AES-2006 new design with VVER reactor and INPRO methodology · PDF fileDenis Kolchinsky Project Chief Engineer AES-2006 – new design with VVER reactor and INPRO methodology State

Denis Kolchinsky

Project Chief Engineer

AES-2006 – new design with

VVER reactor and INPRO

methodology

State Atomic Energy Corporation ROSATOM

Branch of Joint Stock Company «East-European leading scientific research and design

institute for energy technology»

Saint-Petersburg R&D Institute “Atomenergoproject” (SPbAEP)

19-22 November, 2013 INPRO Forum, IAEA, Vienna

Was found in 1929;

Localized in Saint-Petersburg, Russia;

Has designs of power plants at 19 countries among them Finland, Czech Republic, Bulgaria, China, Vietnam, Cuba etc;

Participation in designing of 117 Power Plants, including 18 NPP;

More than 1400 employees;

01.07.2013 had been consolidated with Joint Stock Company «East-European leading scientific research and design institute for energy technology» .

Saint-Petersburg R&D Institute Atomenergoproject” («SPbAEP») – the General Designer of AES-2006


What AES-2006 and MIR.1200 are?


AES-2006 – is an abbreviated name of an evolutionary NPP of VVER-1200 design which was developed on the basis of standard Russian design VVER-1000.

Currently this design is being under construction on several sites.

The reference plant design for AES-2006 is AES-91 design, two units of which type was put into operation in China in 2007.

MIR.1200 (Modernized International Reactor) is based on the design of AES-2006 with taking into account requirements of Bid Specification for Temelin-3,4

VVER Designs Evolution


Existing NPP

for comparison

INS design for

comparison

Main Principles and Approaches for Designing of AES-2006


Maximum using of well-proven solutions and technologies

Minimizing costs and timeframes by using of experience which was got during construction and putting into operation of reference NPP (design AES-91 in China in 2007)

Ensuring of required safety level (also for DEC) by:

optimal safety systems configuration based on active (generally for BDA) and passive (generally for BDBA) elements;

diversity and redundancy of safety functions;

BDBA management;

reduction of human factor influence.

Improvements of AES-2006 in comparison with V-320


Increased thermal power to 3200 MW;

Increased unit electrical output to 1197 MW (Tcw=25oC);

Increased power unit efficiency coefficient to 37 % (gross);

Improved steam parameters at the exit from the steam generator to 7.0 MPa;

Improvement of safety characteristics (FCD<10-6).

Main Technical and Economical Parameters (for one Unit AES-2006)


# Parameter

1 Installed nominal output per one power unit, Mwe 1197

2 Service life, years 60

3 Power plant efficiency, % (gross) 37. 0

4 Power plant efficiency, % (net) 34.5

5 House load consumption, % 7.0

6 Availability factor, % 92.0

7 Number of operating personnel (person/MW) 0.42

8 Fuel campaign duration (i.e. fuel life in the core) (years) 4

9 Design basis maximum fuel burn-up (average per fuel assembly)

(MWd/kgU)

