 — risk of radiation

� The radiation protection system was optimized for the SVBR-100 design. Occupational dose (UR3.1) at the SVBR-100 meet the regulations and is expected to be no worse than for operating plants. The issue of dose to the public (UR3.2) meet the regulations and is expected to be no worse than for operating plants.

� Thus, the goal of basic principle BP3 has been met by the SVBR-100 design.

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5. INPRO basic principle BP4 (RD&D)

�Basic principle BP4: The development of INS shall include associated research, development and demonstration (RD&D) work to bring the knowledge of plant characteristics and the capability of analytical methods used for design and safety assessment to at least the same confidence level as for existing plants. The basic principle BP2 contains 4 user requirement UR4.1-UR4.4

�UR4.1 (safety basis): The safety basis of INS installations should be confidently established prior to commercial deployment.

�UR4.2 (RD&D for understanding): Research, Development and Demonstration on the reliability of components and systems, including passive systems and inherent safety characteristics, should be performed to achieve a thorough understanding of all relevant physical and engineering phenomena required to support the safety assessment.

�UR4.3 (pilot facility): A reduced-scale pilot plant or large-scale demonstration facility should be built for reactors and/or fuel cycle processes, which represent a major departure from existing operating experience.

�UR4.4 (safety analysis): For the safety analysis, both deterministic and probabilistic methods should be used, where feasible, to ensure that a thorough and sufficient safety assessment is made. As the technology matures, “Best Estimate (plus Uncertainty Analysis)” approaches are useful to determine the real hazard, especially for limiting severe accidents.

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Criterion CR4.1.1 (safety concept)

� Indicator IN4.1.1: Safety concept defined?

� Acceptance limit AL4.1.1: Yes.

� The acceptance limit AL4.1.1 for an INS is met if evidence isavailable to the INPRO assessor confirming a consistent safetybasis that demonstrates the safety goals are met.


� The safety concept is based on the advanced properties of the internal self-protection SVBR-100 reactor and the passive principle of operation of safety systems. The details of safety concept presented in the materials of the project.



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Criterion CR4.1.2 (safety issues)

� Indicator IN4.1.2: Clear process for addressing safety issues?


� The acceptance limit AL4.1.2 for an INS is met if evidence is available tothe INPRO assessor that there are well-documented results of theprocess addressing the safety issues including sensitivity and uncertaintyanalyses and independent reviews.

� Final assessment of criterion

� Currently, the license for accommodation SVBR-100 in Dimitrovgrad,Ulyanovsk Oblast was got. Work is underway to obtain a license for theconstruction of nuclear power plants.



� Thus, user requirement UR4.1 has been met by the SVBR-100 design.

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Criterion CR4.2.1 (RD&D)

Indicator IN4.2.1: RD&D defined and performed and database developed?

Acceptance limit AL4.2.1: Yes.

The acceptance limit AL4.2.1 for an INS is met if evidence is available to the INPRO assessor that:

Measured data are available in the region of application;

and that It was demonstrated that all phenomena are understood, data uncertainties are quantified, and documented in reports.

Final assessment of criterion CR 4.2.1

It will be carried out in the framework of R & D in the development of the technical project of SVBR-100

Final assessment of criterion CR 4.2.1:

We can suppose that criterion CR 4.2.1 has been met by the SVBR-100 design.

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Criterion CR4.2.2 (computer codes)

� Indicator IN4.2.2: Computer codes or analytical methods developed and validated?


� The acceptance limit AL4.2.2 for an INS is met if evidence is available to the INPRO assessor that for computer codes used in design and analysis of innovative reactors:

� The region of code application is covered by their validation matrix including quantification of uncertainties and sensitivities;

� Independent reviews have been performed:

� And a complete documentation including detailed code manuals are available.


� Work on the development of computer codes conducted five decades. At this time, a set of computer codes developed and validated.


� The criterion CR 4.2.2 has been met by the SVBR-100 design.

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Criterion CR4.2.3 (scaling)

� Indicator IN4.2.3: Scaling understood and/or full-scale tests performed?


