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Chapler 10 Quality Assurance (QA)

and analyzing their mutual relationships based on facts collected.

  1. Developing, determining and implementing corrective actions

The corrective actions are developed, determined and implemented based on the causes of an occurrence.

In the decision of corrective actions, evaluation of the usefulness of cost-effectiveness, etc., for the corrective action measures considered is carried out to make appropriate actions about the effects of the nonconformities.

In addition, when corrective actions are planned, decided and carried out, reflections onto similar products (subsequent and existing units) and to similar services (subsequent and existing) are considered. This is called ” Horizontal expansion".

Besides the horizontal expansion stated above, the following actions are also carried out, as appropriate;

  1. Reflection onto quality assurance plans

  2. Reflection onto designs, procurement, manufacturing & installation, inspection & test, the instruction manuals of operation & maintenance, etc.

  3. Reflection onto facilities & equipment, etc.

  4. Reflection onto education & training

  1. Records of results of corrective actions taken and follow-up of corrective actions

To show that the corrective actions were carried out reliably, the results of the corrective actions taken are kept as records. It is preferable that these records are added to the nonconforming report or managed in an integrated fashion by clarifying the relationship with the nonconforming report

Furthermore, it is important to carry out the corrective actions reliably by confirming and following up the performance situations. It is also important to improve the effectiveness of the quality management system continuously by implementing the appropriate actions if improvements are found in the corrective actions.

  1. Preventive Action

To prevent the occurrence of potential nonconformities beforehand, it is important to

analyze the causes of nonconformities, etc., deeply, to expose the potential causes, and to determine the actions to eliminate them. This activity is generally called ” Preventive action”. Important items in preventive actions are the information collected for the potential nonconformities and for the identification of their causes, and the analyses of their causes based on facts. The information should include nonconforming examples that have occurred both domestically and internationally. Furthermore, regarding similar nonconformities and nonconformities with a tendency to recur, it is also necessary to perform the root cause analyses and to prevent the occurrence of the potential nonconformities beforehand

The preventive actions are appropriate to the effect of the potential problems. The necessity of the preventive actions is judged in consideration of the importance to nuclear safety. When the importance is low, simplification of the action and a policy of not performing the preventive actions are included as options.

When the preventive actions are judged as necessary as a result of the consideration of the effects of potential problems, it is necessary to establish the procedure manual which prescribes the requirements for the following items and to carry out the preventive actions reliably.

  1. Determining the potential nonconformities and their causes

Analyzing information, identifying the potential nonconformities and their causes

  1. Evaluating the need for actions to prevent the occurrence of nonconformities

Performing the evaluation based on the usefulness of the cost-effectiveness, etc., for implementation of corrective actions considered

  1. Determining and implementing actions needed

Deciding the specific method for preventive actions and carrying it out according to the procedure manual reliably

  1. Records of results of action taken

Recording preventive action taken

  1. Reviewing preventive action taken

Confirming the implementation status of all the above activities and following them up.

10-33

NSRA, Japan

Appendixes

Appendixes

Appendix 1: Chronology of Nuclear Power Plants

Appendix 2: Typical BWR Plant Specifications and Facilities

Appendix 3 : Typical PWR Plant Specifications and Facilities

Appendix 4 : History of Nuclear Technology in Japan and

Transition of Total Generating Capacity of Nuclear Power Plants

Appendix 5 : Items of Improvement and Standardization(I/S) Project

for Light Water Reactor -BWRs-

Appendix 6: Items of Improvement and Standardization (I/S) Project

for Light Water Reactor -PWRs-

Appendix 7 : Key Specifications of BWR, PWR, ABWR and APWR Plants

Appendix 8 : The Outline of International Nuclear Event Scale (INES)

NSRA, Japan

Appendix 1 Chronology of Nuclear Power Plants

NSRA, Japan

JAPC’: Japan Atomic Power Company JPDR™: Japan Power Demonstration Reactor

Appendixes

Appendix 2 Typical BWR Plant Specifications and Facilities

(1) Key specifications of BWRs

Type

Spec.

BWR-1

BWR-2

BWR-3

BWR-4

BWR-5

GE design

Japan improved

ABWR

Fuel type (Initial load)

6x6

7x7

7x7

7x7

8x8

8x8

8x8

Average core power density (kW/t)

31

34

41

51

51

51

44-51

Forced core cooling method

External loops (3-5 loops)

External loops (2 loops)

RPV Internal pumps

Pumps

Pumps and jet pumps

Pumps + 5-nozzle jet pumps

Coolant flow control

Motor-generator sets

Flow control valves

Motor­generator sets

Thyrister control

ECCS

2-CS

HPCI added

LPCI added

3-LPCI+HPCS+LPCS

3-LPCI +2-HPCF

+RCIC

Primary containment vessel type

Dry spherical

MARK-I

MARK-I / II

MARK-II

Improved & standardized MARK-I / JI

ABWR

Steel self-standing

Concrete / Steel

Steel self­standing |

Reinforced concrete

Note : CS (Core spray system), HPCI (High pressure core injection system), LPCI (Low pressure core injection System), HPCS (High pressure core spray system), LPCS (Low pressure core spray system), HPCF (High pressure core flooding system), RCIC (Reactor core isolation cooling system)

Design trend of BWR primary containment vessel in Japan

MARK J PCV (1970-1979) Tsuruga-1 (357 MWe)

