Operational Reactor Safety

22.091 /22.903

Professor Andrew C. Kadak Professor of the Practice

Safety Analysis Report and LOCA Lecture 10

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 1

Topics to be Covered

• Safety Analysis Report – C ontents

• Chapter 15

– T ransients and Accidents Analyzed

• Loss of Coolant Accident Example

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 2

The Hazard (some fission-product isotopes)

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 3

Decay Heat

10 -1

10 -2

10 -3

-4

10 -1

10 10 2

10 3

10 4

10 5

10 6

10 7

10 8

T ime After Shutdown(s)

1- hour 1-day 1-week 1-month 1-year

seconds

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Source: Todreas & Ka zimi, Vol. 1

.

CRITICAL SAFETY FUNCTIONS

KEEP FISSION PRODUCTS WITHIN THE FUEL

• Control Reactor Power Control reactivity additions Shutdown reliably

• Cool the Reactor and Spent Fuel Maintain coolant inventory Maintain coolant flow Maintain coolant heat sinks

KEEP RADIOACTIVE MATERIAL OUT OF THE BIOSPHERE

• M aintain Containment Integrity Prevent over-pressurization Prevent over-heating Prevent containment bypass

• Capture Material Within Contain ment Scrubbing

Deposition Chemical capture

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SH IEL D PERSONNE L FRO M RADIATIO N

Prof. Andrew C. Kadak, 2008

Department of Nuclea r S c ien ce & Engineering

Page 5

Emergency Safety Functions

Reactor Safety

Department of Nuclea r S c ien ce & Engineering Study, WASH-1400

Page 6

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PWR SYSTEMS USED TO PERFORM EMERGENCY FUNCTIONS

Emergency Coolant Injection

React o D r e S p a f r e t t m y ent of Nuclea r S c ien ce & Engineering Study, WASH-1400

PWR SYSTEMS USED TO PERFORM EMERGENCY FUNCTIONS

Emergency Coolant Recirculation

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Reactor Safety Study, WASH-1400

PWR SYSTEMS USED TO PERFORM EMERGENCY FUNCTIONS

Post Accident Radioactivity Removal

Reactor Safety Study, WASH-1400

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 9

PWR SYSTEMS USED TO PERFORM EMERGENCY FUNCTIONS

Post Accident Heat Removal

Reactor Safety Stud y D , W e p A a r S t H-1400

k, 2008

PWR Engineered Safety Systems

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BWR Early Engineered Safety Systems

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Siting Criteria (10 CFR 100)

• Consideration of:

 Characteristics of reactor design

 Population characteristics, exclusion area, low population zone, population center distance

 Assume a bounding fission product release based on a major accident

 Define an exclusion area of such size that an individual located at any point on its boundary for two hours immediately following the accident would not receive a total radiation dose to the whole body in excess of 25 rem (250 mSv) or a total radiation dose in excess of 300 rem (3000 mSv) to the thyroid from iodine exposure.

 Define a low popula tion zone of such size th at an individual located at any point on its outer boundary who is exposed to the radioactive cloud during the entire period of its passage would not receive a total radiation dose to the whol e body in excess of 25 rem (250 mS v) or a total radiation dose in excess of 300 rem (3000 mSv) to the thyroid from iodine exposure.

 A population center distance of at least 1.33 times the distance from the reactor to the outer boundary of the population center distance

 Seismology, meteorology, geology, hydrology.

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 13

General Design Criteria (10 CFR 50 Appendix A)

http://www.nrc.gov/reading-rm/doc- collections/cfr/part050/

• The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components importan t t o safety ; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated withou t undu e ris k t o the healt h an d safet y o f th e public .

• Six major categories:

 Overall requirements

 Protection by multiple fission product barriers

 Protection and reactivity control systems

 Fluid systems

 Reactor containment

 Fue l an d reactivit y contro l

Prof. Andrew C. Kadak, 2008

Department of Nuclea r S c ien ce & Engineering

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The Single-Failure Criterion

• “Fluid and electric systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions.”

• The intent is to achieve high reliability (probability of success) without quantifying it.

• Looking for the worst possible single failure leads to better syst e m understanding .

Prof. Andrew C. Kadak, 2008

Department of Nuclea r S c ien ce & Engineering

Page 15

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GDC 10 and 11

• Criterion 10--Reactor design . The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

• Criterion 11--Reactor inherent protection . The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characte ristics tends to compensate for a rapid increase in reactivity.

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GDC 35

• An ECCS must be designed to withstand the following postulated LOCA: a double-ended break of the largest reactor coolant line, the concurrent loss of offsite power, and a single failure of an active ECCS component in the worst possible place.

