Generation IV Design Concepts 

GE Advanced Liquid Metal Reactor 

S-PRISM 

by 

C. Boardman 

GE Nuclear 

San Jose, CA 

ACRS Wort June 4-5. 2001 

* Topics 

* Incentive for developing S-PRISM 

"* Design and safety approach 

"* Design description and competitive potential 

"• Previous Licensing interactions 

"• Planned approach to Licensing S-PRISM 

"* What, if any, additional initiatives are needed?

A CRS Worfakop June4-5.2001

135

2 Boardman
United States Energy Resources 

.S-PRISM would provide 

600 2.85 TWy was used 238 the LLS. with a long term 

550 in e U.S. 1994 energy source without 

S500 the need for additional 

450 1,900. mining or enrichment 1:400 i "°="")operati -ons. 

C 350 

1,900. TWy 

from tails (w/o further mining) 

0- 250 193.1 + 224. TWy by processing spent L WR fuel •, 200 

200 + 14. TMy by mining U.S. Reserves (< 130S5kig) 

100 2,138. TWy from U.S. Reserves w Fast Reactor 

50 23.1 29.3 5 

coal oil gas U - U - Fast Reactor 

LWR 

Indigenous U. S. Resources 

Energ estimates for fossilffuels are based on "International Energy Outlook 1995", DOE/EZA-0484(95). 

The amount of depleted uranium in the US includes existing stockpile and that expected to result from 

enrichment of uranium to fuel existing LWRs operated over their 40-y design life. The amount of uranium 

available for LWR/Once Through is assumed to be the reasonably assured resource less than $130/kg in 

the US taken from the uranium "Red Book". 

ACRs Wor,•o,,cp .,w4.21)0) 3 Bomrdma,0 

Phased STime Rlative Waste Toxicity (LWR Spent Fuel) 

10~ 

103 2 

S10 Actinide containing L WR spentffue 

FIMON remains toxicfor millions of years S •_PROM lo ='I 

-0 NATUR~AL URNIMabiE 

= • Processing to remove thefission 

o 0 products (-3% of LWRspent fuel), 

" __.l •TMHD uranium (9591), and transuranics

YEARS

ACRS WorbhoP June 4-5.2200 4 Boardman

136

prior to disposal shortens the period 

that the "waste "remains toxic to 

less than 500 years. 

- The recovered U and TR U would 

then be used as fuel and burned.
0' Relative Decay Heat Loads ofLWR and LMR Spent Fuel

Decay Heat 

Decay Heat Load (Watts per kg IM) 

LWR S-PRISM 

Spent Fuel at 

Discharge 2.3 17.8

Normal Process 

Product After 

Processing Spent Fuel 

"* Pu from PUREX 

Process for L WR 

"* Pu + Actinides 

from PYRO 

Process

WeaDons Grade Pu-239

9.62

1.93

25.31

During all stages in the S-PPJSM fuel 

cycle the fissile material is in a highly 

radioactive state that always exceeds the 

"L WR spent fuel standard". 

Diversions 

would be extremely difficult.

5 Boardman AMR WorkhOP

0

Stage oftte Fue Cyde

t

Masterd Ba•tm

19

Tarn•i]• Bwrks

ij -s

Co-Located Fuel Cycle Facility I

Phase I 

These opportunities for 

proliferation are not 

required for S-PRISM. 

Phase 2 

All operations are 

performed within 

heavily shielded 

enclosures or hot cells 

at the S-PRISM site.

Vt L

iL -VtII [ I I

IL 

IL VL VL

M 

Vt 

Vt 

Vt 

Vt Vt 

V7

VL I L 

M Vt 

I1 V

-- PL L-L~- --- ..

Jime 4-5, 2001

137

Phase 3 

All operations are 

performed within heavily 

shielded and inerted 

hot cells at the co-located 

S-PRISM/IFR site. 

6 B.,d.ý

Pbase 3: 

Fýý,l brumn Coerafiow

Rea conmnR 

WMW CM&U=

VL

ACAS Worahop

I I 6 [ I i t I I I -

bý1 E 

4( 

=_

I
o Key Non -Proliferation Attributes of S-PRISM 

1.) The ability to create S-PRISM startup cores by processing 

spent L WR fuel at co-located Spent Fuel Recycle Facilities 

eliminates opportunity for diversion within: 

e Phase I (mining, milling, conversion, and uranium 

enrichment phases) since these processes are not required. 

and 

* Phase II and III (on-site remote processing of highly 

radioactive spent LWR and LMR fuel eliminates the 

transportation vulnerabilities associated with the shipment 

of Pu) 

2.) The fissile material is always in an intensely radioactive 

form. It is difficult to modify a heavily shielded facility designed 

for remote operation in an inert atmosphere without detection. 

3.) The co-located molten salt electro-refining system removes 

the uranium, Pu, and the minor actinides from the waste stream 

thereby avoiding the creation of a uranium/Pu mine at the 

repository. 

ACRS Wohop June 4-5.2001 7 Boardman 

* Incentive for Developing S-PRISM 

> Supports geological repository program: 

0 deployment of one new S-PRISMplant per yearfor 30 years would 

eliminate the 86, 000 metric tons of spent L WR fuel that will be 

discharged by the present fleet ofLWRs during their operating life. 

a reduces required repository volume by a factor of four tofifty 

M All spentfuel processing and waste conditioning operations would be 

paid for through the sale of electricity. 

a limits interim storage to 30years 

> Reduces environmental and diversion risks 

"* repository mission reduced from >> 10,000 to <500 years 

"* facilitates long term CO2 reduction 

"* resource conservation (fossil and uranium) 

"* allows Pu production and utilization to be balanced 

"* utilizes a highly diversion resistant reprocessing technology

June 4-5. 2001

ACAS Wo*T¢A•

8 Boardman

138
8, Topics 

* Incentive for developing S-PRISM 

* Design and safetv approach 

* Design description and competitive potential 

* Previous Licensing interactions 

* Planned approach to Licensing S-PRISM 

* What, if any, additional initiatives are needed? 

AeCAS W O W June 4-5. 2001 9 ardmon 

* S-PRISM Safety Approach 

Exploits Natural Phenomena and Intrinsic Characteristics 

- Low System Pressure 

- Large heat capacity 

- Natural circulation 

- Negative temperature coefficients of reactivity 

June 4-5. 2001 10 Brdma.

ACRS Worb*op 139
Key Features of S-PRISM 

" Compact pool-type reactor modules sized forfactory 

fabrication and an affordable full-scale prototype test for 

design certification 

"• Passive shutdown heat removal 

"* Passive accommodation ofA TWS events 

"* Passive post-accident containment cooling 

"* Nuclear safety-related envelope limited to the nuclear 

steam supply system located in the reactor building 

"* Horizontal seismic isolation of the complete NSSS 

"* Accommodation ofpostulated severe accidents such that a 

a formal public evacuation plan is not required 

"* Can achieve conversion ratio's less than or greater than one 

ACRS Worý June 4-5, 2001 11 Boardm, 

*S-PRISM Design Approach 

Simple Conservative Design 

* Pass'e decayheatremoval 

* Passive accommodatno afATWSEvents S-PRISM Features Contribute to: 

* Automated safetygrade acbons are limfedto., 

- C01nAainment# iSoibn . Simplicity of Operation 

- reactorscram * Reliability - steam side isolabbn andblow-down 

* Maintainability 

Operation and Matintenance 

* SafetygradeenvelopeconfinedatoNSSS Reduced L Risk of Investment 

* Simple compact pdmarysytem boundary Loss 

* Lowpersonnellradiationexposure/evels * Low Cost Commercialization 

Path 

Capita /and Investment Risk Reduction 

* Conservaffve Low Temperature Design 

SModular Construction andseimlc lsolabon 

* Factory fabifcation f/components and facidy modules 

• Modulalty reduces the need for spinnig reserve 

• Cerltication waprotoze testng ofa single 380 MWe module 

A CAS,, W 4 June4-5.2001 12 Boad=ý

140
* S-PRISM Design Approach (continued) 

1. Design basis events (DBEs) 

- Equipment and structures design and life basis 

- Bounding events that end with a reactor scram 

- Example, all rod run out to a reactor scram 

2. Accommodated anticipated transients without 

scram (A-A TWS) 

- In prior reactors, highest probability events that led to boiling 

and Hypothetical Core Disassembly Accidents were A TWS events 

- In S-PRISM, A TWS events are passively accommodated within 

ASME Level D damage limits, without boiling 

- Loss ofprimaryflow without scram (ULOF) 

- Loss of heat sink without scram (ULOHS) 

- Loss offlow and heat sink without scram (ULOF/LOHS) 

- All control rod run out to rod stops without scram (UTOP) 

- Safe shutdown earthquake without scram (USSE) 

3. Residual risk events 

- Very low probability events not normally used in design 

- In S-PRISM, residual events are used to assess performance 

margins 

ACRS WMOW June 4-5, 2001 13 Boa•-d',, 

* Topics 

"* Incentive for developing S-PRISM 

"* Design and safety approach 

• Design description and competitive potential 

• Previous Licensing interactions 

* Planned approach to Licensing S-PRISM 

* What, if any, additional initiatives are needed? 