60

10 Period of refueling, months 12 (18)

1 – Essential Cooling Water pump

2 – Heat exchangers of Intermediate

cooling circuit for important

consumers

3 – Intermediate circuit pump

4 – Exchanger of the spent fuel pool

5 – Low-pressure pump of the

emergency injection system

6 – High-pressure pump of the

emergency injection system

7 – Emergency feed water pump

8 – Storage tanks with high

concentration boric acid

9 – Emergency boration system pump

10 – Storage tanks with boric acid

solution

11 – Emergency boration system

pump

12 – Storage tank of chemical

reagents

13 – Chemical reagents supply pump

14 – Containment Spray system pump

15 – Filter

16 – Deaerator of the volume and

chemical control system

17 – Pump of the volume and

chemical control system

18 – Ventilation stack

19 – Controlled leaks pump

20 – Controlled leaks tank

21 – External containment

22 – Steam generator

23 – Special water treatment plant

24 – After-cooler

25 – Spent fuel pool

26 – Bubbler tank

27 – Regeneration heat exchanger of

the volume and chemical control

system

28 – Reactor

29 – Reactor coolant pump

30 – Molten Core Catcher

31 – Emergency Core Cooling System

Sump and RWST

32 – Alkalis emergency reserve tank

33 – MSIV, safety and relief valves

unit

34 – Containment

35 – Pressurizer

36 – Hydroaccumulators

37 – Passive cooling system tank

38 – Condenser of the containment

passive cooling system

39 – Spray systems collector

40 – Passive hydrogen recombinator

41 – High-pressure heaters

42 – Electric-driven auxiliary feed

water pump

43 – Deaerator

44 – Electric-driven feed water pump

45 – Condenser

46 – Low-pressure heaters

47 – Condensate pumps of the first

stage

48 – Unit demineralised plant

49 – Main condensate treatment

50 – Superheater

51 – Circulation water pumps

(cooling apparatuses)

52 – Service water pump

53 – Machine hall consumers

54 – Standby step down transformer

55 – Generator

56 – Low-pressure part of the turbine

57 – Intermediate-pressure part of the

turbine

58 – High-pressure part of the turbine

59 – Boost pump

60 – Condensate pumps for Unit

demiralisation plant

61 – Emergency feed water pump

62 – Demineralized water storage

tank


Principal Diagram

Plant Layout

Concentration of all building of Nuclear Island around the Reactor Building;

Minimization of radioactive influence on the staff;

Physical division of the buildings which contains safety components into safety trains by the fire protected walls;

Reduction the lengths of communications between buildings: pipes, cabling etc;

Turbine building arrangement excludes the possibility of reactor building damage by the missiles from turbine;

Capabilities of access control to Nuclear Island buildings;

Optimized systems layout for better workflow and reduction of capital and operating costs.

The area of the Site (for two units), including cooling towers and other hydro-technical constructions, is 90 hectare.

Reactor Building

Turbine Building

Safety Building

Control Building Steam Cell

Auxiliary Building Fuel Storage Building


Main Equipment


321

291

The main safety measures in

comparison with V-320


ЯППУ С РУ В-491СОСТОЯНИЕ ТОПЛИВА

ГАЗОНЕПЛОТНЫЕ ТВЭЛЫ-0.2%

ДЕФЕКТНЫЕ ТВЭЛЫ-0.02%

ТЕПЛОНОСИТЕЛЬ ПЕРВОГО КОНТУРА

АКТИВНОСТЬ,Бк/кг:

ПРОДУКТЫ ДЕЛЕНИЯ (ИРГ-64%,ИОДЫ- -14%)-2.0Е8

ПРОДУКТЫ КОРРОЗИИ -1.0Е4ТРИТИЙ -7.4Е6

ПРОТЕЧКАнеорганизованная

0.1 Т/ЧАС

АКТИВНОСТЬ ВОЗДУХА, Бк/м3ПРОДУКТЫ ДЕЛЕНИЯ -2.0Е6

(ИРГ- 94%,ИОДЫ- < 0.1%)

ПРОТЕЧКА ГЦН

4.8 Т/ЧАС

ПРОБООТБОРорганиз. протечки

0.45 Т/ЧАС

ВЫВОДТЕПЛОНОСИТЕЛЯ

1060 Т/ГОД

ПРОТЕЧКА ВОВТОРОЙ КОНТУР

1 КГ/ЧАС

КОНТУРОЧИСТКИ KBE

КОНТУР ОЧИСТКИ

КВА10ВВ001

КОНТУР КВВ

В ПЕРВЫЙ КОНТУР

КОНТУРKBF,KPF,KPK,JNK

КОНТУР ОЧИСТКИKPL-3

4.4Е3

ВТОРОЙ КОНТУР

АКТИВНОСТЬ ПАРА, Бк/кг:

ПРОДУКТЫ ДЕЛЕНИЯ

(ИРГ-87%,ИОДЫ-12%) -3.3Е1ПРОДУКТЫ КОРРОЗИИ -4.0Е-4ТРИТИЙ -1.3Е2

ПРОТЕЧКА

неорганизованная

100 Т/Ч

АКТИВНОСТЬ ВОЗДУХА,Бк/м3НИЖЕ ДОАнас

ВЫБРОС(ГБк/год)