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Defense in Depth

“Defense-in-Depth is an element of the Nuclear Regulatory Commission’s safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused even t occurs at a nuclear facility.”

[Commission’s W h ite Paper, USNRC, 1999]

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DEFENSE-IN-DEPTH MULTILAYER PROTECTION FROM FISSION PRODUCTS

NURE G D / e C p R a r - t 6 m 0 4 e 2 n , t 1 o 9 f 9 N 4 u . clea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 19

DEFENSE-IN-DEPTH, SAFETY STRATEGIES

NURE G D / C e p R a - r 6 t 0 m 4 e 2 n , t 1 9 o 9 f 4 N . uclea r S c ien ce & Engineering

Page 20

Safety Analysis Report Contents

Chapter 1 I ntroduction and General Description of Plant Ch apt er 2 S ite Ch aract eristics

Chapter 3 D esign of Structures, Com ponents, Equipment, and System s Chapter 4 Reactor

Chapter 5 Reactor Coolant Systems and Connected Systems Ch apter 6 E ngineered Safety Features

Chapter 7 Instrum entation and Controls Ch apter 8 E lectric Power

Chapter 9 A uxiliary System s

Chapter 10 Steam and Power Conversion System Ch apter 11 Radio active Waste M a n a g e ment Chapter 12 Radiation Protection

Chapter 13 Conduct of Operations Ch apter 14 Initial Test Program

Ch apter 15 Accid ent An alysis Ch apter 16 Technical Specifications Ch apter 17 Qu ality Assurance

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. K N a d U a R k , E 2 G 0 0 / C 8 R-6042, 1994.

Page 21

Design Basis Accidents

• A DBA is a postulated accident that a facility is designed and built to withstand without exceeding the offsite exposure guidelines of the NRC’s siting regulation (10 CFR Part 100).

• Each DBA includes at least one signi ficant failure of a component. In general, failures beyond those consistent with the single-failure criterion are not required (unlike in PRAs).

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. K N a d U a R k E , G 2 / 0 C 0 R 8 -6042, USN RC, 1994.

Page 22

POSTULATED ACCIDENTS AND OCCURRENCES

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U.S. Atomic Energ y Commission, 1973.

REPRESENTATIVE INITIATING EVENTS

TO BE ANALYZED IN SECTION 15.X.X OF THE SAR

1. I n c r e a s e i n H e a t R e mova l b y th e S e c o nd a r y System

1 . 1 F eed w a t e r s y s t e m m a l f u n c t io n s t h at r e s u lt s in a d e c r ea s e in f e e d w a te r te m p e r a t u r e .

1 . 2 F eed w a t e r s y s t e m m a l f unc t ion s t hat r e s u lt in an inc r e a s e i n f e e dw a te r f l o w .

1. 3 S t e a m pre s sure r e g u l a t o r m a lf u n c t i on or fa i l u re t h a t resu l t s i n i n c r ea s i ng st e a m fl o w .

1 . 4 I nadv e r ten t open i n g o f a s t e a m gene r a to r r e li e f o r sa f e ty valv e .

1 . 5 S pec t r um o f s t e a m s y s t e m p i ping f a ilu r e s in s i de and out s ide o f c onta i nmen t in a P W R .

2 . D e c r ea s e i n He a t R e m ova l b y t h e S e c o n d a r y S y s t e m

2. 1 S t e a m pre s sure s r e g u l a t o r m a lfunc t i o n o r fai l u re t h a t re s u l t s i n d e cr e a s i n g s t e a m fl o w.

2 . 2 L o s s o f ext e r nal el e c t r i c l oad.

2. 3 T u r bi n e tr i p ( s t op v a l v e c l o s ure).

2 . 4 I nadv e r ten t c l osu r e o f ma i n s t e a m i s olat i o n v a lve s .

2 . 5 L o s s o f conden s e r vacuu m .

2. 6 C o i n c i d e n t l o ss o f o n s i t e a n d e x t er n a l ( o ffs i t e ) a . c . p o w e r t o t h e st a t i o n .

2 . 7 L os s of n o r m al f e ed wat er fl ow.

2 . 8 F eed w a t e r p i ping b r eak .

3. De cr e a s e i n Re a c t o r Co o l a n t Sy st e m Fl o w Rat e

3 . 1 S ingl e and mu l tip l e r e ac t o r coo l ant pump t r ip s .

3. 2 B W R r e cir c u l a t i o n l o o p c o n t rol l er m a lf u n c t i o n s t h a t re s u l t i n de cre asi n g f l o w ra t e .