AJune4-5, 2001 14 Boardmm. 

A CRS Wor~hop 141
QPower Train

Safety Grade

I 

I 

I Hiah Grade 

r Industrial Standards 

I Redundant

Safety Grade 

Isolation Valves

from cooling

.R.rAs . . - ..:. . cCondenser S 

Shutdown Heat Removal Systems 

jume 4-5, 2001

a S-PRISM - Principal Design Parameters

Reactor Module 

- Core Thermal Power, MWt 

- Primary Inlet/Outlet Temp., C 

- Secondary Inlet/Outlet Temp., C

Power Block 

- Number of Reactors Modules 

- Gross/Net Electrical, MWe 

- Type of Steam Generator 

- Turbine Type 

- Throttle Conditions, atg/C 

- Feedwater Temperature, C 

Overall Plant 

- Gross/Net Electrical, MWe 

- Gross/Net Cycle Efficiency, % 

- Number of Power Blocks 

- Plant Availability, % 93 

June 4-5, 2001 

ALPgS IWor~1op 142

2 

825/760 

Helical Coil 

TC-4F 3600 rpm 

171/468 

215 

2475/2280 

41.2/38.0 

3

16 Boardman

ACRS WorkOW 15 Boardman

1,000 

363/510

321/496
S__Super PRISM

SS-PRISM Power Block (760 MWe net)

Two 380 MWe N555 per Power Block

June 4-5, 2001

143

ACRS WorbhOP

June 4-5. 2001 17 Boardman

A CRS WarimOP*~~

18 Boardman
a Metal Core Layout

Number of Assemblies

SDriver Fuel 

Sinternal BLanket 

Radial Blanket 

© Gas Expansion Module 

O Reflector 

Shield 

Total 

Jum 4-5, 2001

ACES WorbhAP

138 Fuel: 23 month x 3 cycles 

49 

Blkt 23 month x 4 cycles 

48 

9 

3 

6 

126 

72 

451 

19 Bwardan

* Oxide vs. Metal Fuel 

Attractive features of metal core include: 

- fuel is denser and has a harder neutron spectrum 

- compatible with coolant, RBCB demonstrated at EBR-II 

- axial blankets are not required for break even core 

- high thermal conductivity (low fuel temp.) 

- lower Doppler and harder spectrum reduce the need for GEMs for 

ULOF (6 versus 18) 

* Metalfuelpyro-processing is diversion resistant, compact, 

less complex, and has fewer waste streams than conventional 

aqueous (PURENJ process 

However, an "advanced" aqueous process may be 

competitive and diversion resistant. 

S-PRISM can meet all requirements 

with either fuel type.

Jume 4-5, 2001

144 A CAS Worbhop

20 Badýnw
0, S-PRISM - Three Power Block Plot Plan

Three Power Block Plant 

24 75 MWe (2280 MWe net) 

S13 

1 Reactor Building (2 NSSS/block) 

1 2 Reactor Maintenance Facility 

u I 3 Control Facility 

4 New and Spent Fuel Handling Facility 

5 Assembly Facility 

6 Cask Storage Facility 

7 Turbine-Generator Facility 

p 8 Maintenance Facility 

9 Circulating Water Inlet Pump Station 

10 Circulating Water Discharge 

S11 Waste Treatment 

12 Parking Lot 

,I 13 Switch Yard 

14 Fuel Cycle Facility

21 Boo..dna

0S-PRISM - Seismic Isolation System

Characteristics of 

Seismic Isolation System 

Safe Shutdown Earthquake 

- Licensing Basis 0.3g (ZPA) 

- Design Requirement 0.5g

Lateral Displacement - atO.3g 

- SpaceAllowance 

"o Reactor Cavity 

"o Reactor Bldg. 

* Natural Frequency 

- Horizontal 

- Vertical

7.5 inch. 

20 inch. 

28 inch. 

0. 70 Hz 

21 Hz

- Lateral Load Reduction >3 

Rubber/Steel Shim Plates 

Protective Rubber Barrier

[-<s- (ft. ) 

Seismic Isolators (66)

22 Bordm,

A

I

A CRS wofkhop

Jue 4-5. 2001

A CRS WorbhOP Jne 4-5.2001

145
Reactor Vessel Auxiliaiy Cooling System (RVA,0S)

Flow Annuli & Silo 

Cross Section

96230

Passive Shutdown Heat Removal (R VA CS)

f" 

11* 

typ

iv

I2JrF 

sci 

Equb1mi 

e U axa Ne 299

Jwt4-5. 2001 

ACES Wob&•' 146

ELEVATION

ACkS W07bho

Silo Cavity

Jue 4.-. 2001

S

23 B.,d.

trsraasflflLA

24 Boardman
Natural Circulation Confirmed by 3 Dimensional T/H Analysis 

n 

- 4 , 

t i 

S• WExamples 

Temperature and velocity distribution 

Normal Operation at 4 and 20 minutes after loss of heat sink 

Jane 4-5, 2001 25 Boardman 

* Decay Heat Removal Analysis Model 

---0d -- . -.- • . 

Prfrnaty Coolant Loop Reactor RVACS Air Flow Circuit 

Vessel Containrnent 

Liner Vessel NOTE: Elevations Are Not Representec 

¶SJ 4-3. 2001 26 Bo-dnm 

ACRS WortI'op ••-

I14
SR VA CS Cooling - Nominal System Temperatures

t... 

'0 

Q 'C 

I.. 

0 50 100 150 20 

Time (hr)

?0 250 300

R VA CS Transients Are Slow Quas! Steady State Evens

.hme 4-5, 2001

aR VA CS Heat Rejection and Heat Load versus Time

203D350 40D 5D10 I 20D 

Tim. (hr)

Jme 4-5. 2001

148

Outlet 3Core 5emp 

Vse C 'iwafTep 

A 0tS WOr*hPp

27 Boardan

3

ACRS FforsheOP

28 Boardm'an

350 400
SR VA CS Cooling - Nominal Mixed Core Outlet Temperature

f~Ytn2IW Peak Cm. fMtie 0~t T-mP--tuI

700

I.. 

2..

6" 

500 

400

300 

200

50 100 M50 200 250 300 350 400 

Time (hr)

Jou 4-5. 2001 29 Bo.,d,

* Damage Fraction from Six R VA CS Transients

,.J•.. . ., ... . r.... a .

Sic up - J[L.j - 0 

[Damage from RVACS Transients Is NegligibleA

Avw 4-5. 2001

149

A CAS Wor'ksop

A CRS WarbhOP

30 Baa-dman
Q, S-PRISM Approach to A TWS 

Negative temperature coefficients of reactivity are 

used to accommodate A TWS events. 

- Loss of Normal Heat Sink 

- Loss of Forced Flow 

- Loss of Flow and Heat Sink 

- Transient Overpower w/o Scram 

These events have, in priorLMR designs, led to rapid 

coolant boiling, fuel melting, and core disassembly. 

S-PRISM Requirement: 

Accommodate the above subset of events w/o loss of reactor 

integrity or radiological release using passive or inherent natural 

processes. A loss offunctionality or component life-termination 

is acceptable. 

Ahme 4-5.2001 31 Boardman 

* ARIES-P Power Block Transient Model 

STEA F---- VALVE STEAM 

ILRW PRESSUR 

PUMP TURBINE 

TUAED W AT-R 

EFEEAWATR ARAT -HEATER 

• VC ST0te EAMBOTE UP 

* Two-Reactors Coupled to a Single TG * Once-through Superheat 

* One Group Prompt Jump Core Physics * Control Systems: 

with Multi-Group Decay Heat - Plant control system (global and local controllers) 

* R A CSA CS- Reactivity control system (RCS) M - Reactor protection system (RCIS) 

- EM pump control system and synchronous machined 

ACS O*hPJune 4-S. 2001 32 Soa•-a,, 

A•.S Wvlnho UMPS

$

I DUJ
* Example ATWS- Lossf Flow Without Scram 

S- Loss of Primary Pump Power w/o Scram 

Core Power Fraoonf .. " 

-Core low Fracton (') * Loss of pump pressure allows GEM 

feedback and fission shutdown 

i Continuation of iHTS flow and 

j" I feed water water enhance primary 

natural circulation to 10% 

* Excess cooling of core outlet 

shortens CR drivelines andpulls 

- control rods slightly to balance fissioa 

"power with heat removal 

LiM 

I _ a ....... .f 

- •=--=• r=---.= - I 1 i t± _ • = '

*Example - 0.5 g ZPA Seismic Event Without Scram

151
S-PRISM Transient Performance Conclusions

S-PRISM tolerates A TWS events within the 

safety performance limits 

The passive safety performance of S-PRISM 

is consistent with the earlier ALMR program 

ACAS Wmidiep June 4-5, 200 

* S-PRISM Containment System

t 

Mawinenance En losur Upper Containment 

for Reactor B 

IIJ2 

0ý L_7 .1 All 

S'ic' Cell 

Disk 

Upper Containment 

for Reactor A

.Jwe 4-5. 2001

152

35 Boardman

A CPS Woiihco' 36 Boa•-dm.

... __......._L.

...... ......