Fukushima 1-1

(460 MWe)

Fukushima 1-2

(784 MWe)

Fukushima 1-5

(784 MWe)

Shimane-l

(460MWe)

MARK-I Modified

(1987- )

Shimane-2

(820 MWe)

MARK-H

(1978-1985)

Fukushima 1-6

(llOOMWfe)

Hamaoka-1

Hamaoka-3

Fukushima 2-1

(540 MWe)

(1100 MWe)

(1100 MWe)

Hamaoka-2

Hamaoka-4

Kashi wazaki

(840 MWe)

(1137 MWe)

Kariwa-1 (1100 MWe)

Onagawa2

Tokai-2

(825 MWe) Shika-1

(540 MWe)

(1100 MWe)

MARK-II Modified (1984- ) FukusJiima 2-2,3,4 (1100 MWe)

Kashi wazaki Kariwa-2,3,4,5

(1100 MWe)

ABWRRCCV (1996- ) Kashiwazaki Kariwa-6,7 (1356 MWe)

Hamaoka-5 (138OMWe)

NSRA, Japan

app. 4

Appendixes

Specifications of typical PCVs for BWR

PCV type

MARK-1

MARK Imodified

MARKU

MARK- Ilmodified

ABWR RCCV

Design features

  • Pressure suppression type

  • Steel Self­standing

  • Drywell: cylindrical top & spherical bottom

  • Suppression chamber: toroidal (torus)

Same as left

■ Pressure suppression type

•Steel self­standing

•Drywell: conical ■Suppression chamber: cylindrical

•Vertical vent tubes

Same as left

•Pressure suppression type

■ Reinforced concrete

  • Cylindrical

  • Horizontal vent pipes

Approx, size

Output

460 MWe

1100 MWe

1100 MWe

1100 MWe

1300 MWe

Diameter

11m

24 m

26 m

29 m

29 m

height

34 m

38 m

48 m

48 m

36 m

Volume

Drywell volume

4,240 m3

8,800 m3

5,700 m3

8,700 m3

7,400 m3

Suppression Chamber volume

3,160 m3

5,300 m3

4,100 m3

5,700 m3

6,000 m3

Pool water volume

2,980 m3

3,800 m3

3,400 m3

4,000 m3

3,600 m3

Max. operating pressure

Drywell

4.35 kg/cm2-g

4.35 kg/cm2-g

3.16 kg/cm2-g

3.16 kg/cm2-g

3.16 kg/cm2-g

Suppression chamber

4.35 kg/cm2-g

4.35 kg/cm2 ■ g

3.16 kg/cm2-g

3.16 kg/cm2-g

3.16 kg/cm2-g

app.~5

NSRA, Japan

(2) BWR emergency core cooling system

BWR—3 Outline Flow Chart of Emergency Core Cooling System

BWR— 4 Outline Flow Chart of Emergency Core Cooling System

Rogidunl Hom Removal

BWR— 5 Outline Flow Chart of Emergency Core Cooling System

ABWR Outline Flow Chart of Emergency Core Cooling System

Appendixes

(3) Evolution of BWR fuel major design specifications

Fuel type

7x7

Improved

7x7

8x8

New 8x8

New 8x8 With Zr liner

High burnup

8x8

Year

1971

1974

1978

1982

1986

1991

Burnup (Mwd/t)

21500

27500

27500

295000

33000

39500

for reload

(approx.)

(approx.)

(approx.)

(approx.)

(approx.)

(approx.)

Pellet

uo2

UO,

uo3

UO,

uo2

uo2

Dia. / Length (mm)

12.4/22

12.1 /12

10.6/11

10.3 /11

10.3 /11

10.4 /11

Gd usage

No

Yes

Yes

Yes

Yes

Yes

Cladding

Zry-2

Zry-2

RCA *2

Zry-2

Zry-2

Zry-2

Zry-2

SRA*1

RCA*2

RCA *2

RCA*2 Corrosion resistance w/o autoclaving

RCA*2 Corrosion resistance w/o autoclaving

Thickness (mm)

0.81, 0.90

0.94

0.86

0.86

0.86

0.86

No. of rods

49

49

63

62

62

60

Water rod

0

0

1

2 (large dia.)

2 (large dia.)

l(largedia.)

Inti, press, (atm.)

1

1

1

3

3

5

U blanket

No

No

No

No

Yes

Yes

Spacer

SUS, Inconel

Zry-4

Zry4

Zry-4

Zry-4

Zry-4

X-750

Inco.X-750

Inco.X-750

Inco.X-750

Inco.X-750

Inco.X-750

(wire grid)

(lattice)

(lattice)

(lattice)

Qattice)

(circ. cell)

Design target

Long fuel rod

PCI countermeasure

Lower thermal load (8x8 fuel Arrangement) Even power distribution (water rod)

More thermal margin Better Uranium economy

Higher performance

Better economy & higher performance

Note: * 1SRA stands for stress relief and annealed material *2 RCA stands for re-crystallized and annealed material

7x7

IOOOOOOO’ OOOOOOO ooooooo OOOOQQO OOOQOOO QQQOOOO OOOOOOO

8x8

OOOOOOOO OOOOOOOO OOOOOOOO

OOOOOOOO ooooedoo OOOOOOOO COQOOOOO 0OOOOOOOJ

• Narrow fuel rods

■ One water rod

Reduced LHGR