3 . 3 Re a cto r coo l ant pump sh a f t s e i z u r e .

3 . 4 Re a cto r cool a n t p u m p s h a f t b r e a k .

Prof. Andrew C. Kadak, 2008

Department of Nuclea r S c ien ce & Engineering

Page 24

NUREG/CR-6042, USN RC, 1994.

REPRESENTATIVE INITIATING EVENTS

TO BE ANALYZED IN SECTION 15.X.X OF THE SAR (cont.)

4 . Re a c ti v i t y a n d P o we r Di str i b uti o n A n oma li e s

4.1 U ncont r o l l ed cont r o l r o d a s se m bly w i thd r a w s f r o m a s ubc r i t i c a l o r lo w po w e r st a r tup cond i tion ( a s s u m ing the m o s t u n f avo r abl e r e act i v ity cond i t ion s o f th e co r e and r e a c to r cool a n t s ys t em ) , i nclud i n g c on t r ol r od o r te m po r a r y c o nt r ol devi c e r e m o va l e r r o r d u r i ng r e f u e l ing.

4.2 U n cont r o l l ed cont r o l r o d a s se m b ly w i thd r a w s a t t h e pa r t i c ula r p o w e r l e ve l ( a s s um ing th e m o s t u n f a vo r a ble r e a c tiv i t y c ondit i o ns o f t he c o r e and r e ac t o r coo l ant sy s t e m ) tha t yi e l d s t he m os t se v e re re su l t s (l ow p o wer t o f u l l p o w er) .

4.3 Cont r o l r o d m a lope r a tion ( sy s t e m m a l f u n c tion o r op e r ato r e r r o r ) , in c l uding m a lope r a tion o f pa r t l e ngth cont r o l r o d s .

4.4 A m a l f unc t ion o r f a i l u r e o f t h e f l ow cont r o l le r in BWR loop tha t r e s u l t s in an inco r r e c t te m p e r a t u r e.

4.5 A m a l f unc t ion o r f a i l u r e o f t he f l ow cont r o l le r in BWR loop tha t r e s ul t s in an inc r e a se d rea c t o r c o o l a n t fl o w rat e .

4.6 C h e m i c a l a n d vol u m e c o n t rol sy ste m m a l f u n c t i o n t h a t re su l t s i n a de cre ase i n t h e boron con c ent r a t ion in the r eac t o r c oolan t o f a P WR .

4 . 7 I nadv e r ten t lo a d ing a n d o p e r a t i o n o f a f u e l a s s e m b ly in a n im p r ope r pos i tion .

4 . 8 S pe c t r u m o f r od e j e c tion ac c i den t s in a P W R .

4 . 9 S pe c t r u m o f r od d r o p ac c i den t s in a BWR .

5. In c r e a s e i n Re a c t o r Co o l a n t I nv e nto r y

5 . 1 I nadv e r ten t ope r a tion o f ECC S d u r i ng po w e r ope r a t i on s .

5.2 C h e m i c a l a nd vol u m e c o n t rol s y s t e m m a l f u n c t i o n (or o p e rat or e r ror) t h a t i n c r ea s e s re a c t o r coo l ant inven t o r y

5.3 A n u m b e r o f B W R tra n s ie n t s , i n c l u d i n g i t e m s 2. 1 t h rough 2 . 6 a n d i te m 1. 2 .

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 25

NUREG/CR-6042, USN RC, 1994.

REPRESENTATIVE INITIATING EVENTS

TO BE ANALYZED IN SECTION 15.X.X OF THE SAR (cont.)

6. De cr e a s e i n Re a c to r C o o l a n t In ve n t or y

6. 1 I n a d v er t e n t o p e n i ng of a pre ss uri zer safet y or r e li ef v a l v e i n a PW R or a s af e t y or re l i ef v a lv e in a B W R .

6 . 2 B r e a k in in s t r u m ent lin e o r othe r lin e s f r om r ea c to r c oolan t p r e s s u r e bounda r y t h at pen e t r a t e c onta i nmen t .

6 . 3 S te a m g e ne r a t o r tub e f a ilu r e .

6 . 4 S pec t r um o f BWR st eam s y s t em piping f a i lu r e s ou t s i d e of con t ain m ent .

6 . 5 L o s s - o f - cool a n t a c c i dent s r e su l t ing f r om the sp e c t r u m o f pos t u la t e d p i ping b r eak s w i t h in th e r e a cto r cool a n t p r e s su r e bounda r y , i n clud i n g s t e a m line b r e a ks in s i de o f c o nta i nmen t in a B W R .