• o • Containment ecoUpper 
Example - Large Pool Fire

Beyond Design Basis (Residual Risk) 

events have been used to assess containment margins 

This event assumes that the reactor closure 

disappears at time zero initiating a large poolfire 

Note that the containment pressure peaks at less than 5psig 

and drops below atmospheric pressure in less than 6 hours 

JIM 4-5, 2001 37 oardm

of Emergency Power RequirementsSCompar

Fun don S-PRISM • 

"* Shutdown HeatRemoVa! CompletelyPassive 

"* PostAcddent Passive Air Cooling 

Containment Cooli7g of Upper Conta7nment 

"* Coolant Injecton/Core Flooding NIA 

"* Shrldown System 3,Y Pn'maly or 213 Secondary Rods 

SelfActuated Scram on Secondary Rods 

Passive Accommodation ofA 7WS Events

Generation /I/ L WRs 

Redundant and Diverse Systems 

Redundant and Diverse Systems 

Redundant and Diverse Systems 

Most Rods Must Function 

Boron lnjecton 

NIA

< 200 kWe from Batteries

Jwse 4-5, 2001

- 10, oo0kWe I

38 Boa'dian

153

ACRS WorfsiW

I lmprnonrvA Power

ACRS Workhop

I

I " -•c.I| .21 

-S I I. 

'0 

o ¶ 2 2 6 7 3 

Time (hours)

I EmegenrvAC Power t

L.
0 Layers of Defense

" Containment 

(passive post accident heat removal) 

" Coolant Boundary (Reactor Vessel 

(simple vessel with no penetrations below the Na level) 

" Passive Shutdown Heat Removal 

(R VA CS + A CS) 

"* Passive Core Shutdown 

(inherent negativefeedback's) Incro 

Cha. 

"* RPS Scram of Scram Rods 

(magnetic SelfActuaed Latch backs up FPS) 

"* RPS Scram of Control Rods 

MAPS is independent and close coupled) 

"* Automatic Power Run Back 

(by automated non safety grade Plant Control System

Normal Operating Range A 

Maintained by Fault Tolerant 

Tri-Redundant Control System

June 4-5, 2001

* Adjustments Since End of DOE Program In 1995 U 1 1

Parameter or Feature 1995 ALMR S-PRISM

Core Power, MWt 840. 1000. 

Core Outlet Temp, ,C 499 510 

Main Steam, "C / kg/cm2 454/153 468/177 

Net Electrical, MWe 1243. 1520 

(two power blocks) 

Net Electrical, MWe 1866 2280 

(three power blocks) 

Seismic Isolation Yes. Each NSSS Yes. A single 

placed on a platform supports 

separate isolated two NSSSs 

platform 

Above Reactor Containment Low leakage steel Low leakage steel 

machinery dome lined compartments 

above the reactor 

closure

June 4.5. 2001

154

k

easing 

Henge

AC.RS Workhop

39 Boýda,,.

A CRS WorkdW 40 B.rdmn
Topics

"* Incentive for developing S-PRISM 

"* Design and safety approach 

"• Design description and competitive potential 

"* Previous Licensing interactions 

"* Planned approach to Licensing S-PRISM 

"* What, ifany, additional initiatives are needed?

June 4-5. 2001

*Optimizing the Plant Size

1988 PRISM -i S-PRISM Large Commercial Design

1263 MWe (net)from 3 blocks 1,520 MIWe (net)from two blocks 1,535 MWe Monolithic LMR 

9 NSSS (425 MWt each) 4 NSSS(IOOMWt each) I NSSS (4000 MWt) 

3 421 MWe TG U-1ts 2 825 MWe (gross) TG Units 1 1535 MWe TG Unit 

9 primary Na containing vessels 4 primary Na containing vessels 14 primary Na containing vessels* 

9 SG units/eighteen 1=7loops 4 SG units and eight 1HTS loops (12primary componem vessels. reactor, and EVS7) 

(1000/500 MWt each) 6 SG units and 6 I=TSoops (667 MWt each) 

----- ............... ----- - 4 Shutdown Heat Removal Systems \ Larger module (1000 v. 425 MWt) (DIHX- unis, pzmp, piping, and support systems) 

Once through merheal steam cycle Redundant S-RS also required for EST

.June 4-5. 2001

155

ACRS Wotjwp

41 Board',wn

ACRS Workshop

42 Boadma
Scale Up - - L WR versus Fast Reactor 

1600 MWt Sodium Cooled Fast Reactof600 MW Light Water Cooled Reactor

Three 533 Mit.Loops

3600 MWt FR

Loops

Rating Limited by: 

IHTS Piping: < I m diameter

Two 800 MW Loops

3600 MWt PWR

Two f100 MW( Loops 

Two Loops Viable Because: 

Specific heat of water 5 x sodium 

at operating temperatures

The complexity and availability of a PWR is essentially constant with size 

Due to the lower specific heat of sodium, six or more loops are required in a large FR. 

The Economy of Scale is Much Larger for L WRs then FBRs

Jwie 4-5. 200)

SModular versus Monolithic (Fast Reactors)

To TG

Modular (S-PRISMF

Monolithic Fast Reactor

The one-on-one arrangement: 

* simplifies operation, 

• minimizes the size of the reactor building 

* improves theplant capacity factor 

* reduced the need for backup spinning reserve 

June 4-3. 2001 

156

AtCS Wosahop

-ii

43 B.rd.

ACPS WO,*-howP

44 B.ad.

I

I

0 Modular versus Monolithic (Fast Reactors) I
SNSSS Size, ALMR verses S-PRISM

22 % More Power 

from 

Smaller NI

Jlune 4.5. 2001

168 ft. - I 

SV RV 

123 RI 

SO

Seismically /Isolated

S-PRISM

45 Boardman

A CAS Wordi-p

* Learning Effect Favors Modular Plant Designs

'Ieooo o o 00 

Cumulative Plant Capacity, MWe 

.hme 4-5, 2001 

157

i ~210 O

I

ALMR

AMAS WorbhOp

46 Boardrc

ii - --.. j
SModular vs. Monolithic Availability and Spinning Reserve

S67, 

50 

S33 

17

0% 20% 40% 60% 30% 100% 

86%

Percent Time at Load (%)

6 Module S-PRISM Plant

S.... . - 01.31 7.$

% 97.9% 

% 99.3% 

% 99.95% 

% ge 99.99% 

0% 20% 40% 60% 80% 100% 

93 % 

Percent Time at Load (%)

June 4-5. 2001

*_Comparison ofPlant Construction Schedules

NOAKModular 

Simultaneous 

NOAK Modular 

Staggered

First Commercial Modular

Simultaneous

Firsta Commercial Modular 

Staggered

First Commercial Large 

Reactor

A CRS ork*Toh

EiIIIIII i

II IEhIIEIEIEI 11/

II IEI JhII

1,520 MWe 

S-PRISMPlant

Monolithic Plant - 1520 M"We

48 Bardman

158

Monolithic Plant 

6 Loops 

Six Loops

83% 

~67% • 85.80% 

5 7.0%

Six Modules 

9 83"%. Five Modules 

• ... " Four ModuleY1

Seven point advantage caused by: 

"* Relative simplicity of each NSSS (one SG System rather than 6) 

"* Ability to operate each NSSS independently of the others

ACAS W'm•:oh 47 B•adan

5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 

Duration, months

June 4-5. 2001

g,

: . .. Monolithic Plant- 1520 MWe
*NSSS Size, CRBRP/ALMR IS-PRISM 

CRBRP - ALMR 

S-PRISM 

00760 MWe 

A S 4-5, de 2001 49 Boardman 

ACRCS Wonmhop 

O Topics 

"* Incentive for developing S-PRISM 

* Design and safety approach 

* Design description and competitive potential 

* Previous licensing interactions 

* Planned approach to licensing S-PRISM 

* What, if any, additional initiatives are needed? 

ACRS WorbAhop A=4-S.200) 50 &

159
Q' ALMR Design and Licensing History

Funded

DOE Program j

GE Funded 

Imovative Design Studies

Jhne 4-5. 2001

S-PRISM is supported 

by a 100 million dollar 

Data BaseJ

SI Boardman

(NUREG-1368) concluded:

"the staff, with the ACRS in agreement, concludes that 

no obvious impediments to licensing the PRISM (ALMR) 

design have been identified." 

..he 4-5. 2001 

160

AcRS WMAO-P

A CRS Worbhop

52 SBordma,

51 Bmd--

The NRCS pre-application Safety Evaluation of the A LAII.R
*Topics

"• Incentive for developing S-PRISM 

"* Design and safety approach 

"* Design description and competitive potential 

"* Previous Licensing interactions 

"• Planned approach to Licensing S-PRISM 

"* What, if any, additional initiatives are needed?

June 4-5. 2001

0 Detailed Design, Construction, and Prototype Testing

Y_ _ 3 1 _ _1 2 3 5_ 1, 1 6 12 3 14 

I h~1 --- '1 _________________ 12i 1 Phas

Standard Plant 

- NRC Licensing 

S ,Certification 

-R&D 

Prototype Plant 

- NRC Licensing 

- DeslgnrCertfficatlon 

- Site Permlt/Envioro. Impact 

- Equip.Fab. & Site Construct 

- Safety Testing 

- Comm. Power Generation

R PS 

Key Feature Tesi

- I i -

Pt

-t

Suyte, f Tess 

SafetyTest 

FSAR Plan Agmt.