� The acceptance limits AL4.2.3 for an INS is met if evidence is available to the INPRO assessor showing that scaling considerations including uncertainty analyses have been performed and are well documented.


� The effects of scaling realized. Basically, experiments are performed on full-scale models.


The criterion CR 4.2.3 has been met by the SVBR-100 design.

Thus, user requirement UR4.2 has been met by the SVBR-100 design.

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Criterion CR4.3.1 (novelty)

� Indicator IN4.3.1: Degree of novelty of process.

� Acceptance limit AL4.3.1: In case of high degree of novelty: Facility specified, built, operated, and lessons learned documented. In case of low degree of novelty: Rationale provided for bypassing pilot plant.

� The acceptance limit AL4.3.1 for an INS is met if evidence is available to the INPRO assessor that the degree of novelty of new SSC has been identified and an appropriate RD&D program has been established.


� The degree of novelty in the SVBR-100 has been established, i.e. it is classified as an evolutionary design, and necessary R&D activities have been performed.


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Criterion CR4.3.2 (pilot facility)

� Indicator IN4.3.2: Level of adequacy of the pilot facility.

� Acceptance limit AL4.3.2: Results sufficient to be extrapolated.

� The acceptance limit AL4.3.2 for an INS is met if evidence is available to the INPRO assessor that a peer review about the adequacy to build and operate a pilot plant has been performed.


� SVBR-100 will be the pilot plant, which will be worked out this technology


� The criterion CR 4.3.2 has been met by the SVBR-100 design.

Thus, user requirement UR4.3 has been met by the SVBR-100 design.

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Criterion CR4.4.1 (risk informed approach)

� Indicator IN4.4.1: Use of a risk informed approach?

�Acceptance limit AL4.4.1: Yes.

�The acceptance limit AL4.4.1 for an INS is met if evidence is available to the INPRO assessor that a careful use of risk informed approaches based on proven data sets has been performed by the designer.

�Final assessment of criterion CR 4.4.1

At the stage of obtaining a license for construction and operation will be used the risk-informed approach

Final assessment of criterion CR 4.4.1:


65

Criterion CR4.4.2 (uncertainties)

� Indicator IN4.4.2: Uncertainties and sensitivities identified and appropriately dealt with?


�Final assessment of criterion� When developing the project of the active zone SVBR-100 was taken

into account the technological and computational uncertainty. The design of the reactor core and the control rods allow the reserves to compensate for the most unfavorable from the point of view of nuclear safety the combination of uncertainties. Later in the justification for the reactor plant SVBR-100 certainly will take into account the uncertainty of any nature affecting safety RU



User requirement UR4.4 has been met by the SVBR-100 design.

Thus, BP4 is deemed to have been met by the SVBR-100.

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Conclusion

� For the SVBR-100 type RF the inherent self-protection properties

assure deterministically elimination of the certain severe

accidents, lack of the accidents, which are within the design basis,

and elimination of the catastrophic consequences of the

postulated accidents, which are beyond the design basis and

require the population evacuation beyond the NPP site.

� The SVBR-100 technology is based on forty-year experience (80

reactor-years of operating) of development and operation of LBC

cooled RFs at nuclear submarines (NS) and ground facilities-

prototypes.

� Within the SVBR-100 reactor design, clearly the defence-in-depth

concept has been enhanced (BP1).

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Conclusion

� The SVBR-100 design has placed increased emphasis on

inherently safe characteristics and passive systems in its safety

approach (BP2).

� Radiation exposure to workers in the SVBR-100 reactor is below

the regulatory limits (BP3). Not sufficient data were available to

the assessor to compare the radiation risk in an SVBR-100 with

the risk in non-NPPs.

� Intensive R&D activities have been performed (BP4) to ensure

reliable behavior of all new engineered features in the SVBR-

100 design, such as the use of additional passive systems

especially for prevention and mitigation of severe accidents.

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Thank you very much

for your attention

Documents

Preliminary assessment of innovative reactor SVBR … · Preliminary assessment of innovative reactor SVBR-100 in ... compared with LBC cooled ship ... Parameters EP1.1.1.1-EP1.1.1.5