6 . 6 A nu m b e r o f BWR t r an s i e n t s , inc l uding it e m s 2 . 7, 2.8 , and 1.3 .

7. Ra di o a c ti v e Re le a s e fr o m a Sub s y ste m o r Compon e n t

7.1 R a d i o a c t i v e g a s wa st e s y ste m l e ak or fa ilure .

7 . 2 R a d io a c ti v e l i q u id wa st e sy ste m l e ak o r fa ilu r e .

7 . 3 P os t u la t ed r a dioa c tiv e r e l ea s e s due t o l i quid tank f ai l u r e s .

7 . 4 D e s ign b a si s f u e l hand l i ng acc i d ent s in t h e conta i nmen t a n d s p ent f u e l s t o r ag e bui l d ing s .

7 . 5 S pent f u el c a sk d r op a c cid e nt s .

Department of Nuclea r S c ien ce & Engineering

Prof. Andrew C. Kadak, 2008 Page 26

NUREG/CR-6042, USN RC, 1994.

Emergency Core Cooling System (ECCS) (January 1974, 10 CFR 50.46)

• Postulate several LOCAs of different sizes and locations to provide assurance that the most severe LOCAs are considered.

• Postulate concurrent loss of offsite or onsite power and the most damaging single failure of ECCS equipment (GDC 35).

• Acceptance Criteria

 Peak cladding temperature cannot exceed 2200 ºF (1204 ºC)

 Oxidation cannot exceed 17% of cladding thickness

 Hydrogen generation from hot cladding-steam interaction cannot exceed 1% of its potential

 Core geometry must be coolable

 Long-term cooling must be provided

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Double Ended Guillotine Break

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Figures © Hemisphere. All rights reserved.

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Department of Nuclea r

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2500

2400 "

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Safety Valv e s Open

High Pressure Trip - Both Poxer

Operded Relief Vatves Open

2350

Both

Spr a g y r i V g alVe S Full Open

Above

2350 psia

High

Pres sure

Both SpraY Valves Full

Closed

2300 ,

2275 ;'

Belo \ ’/ 2300 psia

Proportional Hea \ er Grou p " 0Fr

2250 ”

2225

22@ -”

tontrnl Set P0lflf

All Batkug Heaters ” 0fF" ét<ve 2223

Proportional

y l l Backup Heaters "Ok’" 8elow 22@

Low Pressure Alarn

psia

Thermal Marain/Loa Pressure Trip

fi-°! .^1int \ Y'• h Varv BP‹hveen

1730 and ZSOS psia’

1750

Thermal h \ arain/Low MininuwV»fue

P ress u T r r e ip

LP.':-LR Pressure Alarm and Satet

Injection Actuation Si¿nal

Pressure Control PrO§fan

7- 13

NEP 1 & 2

TABI• E 15 . 4- 1 c

LAR GE BREAK

TIM E SEQUENC E O G EVENTS

Amendment Nl0 March 1978

* From Containment Pressure Signal.

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NEP 1 & 2

TABL E l5.4-1d

Amendment N10

March 1978

Net Free Volume

Initial Conditions Pressure Temperature RWST temperature

CONTAINNEN Z DAT A — hE P CONTAINMEN T

2.987 x 10’ f 3

14. 7 psla 90°F

4O°F

Raw water temperature Outside temperature Relative humidity

Spray System

Number of pumps operating Runout flow rate ( totaL ) Actuation time

NA

0°F

99$

2

6600 u

35 sec

Structural Eeat Sinks

Item

Thicknes s ( ft )

Area ft 2

Containment Cylinder Containment Dome Containment Floor Containment Sump Miscellaneous Concrete Hiscellaneous Steel

.0313 steel, 4.5 concrete 70151

.0417 steel, 3 5 concrete 33867

4.0 concrete, .0208 steel, 9.0 concrete 13820

9.0 coocrete 972

1.0 concrete 160,000

0.2 steel 5,000

0.05 steel 65,000

0.03125 steel 90,000

0.030 steel 100,000

0.020 steel 70,000

0.0057 steel 45,000

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NEP 1 6 2

TABL E 1 3 . 4— 1 a

LARG E BREAK

Amendment N10 March 1978

DECL

0. 6 DCCL

0. 4 DECL

0.8 DECI.

Peak Clad Temp. °F Peak Clad Locaclon Ft- .

Local Z • ' 2 0 Rxn ( »a= ) z

Local Z •° 2° Location Ft.