FE

U6II8O '

Fuel L 

Au.o.

Full 

Power

A uthorizm 

-I Load Safety Te 

hmAark

FL 

Ben 

TeS4

k Des 

Certific.

Safety Test Report Agnt

ition

n tion

Design Certification would be obtained through the construction 

~and testing of a single 380 MWe module

161

A CRS WobWhop

53 Boardnan

A CRS Wo,",d:j

Boardman CIJl '-J,,•V J'T

prwhirlm

)I -1.1k r-_- -

I n etn f singl 38 M emdl: j•

I

Ption

Environ. Repon SiePefd 

start Corr

t Report 

F 

COrn'nOp.
Topics

"* Incentive for developing S-PRISM 

"* Design and safety approach 

"* Design description and competitive potential 

"• Previous Licensing interactions 

"* Planned approach to Licensing S-PRISM 

"* What, if any, additional initiatives are needed?

June 4-5, 2001

* Safety Review/Key Issues

Safety Methods 

Containment

S 

Fuels 

0 

Waste 6

I Reseazrch 1956, - I 0.1 1 - I PUI Hg

More than 20 Sodium cooled Fast Reactors have been built 

Most have operated as expected (EBR-II and FFTF for example) 

The next one must be commercially viable I l .I=•l l l a

56 Bowrdman

ACRS WoeWh• 55 Boaon

NAME 

Franct 

INDIA 

IT1ALY 

•PEC 

JAPAN 

JOYD 

UK 

USA Q•in 

EBR-1 

EBR-2 

Emico Fermi 

SEFOR 

ým

USSR 

BR-2 

BR-5 

BOR-60 

BN-350 

BN-60O 

S1N-BO BN-1600 

W. Germany 

KNK 

SNR-300 S•NR-2

LCATIONh 

M22COWe 

Dowury 

Los Alamos 

Idabo 

Ricbhand

Obmiosk 

Melekess 

Shevchen 

Beloyarsk 

Kariruhe 

KRikar

Core energetic potential 

Analysis of Design Basis SG Leaks 

PRA 

Nuclear Methods 

T/H Methods 

Validation offuels data base (metal/oxide) 

Fission Product Treatment and Disposal

ACRS Worksop June 4-5. 2001 

162

M 9 --" - Oak Ridve

SNE-2 Kal I Odemrv~auqn I -- I -- 1'z ..34 14- I yv~r I -- ,
Component Verification and Prototype Testing 

Final component performance verification can be performed during 

a graduatedprototype testingprogram. 

Example: The performance of the passive decay heat removal 

system can be verified prior to start up by using the Electromagnetic 

Pumps that add a measurable amount of heat to the reactor system 

Licensing through the testing of a prototypical 

reactor module should be an efficient approach to 

obtaining the data needed for design certification. 

Defining the T/H and component tests needed to 

proceed with the the construction and testing of the 

prototype as well as defining the prototype test 

program will require considerable interaction with 

the NRC 

ACuci w-SJune 2001 57 n

163
G. Apostolakis, ACRS Chairman: What are the most dangerous mines in the mine 

field that you feel we ought to be working on? 

A. Rao, General Electric Nuclear Energy: Our experience on the last go round was a 

time and material effort. There tends to be no closure when you're having NRC review 

of the licensing submittals, whether it's with the national labs which are consultants to 

the NRC staff or the NRC staff. So there is a minimum incentive for closure of some of 

the items. That was our experience with the SBWR in the past. 

We don't think there are any technical issues that are there because we've had -- I 

haven't emphasized the international part of our meetings. Typically we meet twice a 

year and have 30 or 40 people from national labs and people from all different parts of 

industry. So we don't think there's any technical issues. It's just bringing the NRC staff 

up to the same state where we are. That's one thing. 

The other question is do the people who reviewed the SBWR in the NRC staff, are they 

still there? I think some of them are still there. That would make it go faster. The 

process of someone else coming up to the same level of understanding as those who 

worked on it is, I think, one of the major challenges we faced in the SBWR. I remember -- I don't know whether it was Ivan Catton or someone on the ACRS. It took several 

years before we got people to appreciate how simple our passive containment cooling 

system was, for example. It was actually not a natural circulation system. It was a - circulation system. And so if the same members of the NRC staff are not there, we 

might have to go through that same process again. So it's those kind of institutional 

issues, I think, which will be a harder challenge for us.

164
Summary of NRR Future Licensing Activities

In response to a renewed interest in building nuclear power plants, the NRC has 

created organizations within its major program offices to prepare the NRC staff for new 

applications (early site permits [ESPs], design certifications, and combined licenses) 

and to manage special task groups and pre-application reviews of new reactor designs. 

Activities planned in FY2001 and FY2002 include: (1) evaluating the ability of the NRC 

staff to support future application reviews under 10 CFR Parts 50 and 52; (2) 

performing pre-application reviews of the AP1 000 (a light-water reactor design with 

passive safety systems), Pebble Bed Modular Reactor (PBMR - a high-temperature 

gas-cooled reactor design), ESPs, IRIS (an advanced light-water reactor design), and 

GT-MHR (a high-temperature gas-cooled reactor design); (3) initiating and/or 

performing related rulemakings that will update 10 CFR Part 52 to reflect lessons 

learned from certifying three nuclear plant designs, update Tables S-3 and S-4 of 

10 CFR Part 51 to address higher burnup fuel considerations and non-LWR advanced 

designs, and address alternative siting considerations; (4) reactivating the construction 

inspection program; and, (5) interacting with stakeholders to ensure there is a clear 

understanding of upcoming activities related to future applications and to solicit 

stakeholder input. 

In FY2002 and FY2003, activities are expected to include: (1) managing the reviews of 

fine new applications resulting from the pre-application reviews (including one design 

certification, one combined license, and three ESP reviews); (2) managing two pre

application reviews (IRIS and GT-MHR); (3) updating regulatory and review guidance 

for new applications, i.e., Standard Review Plans (SRPs), Regulatory Guides, and 

referenced codes and standards, and identifying where enhancements are needed; 

(4) developing independent codes to analyze the safety of non-LWR designs, with 

supporting validation testing; and, (5) addressing regulatory infrastructure issues, 

including NEI's proposed New Plant Regulatory Framework initiative, and NRC 

regulations governing financial issues and operator staffing.

165
SREGUZ 

i?? 

F-1 

ACRS WORKSHOP ON ADVANCED REACTORS 

JUNE 4, 2001 

NRR FUTURE LICENSING ACTIVITIES 

INTRODUCTION: M. Gamberoni 

FUTURE LICENSING AND INSPECTION READINESS: N. Gilles 

EARLY SITE PERMITS: T. Kenyon 

ITAAC/CONSTRUCTION: T. Kenyon 

AP1 000: A. Rae 

REGULATORY INFRASTRUCTURE: E. Benner 

2

166
FUTURE LICENSING ORGANIZATION 

Associate Director for Inspection and 

Programs

:Richard B*arett 

SES Manager J 

:Marsha -Gamberoni 

;Section Chief 

J. . ~ Wlso ~ae-. E..Benner A.CbbglD Jackson: 

Sr.-PolicyAnalyst A. 000113M PeguIatory Infrastrcture ;PBMRIGT-MHRARIRS PMs; 

--SiflgPM4uRA1.ead 1TAA.C/Consthiction.PM -SenioPt 

FUTURE LICENSING AND INSPECTION READINESS 

ASSESSMENT (FLIRA) 

* Evaluate Full Range of Licensing Scenarios 

"* Assess Readiness to Review Applications & Perform Inspections 

- Staff Capabilities 

- Schedule and Resources 

- External Support 

- Regulatory Infrastructure 

"• Recommendations: 

- Staffing 

- Training 

- Contractor Support 

- Schedules 

- Rulemakings & Guidance Documents 

" Complete Assessment by September 28, 2001

167 4
EARLY SITE PERMITS 

Early Site Permits (ESP) 

- Site Safety 

- Environmental Protection - Emergency Planning 

* 10 CFR Part 52, Subpart A 

- Regulatory Guides - Environmental SRP - Experience with Environmental Reviews on License Renewal 

a Initial efforts 

- Coordinate Preparations for ESP Reviews 

- Interact with Stakeholders 

- Recent Meetings with NEI ESP Task Force 

* Applications 

- One in 2002, Two in 2003, Three in 2004 

5 

ITAAC/CONSTRUCTION 

"* Construction Inspection Program Re-activation 

- Develop Guidance for Inspection of Critical Attributes - Include Inspections for Plant Components & Modules at Fabrication Site - Initiate Development of Training for Inspection Staff 

"* Reactivation of Construction Permit (WNP-1) 

"• Resolution of "Programmatic" ITAAC

168 6
AP1000 PRE-APPLICATION REVIEW 

" Phase 1 Complete 

- July 27, 2000 Letter Identified 6 issues that Could Impact Cost and 

Schedule of Design Certification 

" Phase 2 Scope 

- Applicability of AP600 Test Program to AP1 000 Design 

- Applicability of AP600 Analyses Codes to AP1 000 Design 

- Acceptability of Design Acceptance Criteria in Selected Areas 

- Applicability of Exemptions Granted to AP600 Design 

" Phase 2 Schedule 

- Receipt of Analyses Codes Will "Officially" Start Phase 2 - Estimated Duration of Review - 9 Months 

"* Phase 3 - Westinghouse Application 2002? 