2.6Е-1(ИРГ-87%)

ТРИТИЙ 1.2Е0

ЖИДКИЕ СРЕДЫ

ВОЗДУХ

ТЕХНОЛОГИЧЕСКИЕ СДУВКИ

ГРАНИЦА ЗКД

БАККТА10ВВ001


КВА10ВВ001

ПРОТЕЧКАнеорганизованная

30 КГ/ЧАС

АКТИВНОСТЬ ВОЗДУХА,Бк/м3ПРОДУКТЫ ДЕЛЕНИЯ - 8.6Е2

(ИРГ-90%,ИОДЫ- - 0.2%)


KLD-20

КОНТУР ОЧИСТКИKLD-10

КОНТУР ОЧИСТКИKLЕ-30


KPL-2

ОСТАНОВ БЛОКА

РА

БО

ТА

НА

МО

ЩН

ОС

ТИ

7.0Е21.4Е2

ТРИТИЙ-5.0Е14.0Е4

ТРИТИЙ-3.9Е3

4.6Е4(ИРГ-99%,ИОДЫ<0.1%

АЭРОЗОЛИ<0.1%)

ТРИТИЙ-3.9Е3

ГЕ

РМ

ЕТ

ИЧ

НЫ

Й Б

ОК

С О

СН

ОВ

НО

ГО

ОБ

ОР

УД

ОВ

АН

ИЯ

1Е

10

М3

/ГО

Д

ЗД

АН

ИЕ

ТУ

РБ

ИН

Ы

ВЫБРОС(ГБк/год)ВЫБРОС(ГБк/год)ВЫБРОС(ГБк/год)ВЫБРОС(ГБк/год)

ВЫБРОС(ГБк/год)

1.2

Е9

м3

/го

д

РК

ПАР НА ТУРБИНУ

6468.5 Т/Ч

ВО ВТОРОЙ КОНТУР

ОСТАНОВБЛОКА

ГЕ

РМ

ЕТ

ИЧ

НЫ

Й Б

ОК

С О

СН

ОВ

НО

ГО

ОБ

ОР

УД

ОВ

АН

ИЯ

БО

КС

Ы

ВС

ПО

МО

ГА

ТЕ

ЛЬ

НО

ГО

О

БО

РУ

ДО

ВА

НИ

Я

КОНТУР ОЧИСТКИKLA13

РК

РК РКРК РК

Redundancy and

physical separation

Low volume of radioactive

emission

External impacts

protection

Defense in

Depth and

Diversity

I&C for safety functions

Safety systems for

BDA and measures

for BDBA

management 4 versus

3 trains

Diversity

of all SF Core Catch.,

PHRS…

D-Cont,

Aircraft Crash,

DBE=0,25g

Independence

DiD levels

Diversity, CCF

Designs Features Comparison

V-320 AES-91 MIR-1200/AES-2006

3 safety trains 4 safety trains 4 safety trains

Single containment Double containment Double containment

- Engineering measures for

BDBA management (H2-

recombiners, core catcher,

etc.)

Engineering measures for

BDBA management (H2-

recombiners, core catcher,

etc.)

- Enhanced seismic stability Enhanced seismic stability

- - C-PHRS and PHRS-SG

(BDBA)

- - Independence from outer

power supply – 72 hours

- - Emergency Storage Water

Tank inside the Containment


1

2

3

4

5

67

Main features of the Design

Optimal combination of passive and active

heat removal systems (HRS) Conventional active HRS

Newest additional Passive HRS

via Steam Generators

from the Containment


Measures for severe accident

management

Main features of the Design

Hydrogen Removal System

Core Catcher


Design Basis Conditions (DBC) and Design Extension Conditions (DEC)


For each category of design conditions the acceptance criteria are stated and made

safety analyzes to justify them.

CCF,

EEI*

Severe

Accidents

*) CCF – some of Common Cause

Failure events

EEI – some of Extremely External

Impacts

In the deterministic safety analysis, as per the level of possible negative consequences and an occurrence probability, the list of Design Conditions is divided into the several categories

For previous designs there was

only three categories:

- Normal operation;

- Anticipated transients;

- Accidents.