2148

7.5

6.7

7.5

2137

7.5

6. 7

7.5

1790

9.0

2. 1

8.0

2144

7.3

6.8

6.0

Total Z 2 0 Rxn Z

*0.3 *0.3

<O.3

<0.3

Hot Rod Burst Time sec

21.0

23.2

84.8

20.2

Hot Rod Burst Location Ft. 5.75

NSSS Pou’e:c J ct 102Z oC

Peak I.1neac Power kw/it 102Z o€

Peaking Factor ( At License Rating )

3. 75

3425

12.6

2.32

7.O

Accuaulatoz Vatez Vo1uzae ( Cub1c J eeL ) 950

Puel region t cycle soslysed

WIT 1

Cycle

1

Rnglon

All

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( 0d/d ) ¥3k0d

AMENDMENT NIJ MARCH 197i

N EW ENGLAND POWEft COMPANY N E P & 2

Preliminary Safety Analysis Fleport

CORE POWER TRANSIENT - DECLG ( C - 1.0 )

A

FIG. 15.4.1-14A •• = = ° ^

( 0’l ’O 59O3O - 3lVb AO98 >V@b8

Z 6 \ d 3 N

AN'o’dNOO tJ3/ \ / \ Od ON \ r1ON3 M3N

8Z6£ HOtJYN

0tN 1N1WONZWV

BREAK FLOW ( LB/SEC )

( ‹ I Sd ) 3¥flSS3 \ Id

AMENDMENT N1o MAICH 1978

NEW ENGLAND POWER COMPANY N E P & 2

PreliIT \ inSry Safety Analysis Report

CORE PRESSURE — DECLG ( C D 1.0 )

FIG. 15.4.1-4A

( 1‹ ) 1353i k31Y/4

AMENDMENT N1f

MARCH 197t

NEW ENGLAND POWER COMPANY

N E P 1 & 2

Prel imii›ary Sule‹y An a vi is Fleport

REFLOOD TRANSIENT - DECLG ( C D - 1.0} DOWNCOMER AND CORE WATER LEVELS

FIG. 15.4.1-UA • • •

( 0"t ’3 ) DOOZO — 3dFlJ-Yu3dN3A O'b’13 1YZd

taodag stsAjeug Atajeg ‹ •^ u ! ^ !I ^* d

g 9 I d 3 N

ANVdINO3 UZMOd ON'V’0DN1 MZN

CLAD A VERA G T E EMPERATURE HOT g0O ( “F )

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›

2500

2000

@l500

TY/0 PHASE VOLUME

42 in. DOUBLE-ENDED HOT LEG BREAK ¥/ITH

MINIMUM SAFETY INJECTION

4000

3000

<<* I000

PRESSURE

Ll§UlD-

VOLUME

TIME, SECONBS

T0P0FC0R2 2000

BOHOM OF CORE

30

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0 20 40 60 80 100 120 140 160 180 2oo TlHE AFTER BREAS - SEC08DS

Figure 15. 4- 6, Double Snded Cold Leg Break ( Guillotine )

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15 . 4

NEP 1 6 2

CONDITIO N t Y - LIMITIN G FAULTS

Amendment NlO

Harch 1978

Refer to RESAR-3 ( 4-loop, without loop stop valves ) Section 15.4, with

the following modifications:

15 . k • 1 Ha jor Reac tor Coo l ane S s tern PI e Ru tures

Los s o I Co o lan d Acc ident )

The analysis specified by 10 CIR Part 50.46 ( 1 ) Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors is presented in this section. The results of the LOPA analyses

are shown in Table 15.4-1a and show compliance with the Acceptance Criteria.

The analytical techniques used are in compliance with Appendix K of 10 CFR Part 50, and are described in Reference ( 2 ) The results for the small break LOCA are presented in subsection 15.3.1 of the PSAR and are in conformance with 10 CFR Part 50.46 and Appendix K of 10 CFR Part 50.

The boundary considered for the IDCA as related to connecting piping is

defined in RESAR-3, Section 3.6.

Should a major break occur, depressurization of the reactor coolant system results in a pressure decrease in the pressurizer. Reactor trip signal occurs when the pressurizer low-pressure trip set point is reached. A safety injection system signal is actuated wLen the appropriate set point is reached. These countermeasures will limit the consequences of the accident in two ways:

a. Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

b. Injection of borated water provides heat transfer from the core

and prevents excessive clad temperatures.

At the beginning of the blowdown phase, the entire reactor coolant system contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boilin g is calculated, consistent with Appendix K of 10 CFR Part 50. Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms. During the refill period rod-to-rod radiation is the only heat transfer mechanism.

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