7 

REGULATORY INFRASTRUCTURE 

Current Activities: 

" Rulemaking to Update 10 CFR Part 52 

- Incorporate Previous Design Certification Rulemaking Experience 

- Update Licensing Processes to Prepare for Future Applications 

- Proposed Rule Package (9/01) 

" Rulemaking on Alternative Site Reviews 

- Amend Requirements in 10 CFR Parts 51 and 52 for NEPA Review of 

Alternative Sites for New Power Plants - Initiation of Rulemaking - Mid-FY2002 

" Rulemaking on 10 CFR Part 51, Tables S3 and S4 

- Amend Part 51 Tables S-3 & S-4 for Fuel Performance Considerations 

and Other Issues to Reflect Current and Emerging Conditions in the 

Various Stages of the Nuclear Fuel Cycle

169 8
REGULATORY INFRASTRUCTURE 

Financial-Related Regulations 

- NRC Antitrust Review Requirements 

- Decommissioning Funding Requirements 

- Modular Plant Requirements (Price-Anderson) 

Future Activities: 

-NEI Petition for Generic Regulatory Framework 

- NEI Intends to Propose Risk-Informed GDC, GOC and Regulations - Petition Anticipated in December 2001 

- NEI Proposal May Be Similar to Option 3 of RIP50 

• Licensing of New Technologies 

- Short-Term: Address via Existing Regulations, License Conditions and 

Exemptions - Long-Term: Address via Rulemaking 

9

170
Office of Nuclear Regulatory Research 

Summary of Advanced Reactors Activities 

June 4, 2001 

John H.Flack (Branch Chief) and Stuart D.Rubin (Senior Advisor) from the Division of 

Systems Analysis and Regulatory Effectiveness (RES), Regulatory Effectiveness and 

Human Factors Branch (REAHFB), provided an overview of the historical and current 

role of RES in pre-application reviews of advanced reactors. Pre-application 

interactions with potential licensee applicants will help NRC prepare for future 

submittals, through the development of the infrastructure necessary for licensing 

application reviews. RES has the lead for non-LWR advanced reactor pre-application 

initiatives and longer-range new technology initiatives. An advanced reactor group has 

been formed in REAHFB, and is currently performing a pre-application review of 

Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre

application interaction include General Atomics' Gas Turbine-Modular Helium Reactor 

(GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) 

design. RES advanced reactors activities also include participation as an observer in 

DOE's Generation IV initiative. 

Pre-Application review objectives include the development of regulatory guidance, 

licensing approach, and technology-basis expectations for licensing advanced designs, 

including identifying significant technology, design, safety, licensing and policy issues 

that would need to be addressed in the licensing process. In addition, the pre

application review will help to develop necessary analytical tools, obtain contractor 

support, train staff to achieve fully the capacity and the capability to review advanced 

reactor license applications. 

The presentation described the pre-application process for the Exelon PBMR. NRC 

first identifies additional information following topical meetings with Exelon, and Exelon 

formally documents and submits required topical Information. The staff then develops 

a preliminary assessment and drafts a response which is followed by stakeholder input 

and comments at a public workshop. Preliminary assessments are discussed with 

ACRS and ACNW, and Commission papers are written which provide staff positions 

and recommendations on proposed policy decisions. Some of the significant areas for 

the PBMR include: 

Process Issues, Legal & Financial Issues 

Regulatory Framework 

Fuel Performance and Qualification 

Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials) 

Fuel Cycle Safety Areas 

PRA, SSC Safety Classification 

PBMR Prototype Testing 

Sources of expertise for the PBMR include, RES, NRR, NMSS, OGC technical 

expertise and associated regulatory experience, contractor support from National Labs,

171
prior NRC Modular HTGR pre-application review experience, design, operating and 

safety review experience for Fort St. Vrain HTGR, International HTGR experience 

including IAEA, Japan, China, Germany, UK, and external stakeholder comments, 

ACRS and ACNW advice and insights.

172
United States 

Nuclear Regulatory Commission

Office of Nuclear Regulatory Research 

Advanced Reactors Activities 

June 4, 2001 

John H.Flack 

Stuart D.Rubin

Introduction

Historical role of RES in preapplication reviews 

Preapplication review of advanced reactors 

Current role of RES in advanced reactor reviews 

Advanced reactor group in Division of Systems Analysis and Regulatory 

Effectiveness (RES)

173

0. 

2 

., ...1-to.

0 

0 

0 

0
Advanced Reactor Activities 

* Advanced reactors have greater reliance on new technology and safety features. 

0 Preapplication interactions and reviews will help NRC prepare for licensing application 

* NRR has lead with RES support for LWR advanced reactor preapplication initiatives and 

licensing application reviews 

0 NMSS has lead for fuel cycle, transportation and safeguards 

0 RES has lead for non-LWR advanced reactor preapplication initiatives and longer-range 

new technology initiatives 

* Recent industry requests for preapplication interactions: 

Westinghouse: AP1000 (5/4/00) 

Exelon: Pebble Bed Modular Reactor (12/5/00) 

General Atomics: Gas Turbine-Modular Helium Reactor (3/22/01) 

Westinghouse: International Reactor Innovative and Secure (4/06/01) 

0 NEI Risk-Informed framework for Advanced Reactor Licensing 

RES Advanced Reactors Activities 

* PBMR: 

- Request for pre-application interactions received from Exelon 

- NRC response 

- Plan developed (SECY-01-0070) 

- Pre-application work underway (FY2001-2002) 

- Objective - identify issues, infrastructure needs and framework for 

PBMR licensing 

- Develop nucleus of staff familiar with HTGR technology 

* GT-MHR 

- Request for pre-application interactions received from General Atomic 

- NRC Response

174
RES Advanced Reactors Activities (cont.) 

"* IRIS 

- Developed under DOE-NERI program 

- Initial meeting on 05/07/01 

"* Generation IV 

- International activity coordinated by DOE 

- Longer term 

- NRC participating as an observer 

"* Generic Framework: 

- NEI developing proposal 

- Need for NRC to establish an effective and efficient risk-informed,and 

where appropriate, performance-based licensing framework 

Significant Technology Issues: 

* Unique, First of a Kind Major Components 

* Fuel Design, Performance, Qualification, & Manufacture 

* Source Term 

* Thermal-Fluid Flow Design 

* Hi-Temperature Performance 

* Containment 

* Fuel Cycle Safety & Safeguards 

* Prototype Testing and Experiments 

* Human Performance and I&C 

* Probabilistic Risk Assessment Methodology and Data 

* Emergency Planning 

* Regulations Framework 

- design basis accident selection 

- safety classification 

- acceptance criteria 

- GDC, 

- use of PRA 

- Safety Goals

175
PBMR Pre-Application Review Objectives

To develop guidance on the regulatory process, regulations framework and the 

technology-basis expectations for licensing a PBMR, including identifying 

significant technology, design, safety, licensing and policy issues that would 

need to be addressed in licensing a PBMR. 

To develop a core infrastructure of analytical tools, contractor support, staff 

training and NRC staff expertise needed for NRC to fully achieve the capacity 

and the capability to review a modular HTGR license application. 

PBMR Pre-Application Review Guidance

"* Commission Advanced Reactor Policy Statement 

"• NUREG-1226 on the Development And Utilization of the Policy Statement 

"* Previous Experience with MHTGR Pre-Application Review 

"* Identify Safety, Technology, Research, Regulatory & Policy Issues

176
PBMR Pre-Application Review Scope

S•.!•.nted Desian. Technoloav and Reaulatorv Review Areas:

Fuel Design, Performance and 

Qualification 

" Nuclear Design 

" Thermal-Fluid Design 

"* Hi-Temp Materials Performance 

" Source Term 

"* Containment Design 

"* PBMR Regulatory Framework

"* Human Performance and Digital I&C 

"* Prototype Testing Program 

"* Probabilistic Risk Assessment 

"* Postulated Licensing-Basis Events 

"* Fuel Cycle Safety 

"* Emergency Planning 

"• SSC Safety Classifications

PBMR Pre-Application Review Process

Conduct Periodic Public Meetings on Selected Topics: 

Process Issues, Legal & Financial Issues, Regulatory Framework (4/30) 

Fuel Performance and Qualification (6/12-13) 

Traditional Engineering Design (e.g., Nuclear, Thermal-Fluid, Materials) 

Fuel Cycle Safety Areas 

PRA, SSC Safety Classification 

PBMR Prototype Testing 

* NRC Identifies Additional Information Following Topical Meetings 

• Exelon/DOE Formally Documents and Submits Topical Information 

"• NRC Develops Preliminary Assessment and Drafts Documented Response 

"* Obtain Stakeholder Input and Comments at a Public Workshop 

"* Discuss Preliminary Assessments With ACRS and ACNW 

"* Commission Papers Provide Staff Positions and Recommend Policy Decisions 

"* Commission Provides Policy Guidance and Decisions 

"* NRC Staff Formally Responds to Exelon with Positions and Policy Decisions

177

elected Desion Technology and Regulatory
PBMR Pre-Application Review Sources of Expertise