Radiation Safety for Population at Accidents


DBA:

Dose incurred by population will not exceed the operational dose limit established for normal NPP operation;

Radius of emergency protection area doesn’t exceed 0.8 km around reactor (i.e. limited by Site boundary).

Severe accidents:

Evacuation of people closely living near the NPP are not required;

Radius of area where protection measures for population are planned doesn’t exceed 3 km.

Safety Plant Assurance from the Fukushima Point of View


The “stress test” analyze was fulfilled for Leningrad NPP-2. Full list of

external impacts (including: flooding, tsunami, tornado, etc) was

considered.

Within the framework of “Stress tests” an expert engineering evaluation of

seismic strength of inner containment was performed to determine threshold

seismic impact value.

There was made analyze of following initiating events:

loss of all electrical power supply sources, including station black out;

loss of the ultimate heat sink;

combination of both.

Seismicity Resistance Analyze


The evaluation is performed by means of the following methods:

Direct strength analysis of containment as per linear and spectra theory of seismic stability at step-by-step increasing of acceleration level;

Evaluation based on design experience and seismic fragility test of this type of construction is performed as per recommendations of EPRI-NP-6041.

Conclusion:

Taking the minimum value of acceleration from obtained by these two methods, threshold value of maximum acceleration on the ground level is 0,51 g.

Current stage of the Designs


8 units of AES-2006:

Leningrad NPP – 2 Units under construction, 2 – in the process for construction permit;

Baltic NPP - 1 Unit under construction, 1Unit has Site Decision Permit;

Belorussia NPP – 2 Units under construction;

2 units of AES-91 under construction in China;

2 units of MIR.1200 in bid process for Temelin 3 and 4 in Czech Republic;

1 unit of AES-2006 is being considered for construction in Finland (Feasibility Study);

other possible bid competitions and contracts.

Assessment Methodology

for Innovative Nuclear Energy Systems (general comments and notes)

Recently the final report of an INPRO assessment of the

planned nuclear energy system (AES-2006), performed by

Belarusian experts in 2009-2011, was published by IAEA.

This publication is a clear evidence that it is necessary for

designers and/or technology holders make support or assist

of potential future NPP owners in this activity. Because

whether assessor doesn’t have actual design information, he

can take it from not-reliable sources, than, based on it, made

not correct evaluation and could disserve to the vendor (or

technology holder).


The Positives Effects from Cooperation


The cooperation of technology holder and potential NPP owners in

INPRO assessment may have following positive effects for each

party:

Popularization of certain design;

Additional self estimation of designer;

Independence estimation implemented by owner’s and IAEA’s

experts;

To contribute more opening of nuclear technology for public;

Increasing of owners competence level and knowledge of the

design and nuclear technology as a whole;

Stimulation of healthy competition among technology holders.

Criteria for the Assessment

For consideration at the Forum there were chosen following

criteria:

2-nd presentation (Mr. Michael Bykov):

CR 1.1.1 Robustness;

CR 1.2.1 Inherent characteristics;

CR 1.2.3 Inertia;

3-rd presentation (Mr. Michael Bykov):

CR 1.2.2 Grace period;

CR 1.3.2 Grace period;

CR 1.3.3 Safety features;


4-th presentation (Mr. Michael Bykov):

CR 1.3.4 Barriers;

CR 1.3.5 Controlled state;

CR 1.3.6 Subcriticality

5-th presentation (Mr. Denis Kolchinsky):

CR 1.4.1 Major release into containment

CR 1.4.2 Process

CR 1.4.3 Accident management;

6-th presentation (Mr. Denis Kolchinsky):

CR 1.6.1 Independent of defense in depth.

Criteria for the Assessment

(continuation)


Thank you for the attention!


Documents

AES-2006 new design with VVER reactor and INPRO methodology · PDF fileDenis Kolchinsky Project Chief Engineer AES-2006 – new design with VVER reactor and INPRO methodology State