RES, NRR, NMSS, OGC Technical Expertise and Regulatory Experience 

"* Contractor Support From National Labs and Design/Technology Experts 

"* Prior NRC Modular HTGR Pre-Application Review Experience 

"* Design, Operating and Safety Review Experience for Fort St. Vrain HTGR 

International HTGR Experience: IAEA, Japan, China, Germany, UK 

"* Exelon and DOE Design, Technology and Safety Assessments 

"* External Stakeholder Comments 

"* ACRS and ACNW Advice and Insights 

PBMR Safety Significant Review Issues/Topics 

• Fuel Performance and Qualification 

• High Temperature Material Issues 

• Passive Design and Safety Characteristics 

• Accident Source Term and Basis* 

0 Postulated Licensing Basis Events* 

* Prototype Testing Scope and Regulatory Credit 

* Containment Functional Design Basis* 

0 Emergency Planning Basis* 

• Risk-Informed Regulatory Framework* 

• Probabilistic Risk Assessment 

Commission Policy Decision Likely Is Needed 

178
PBMR Pre-Application Review Schedule

" About 18 months to Complete 

" Monthly Public Meetings To Discuss Topics 

" Feedback on Legal, Financial and Licensing Process Issues (-9/01) 

Feedback on Regulatory Framework (-12/01) 

Feedback on Design, Safety, Technology & Research Issues (-6/02) 

Feedback on Policy Issues (-10/02) 

Regulatory Infrastructure Development Needs 

" Staff Training Course for HTGR Technology 

" Analytical Codes and Methods for Advanced Reactor Licensing Reviews 

"* Regulatory Framework for Advanced Reactor Licensing Reviews 

"* Core Staff Capabilities for Advanced Reactor Licensing Reviews 

"* Contractor Technical Support Capabilities 

"* Possible RES Confirmatory Testing and Experiments 

Possible Codes and Standards for Advanced Reactor Design and Technology

179
G. Apostolakis, Chairman, ACRS: If someone comes to you using Part 52, is there 

anything there that says that you need the risk-informed, performance-based system? 

J. Flack, RES: There's nothing in Part 52 that says that we need to have a risk

informed, performance-based licensing approach. 

G. Apostolakis, Chairman, ACRS: So they could approach the licensing issue without 

using risk information. Could they? 

J. Flack, RES: Yes, I would expect that would be the case. 

G. Apostolakis, Chairman, ACRS: Is there anything that gives you the authority to 

request risk information? 

J. Wilson, NRR: The Part 52 licensing process is just that, it's a licensing process, and 

so it references back to parts 20, 50, 70 and 100 for the actual safety requirements. So 

whether or not those safety requirements remain as they are or change as a result of 

some risk-informed process, it will use whatever is the requirement that's currently in 

place. 

G. Apostolakis, Chairman, ACRS: What if the industry doesn't want to use risk 

information? What if they just want to use existing regulations with exemptions or 

changes and maybe they feel that going to a risk-informed system adds an impediment 

because we have to understand it and do it. It's new. And try to go with the existing 

system and maybe a PRA would be an assessment at the end if you guys request it but 

maybe it will be a good idea not to bring it up at all. Why is that the need? 

J. Flack, RES I think it would be to their advantage to come in that way. 

G. Apostolakis, Chairman, ACRS: We heard today from several speakers, I think, that 

they're trying to reduce involvement of the humans. Do you think that the human 

performance issue will be as important here as in the current reactors? 

J. Flack, RES: I've discussed this at length. I don't know whether we can say it's going 

to be less important. I mean it's going to be a different environment which that human 

operates in, and one has to understand that environment and what's changing in that 

environment. So it's something that one has to look at very carefully. So it's hard to 

say. 

D. Powers, ACRS Member: It seems to me that the change is really entertaining and 

in the direction that's most difficult for us because as they design the plants to be less 

and less dependent on the human operator intervening. We become more and more 

worried about the fact that the operators are not going to sit there and do nothing. They 

will intervene and the potential for them to intervene incorrectly in a system that's 

designed to operate with rather minor low head forces operating on it. So you get into

180
the problem of errors of commission that we are most incapable of addressing. It's a 

subtle problem. 

J. Flack, RES: Yes. The environment changes and you don't really have as much 

data as you wish you'd have to go on. 

J. Garrick, Chairman, ACNW: This is probably the question that I was half asleep on 

when George asked the question about the risk assessment. But you mentioned that 

on the PBMR you're going to get a risk assessment. What's the nature of that? Has 

that been requested? 

S. Rubin, RES: We have urged Exelon to provide as much information on the current 

risk assessment that they've done for the plan to support our review of this risk

informed framework for making licensing decisions. I wouldn't call it a risk-informed 

regulations framework as the extent of wholly replacing Part 50 but we think we now 

understand that this framework is not quite going to do that but will through risk insights 

be able to identify systems requirements for mitigation, prevention, the level of 

redundancy in those systems, which systems should be designated as safety significant 

and also things like what are the special treatment requirements on the system. But 

we're not talking about a regulations framework which covers all of Part 50. 

But to answer your question, we have asked for that and we've also asked, to the extent 

possible, that we get information on the design itself. We have not yet, except for these 

kinds of viewgraphs that we've seen today, gotten what I would call a significant design 

description and principles of operation document from Exelon. I think the staff would 

very much like to get both a PRA and a design description so we have a context for 

reviewing this framework. It is on our schedule. We talked about that. It's not now but 

it is later.

181
Nuclear Energy Institute (NEI) Summary

Prepared by ACRS Staff for R. Simard 

Mr. Ron Simard of the Nuclear Energy Institute (NEI) provided a brief presentation on 

the state of energy demand in the United States and discussed the improving 

economics for new nuclear power plants. He discussed the consolidation of companies 

under deregulation and the ability of these larger companies to undertake large capital 

projects such as nuclear power plant construction. He discussed efforts under way to 

support a new generation of plants but noted that there needs to be greater certainty in 

the licensing process. He discussed infrastructure challenges in terms of people, 

hardware, and services to support new and current plants. He stated that there needs 

to be fair and equitable licensing fees and decommissioning funding assurance for 

innovative modular designs such as the PBMR. He concluded that NRC challenges will 

include resolving 10 CFR Part 52 implementation issues, establishing an efficient and 

predictable process for siting, COL permits and inspection, and an increasing regulatory 

workload.

183
184

Licensing needs for future plants 

ACRS workshop on Regulatory Challenges 

for Future Nuclear Power Plants 

Ron Simard 

Senior Director, Business Services 

Nuclear Energy Institute 

June 5, 2001

New Nuclear Power Plants - New 

Momentum 

No Growing electricity demand, need for new generating 

capacity 

0, Fossil fuel price volatility, clean air constraints 

O Improving economics of new nuclear power plants 

0,ý Industry consolidation = companies large enough to 

undertake large capital projects 

No Significant public and political support 

0 Potential for greater certainty in the licensing process
185

Focus of efforts to pave the way for 

new plants 

"• Policy, legislative, regulatory changes needed 

to support new approaches to ownership, risk 

sharing and project financing 

"• Policymaker support (Administration, 

Congress and others) 

"* Infrastructure (people, hardware, services) to 

support new and current plants 

"* Licensing, licensing, licensing
Licensing needs with respect to ...

186

"• working out the Part 52 implementation details 

"• assuring safety and equitable application of 

regulations to new types of designs 

"* clarifying how financial related requirements 

apply in the new business environment 

Um

Examples of Part 52 licensing needs 

"• a timely and efficient ESP process (e.g., 

focusing on the incremental impacts of 

additional reactors at existing sites) 

"• a timely and efficient process for COL 

applications and reviews 

" an efficient process for construction inspection 

and ITAAC verification 

U
New Licensing Process Significantly 

Reduces Project Risk

Old'Poes Construction 

Permit 

Application 7 

NeW Process 

Design Construction Permit! 

Certification or I Operatin License 

Plant-Specific FSAR+ITAAC -Appliction Aplcto 

Early Site Permit non 

Project Decision

Operating 

License 

Application -V 

Operation

Acceptance 

Criteria Met 

Operation 

,triition

187

"New design" licensing needs (in 

addition to safety determinations) 

" For modular designs, clarification of 

"• number of licenses per facility 

"* application of Price Anderson requirements 

"• basis for Part 171 annual fees 

"° basis for control room staffing 

"• For gas cooled designs, clarification of 

- decommissioning funding levels 

- generic environmental impacts (Tables S-3, S-4) 

- basis for EP action levels, reporting requirements, 

implementation of NUREG-0654

S... ........ . I •n•TWTI•TI•
188

Licensing needs for the new business 

environment 

• Clarification of how financial related 

requirements apply to merchant nuclear plants 

"* no need for an NRC antitrust review 

"• nature of financial qualifications 

"* appropriate mechanisms for decommissioning 

funding assurance 

U

The nuclear energy imperative 

* DOE projects 400,000 MW of additional 

capacity needed by 2020 (to replace 

existing plants that reach end of life and to 

meet new demand) 

* 30% of our current generation is non 

emitting (nuclear, hydro, renewables) 

* maintaining that contribution to clean air 

will require 50,000 Mwe of new nuclear 

U
189

Vision 2020 

5000 

4500- "4. 400- ., 31% 

2 4000- non

0 lJemnitting i:3600

- 30% 0 

" 3000- • non-: 

_ 2000- emttn ...... ie Cpailt 

S2000 E nhanced Nuclear Caa~lity 

1500- , " Existing Nuclear Cqaability 

1000- H ,ydro & Ronewcrles 

500 Oil, Gas & Cod 

2000 2020

The Future isn't what it used to be 

because ... 

0 Electricity demand will continue to grow 

i New nuclear generation is no longer an option 

- it is an imperative 

' The business case for new 

nuclear plants will be clear 

0- The cost and schedule drivers must 

be known and manageable to much 

more certainty than in the past
The Future isn't what it used to be 

because ... 

NRC will be challenged to 

l resolve Part 52 implementation issues in a timely 

manner 

o establish efficient and predictable processes 

for siting, COL license applications, 

construction inspection 

0 respond to an increasing workload 

with new focus, discipline and 

efficiency

190
D. Powers, ACRS Member: There seems to be a body of opinion that maybe we don't need 

that much electrical energy, and that we, in fact, can achieve the necessary energy supplied by 

conservation. 

R. Simard, NEI: No question that conservation and efficiency are important, but it's folly to 

think that you're going to conserve your way out of having to add almost a 50 percent increase. 

The gains that we have made in conservation have been impressive, at times, and efficiency 

has really helped quite a bit, but there's no way to conserve your way out of the low end of this 

projection without disruptive impacts on the economy. 

There are still some people who will question the need to have that much electricity and they 

might even go so far as to say that we can keep our current demand steady. The future isn't 

what it used to be because I think the consensus is here now that the demand will grow, and we 

used to talk about the nuclear option. It's not an option anymore. It's an imperative. 

The business case for new plants is pretty clear, but we have to have cost and schedules 

known to a greater degree of certainty than we ever had before, which leads us into the 

challenge for the NRC because, the ability to bring this plant to make depends upon being able 

to work out these Part 52 implementation issues in a timely manner, and having in place 

efficient and, Commissioner Diaz's word, "scrutable" processes for early siting and licensing 

and construction inspection. And what's emerging here from this day and a half is the 

challenge for NRC to be able to respond to this with a whole new focus and discipline and 

efficiency. 

D. Powers, ACRS Member: One of the persistent problems that we encounter when new 

things are brought to this particular body is the documentation is incomplete; the documentation 

is not rigorous. Those kinds of things slow the process substantially. 

Is the industry doing anything to try to address those kinds of questions? 

R. Simard, NEI: I think the challenge on our side is to bring in an unprecedented quality of 

application. On our side, we need to bring to the NRC the highest quality of information and 

application. What you're seeing both with the Westinghouse and PBMR North America 

International with NRC, is an effort early on to really clearly identify exactly what the staff needs 

are going to be to be able to do their review. 

G. Apostolakis, ACRS Chairman: In one of your earlier slides, it says "acceptance criteria 

met," do we have those criteria? 

R. Simard, NEI: Yes, in the three designs that have been certified, a key feature and a high 

level of detail in those certifications are the ITAAC. So they're clearly specified. In the ABWR, 

for example, the high pressure core flooder system, there were 31 separate ITAAC that clearly 

focus on the performance of a pump. For example, what inspections or tests will be done on 

that pump and what acceptance criteria will be necessary to show that, in fact, that pump is 

going to deliver the amount of water you need at the time you need it? So in the design 

certification, a key feature of them has been these ITAAC. We need to add a few more that are 

site specific when the licensee brings the application. 

T. Quinn, Consultant, General Atomics: The reason for success in the license renewal 

process to a large extent was the project management role that was put in place with a lot of 

work by NEI, with a lot of work by the NRC, and a suite of documents that became part of the

191
process, (e.g., the GALL report, and the NEI guideline). Have you considered working with 

NRC on a similar type of suite of documents to help us make this a more stable framework? 

R. Simard, NEI: Yes, I think you're right, Ted. That's been a good model in the past. By 

bringing to bear the range of industry resources and expertise on an area and combining that 

with the NRC, I think we've wound up with a better quality product in the end and improved the 

efficiency of the process. 

So building on our success with license renewal, maintenance rule or other things like that, it is 

our intent to put a lot of thought from our side into how -- for example, the format of an early site 

permit application, and that's something we actually have underway, or with respect to 

construction inspection at ITAAC verification, it's our intent to bring together the folks who still 

have construction experience in the industry, if we can find them, and again, drawing upon their 

expertise and our knowledge of how Part 52 -- the basic principles of Part 52. 

Again, it would be our intent in cases like that to bring in a document and ask the NRC for its 

review and reactions and use that as the framework for these productive discussions.

192
ACRS WORKSHOP 

Regulatory Challenges for Future Nuclear Power Plants 

"Safety Goals for Future Nuclear Power Plants" 

Neil E. Todreas 

KEPCO Professor of Nuclear Engineering 

Massachusetts Institute of Technology 

This talk presents technology goals developed for Generation IV nuclear energy 

systems that can be made available to the market by 2030 or earlier. These 

goals are defined in the broad areas of sustainability, safety and reliability, and 

economics. Sustainability goals focus on fuel utilization, waste management, 

and proliferation resistance. Safety and reliability goals focus on safe and 

reliable operation, investment protection, and essentially eliminating the need for 

emergency response. Economics goals focus on competitive life cycle and 

energy production costs and financial risk. 

The goals have three purposes: First, they define and guide the development 

and design of Generation IV systems. Second, they are challenging and will 

stimulate the search for innovative nuclear energy systems-both fuel cycles and 

reactor technologies. Third, they serve as the basis for developing criteria to 

assess and compare the systems in a technology roadmap. 

The Generation IV technology goals derive from a set of guiding principles: 

Technology goals for Generation IV systems must be challenging and 

stimulate innovation. 

Generation IV systems must be responsive to energy needs worldwide. 

Generation IV concepts must define complete nuclear energy systems, 

not simply reactor technologies. 

All candidates should be evaluated against the goals on the basis of their 

benefits, costs, risks, and uncertainties, with no technologies excluded at 

the outset. 

The Generation IV technology goals are intended to stretch the envelope of 

current technologies. Hence, the following caveats are important to note: 

The goals will guide the development of new nuclear energy systems. 

The objective of Generation IV systems is to meet as many goals as 

possible. 

The goals are not overly specific because the social, regulatory, 

economic, and technological conditions of 2030 and beyond are 

uncertain. 

The goals must not be construed as regulatory requirements. 

Future designs will likely (but not necessarily) involve new fuel cycles and the 

capability to produce a broader range of energy products. For these reasons and

193
to enhance the economic performance of electricity-only producing systems, I 

anticipate: 

New Fuel Materials 

Higher Burnups 

Longer Operating Cycles 

Higher Temperature Operation 

These trends will be driven by the Sustainability ( SU 1, 2, +3 ) and the Economic 

(EC 1+2) Goals. 

Since these trends involve significant safety issues, all the goals should be 

considered as relevant contributors to the safety profile of future Generation IV 

energy systems. 

Each of the eight goals is presented and the key issues debated and decided 

upon in their formulation will be discussed. The illumination of this debate is 

reflected in the Viewgraphs by highlighting the wording in the Goals Statements 

that best embodies the deliberations. 

For the Sustainability Goals the following observations are relevant: 

Fuel cycle development offers the only way to address objectives of 

availability, waste management and nonproliferation in an integrated manner. 

Hence, for the US, R+D on fuel cycle options needs to be reinvigorated. 

The once -thru fuel cycle will likely be hard to beat considering that the 

objective of effective fuel utilization involves the following elements: 

- Economics (fuel cycle plus effect on O+M cost). 

- Nonproliferation concerns - challenge remains on cross-rating 

individual intrinsic and extrinsic barriers. 

- Environmental concerns - to what degree are externalities to be 

internalized in the nuclear fuel cycle and in other competing 

energy supply systems. 

For the Economic Goals the following observations are relevant: 

Legitimate differing views exist on whether "clear" life-cycle cost advantage 

will be needed over the 30 year horizon for introduction of GENlV systems or 

whether breakeven will suffice because of recognition/credit for 

environmental benefits derived from nuclear systems. 

A judgement has been made that the history of deployment of nuclear 

systems has so raised the specter of risk and uncertainty of deployment cost 

that a "clear" advantage will be necessary to induce a commercial 

194
commitment to GENIV (or any nuclear system i.e. NTD) systems. Further, 

although allowable financial risk is limited by the need to achieve the life cycle 

cost advantage, the nuclear deployment history has also so raised the issues 

of risk & uncertainty of deployment that the separate and specific EC-2 Goal 

on "level of financial" risk was deemed necessary. 

Anticipating enhanced interest in this audience in the three Safety and Reliability 

Goals, the text of the discussion which follows and supports each of these goals 

is also presented with the relevant wording similarly highlighted. 

The latest statement of these Goals which was presented to NERAC on May 1, 

2001 and subsequently accepted by DOE for final presentation to GIF is 

appended. 

Conclusion: 

Future reactors fall in three categories - those which are: 

* Certified or derivatives of certified designs. 

* Designed to a reasonable extent and based on available 

technology. 

* In Conceptual form only with potential to most fully satisfy the 

GENIV goals. 

My focus has been on goals for the third category. 

It will be desirable to develop a range of design options in this third category to 

enable response to a range of possible market demands such as: 

"* cheap versus expensive uranium 

"* small versus large power ratings 

"* significant reduction of greenhouse emissions 

"* new fuel cycles to achieve a significant response to the 

sustainability goals 

Considerable R+D activity will be required to achieve these goals among which 

fuels, materials, and coolant corrosion research are the most intensive and long 

term. 

Consequently it is important that while an early dialogue between designers and 

regulators occur, the dialogue be framed to encourage & promote fundamental 

design directions which inherently promote safety. Development of a new 

regulatory process using risk-based principles is an important element of this 

dialogue. Interactions which frame the dialogue around the current regulatory 

framework can have the undesirable intent of discouraging the necessary and 

desirable exploration of technology and design alternatives.

195
196

ACRS WORKSHOP 

Regulatory Challenges for Future Nuclear Power Plants 

Safety Goals for Future Nuclear Power Plants 

Neil E. Todreas 

KEPCO Professor of Nuclear Engineering 

Massachusetts Institute of Technology 

AM June 5, 2001 

.MI.T. Dept. of Nuclear Engineenug

HOW TO MISCONSTRUE THIS TALK 

I am not talking about: 

* NRC Safety Goals - Quantitative Health Objectives - CDF and LERF. 

* Suggested Regulatory Requirements for Future Power Plants. 

* Soley about Future Power Reactors. 

Goals for Near Term Deployment* Plants ( by 2010). 

I am talking about: 

* DOE and GIF Generation IV Technology Goals. 

* Technology Goals formulated to 

- stimulate innovation. 

- suggest metrics for downselection which specifically are not to be 

construed as regulatory requirements. 

" Nuclear Energy Systems Including 

- Fuel Cycles 

• Goals for Systems to be Deployed from 2011 to 220330. 

*Deploymet: Manufactur, constructrin, and startup of erttfled planxs ready to produce energy in their ehase market. 

MI.T. Dept. of Nucear Engineering 2
197

HOW TO MISCONSTRUE THE GOALS 

- Assume that new nuclear energy systems must meet every new goal 

- Tradeoffs among goal parameters must be made for each design. 

Future markets may value different parameters. 

Desirable outcome is a spectrum of designs each best suiting 

different market conditions hence different goals. 

- Some goals presently appear unattainable ( S+R 3). 

- Most goals are not overly specific because the social regulatory, 

economic and technological conditions of 2030 and beyond are 

uncertain. 

MLT. Dept ofNucear Engineering

HOW TO MISCONSTRUE THE GOALS (cont) 

Assume that all safety considerations are encompassed in the Safety and 

Reliability Goal grouping ( S+R 1, 2, +3 ) 

- Future designs will likely (but not necessarily) involve new fuel 

cycles and the capability to produce a broader range of energy 

products. For these reasons and to enhance the economic 

performance of electricity-only producing systems, 

I anticipate: "* New Fuel Materials 

"* Higher Burnups 

"* Longer Operating Cycles 

"* Higher Temperature Operation 

- These trends will be driven by the Sustainability (5SU 1,2, +3) and 

the Economic ( EC 1+2 ) Goals. 

M.IT. Dept of Nudear Enpocering
198

SUSTAINABILITY 

Sustainability is the ability to meet the needs of present generations while enhancing and not jeopardizing the ability offuture generations to meet society's needs indefinitely into 

the future. 

Sustainability-1. 

Generation IV nuclear energy systems including fuel cycles will provide sustainable 

energy generation that meets clean air objectives and promotes long-term availability 

of systems and effective fuel utilization for worldwide energy production. 

Sustainability-2. 

Generation IV nuclear energy systems including fuel cycles will minimize and manage 

their nuclear waste and notably reduce the long term stewardship burden in the 

future, thereby improving protection for the public health and the environment 

Sustainability-3. Generation IV nuclear energy systems including fuel cycles 

will increase the assurance that they are a very unattractive and least desirable route 

for diversion or theft of weapons.usable materials. 

Dept. or Nudear Enpneening

SAFETY AND RELIABILITY 

Safety and reliabilty are essential priorities in the development and operation of nuclear 

energy systems. 

Safety and Reliability -1. 

Generation IV nuclear energy systems operations will excel in safety and reliability. 

Safety and Reliability-2. 

Generation IV nuclear energy systems will have a very low likelihood and degree of 

reactor core damage. 

Safety and Reliability-3. 

Generation IV nuclear energy systems will eliminate the need for offsite emergency 

response. 

M.LT. Dept of Nudear Engineering
199

Safety and Reliability -1. Generation IV nuclear energy systems operations will 

excel in safety and reliability. 

This goal aims at inasng operational safety by reducing the number of events, equipmemt problems, 

and buman perfornance issues that can initiate acwidents or cause them to deteriorate into more severe 

accidents. it also aims at achieving increased nuclear energy systems reliability that will benefit their 

enoeucs. Appropriate requirements and robust designs are Deeded to advance sidh operational 

objecives and to support the demonstration of safet that enhances public cvnfidn 

During the last two decades. operatng nucear power plants have improved their safety levels 

significantly, as tracked by die World Association ofNudear Power Operators (WANO). At thesame 

time, design requirements have been developed to simplify their design, enhance their defense-in-depth 

in nuclear safety, and improve their constructahility, operability, maintainability, and economics

Increased emphass is being put on preventng abnormal events and on improving hbman performance 

by using advanced instrumentation and digital system Also, the demonstration of safety is being 

strengthened through prototype demonstration that is supported by validated analysis tools and testing, 

or by showing that the design relies on proven technology supported by ample analysis. testing, and 

research results. Radiation protection is being maintained over the total system lifetime by operating 

within the applicable standards and regulations. The concept of keeping radiation exposure as low as 

reasonably adchevable (ALARA) is being succesflly employed to lower radiation csposur 

Generation IV nducear energy syswm must continue to promote the highest levels of safety and 

reliability by adopting established principles and best practices developed by the industry and 

regulators to enhance public confidence, and by employing future technological advance. The 

continued and judicious purrsit of eseellence in safety and reliability is important to improving 

ecnoomis 

M.LT. Dept. of NudlearE.ngl rig 7

Safety and Reliability-2. Generation IV nuclear energy systems will have a very low 

likelihood and degree of reactor core damage. 

This goal is vital to achieve investment protection for the 

owner/operators and to preserve the plant's ability to return to power. 

There has been a strong trend over the years to reduce the possibility 

of reactor core damage. Probabilistic risk assessment (PRA) identifies 

and helps prevent accident sequences that could result in core damage 

and off-site radiation releases and reduces the uncertainties associated 

with them. For example, the U.S. Advanced Light Water Reactor 

(ALWR) Utility Requirements Document requires the plant designer to demonstrate a core damage frequency of less than 10" per reactor 

year by PRA. This is a factor of about 10 lower in frequency by 

comparison to the previous generation of light water reactor energy 

systems. Additional means, such as passive features to provide cooling 

of the fuel and reducing the need for uninterrupted electrical power, 

have been valuable factors in establishing this trend. The evaluation 

of passive safety should be continued and passive safety features 

incorporated into Generation IV nuclear energy systems whenever 

appropriate. 

M.LT. Dept. of Nuclear Engineering
200

Safety and Reliabilit'-3. Generation IV nuclear energy systems will eliminate the need 

for offsite emergency response. 

The intent of this goal is, through design and application of advanced 

technology, to eliminate the need for offisite emergency response. 

Although its demonstration may eventually prove to be unachievable, 

this goal is intended to stimulate innovation, leading to the 

development of designs that could meet it. The strategy is to identify 

severe accidents that lead to offexte radioactive releases, and then to evaluate the effectiveness and impact on economics of design features 

that eliminate the need for offsite emergency response. 

The need for offsite emergency response has been interpreted as a safety weakness by the public and especially by people living near nuclear facilities. Hence, for Generation IV systems a design effort 

focused on elimination of the need for offsite emergency response is 

warranted. This effort is in addition to actions which will be taken to reduce the likelihood and degree of core damage required by the 

previous goal. 

MI.T. DepL of Nucear Engineering

ECONOMICS 

Economic competitiveness is a requirement of the marketplace and is essential for 

Generation IV nuclear energy systems. 

Economics-1. 

Generation IV nuclear energy systems will have a clear life-cycle cost advantage over 

other energy sources. 

Economics-2. 

Generation IV nuclear energy systems will have a level of financial risk comparable to 

other energy projects. 

SMIT. D.pL or•Nuder Engnerri 10
201

CONCLUSIONS 

* Future reactors fa in tree categories - those which ame: 

Certified or deriva•,ne of certified desi,. 

* Designed to a reasonable extent and based on available tedhnoio'. 

* In Conceptual form only with pote-•il to most fully atisfy the GENV" goals. 

My focus has been on goals for the thirdt category. 

* It will be desirable to develop a range of design options in this third category to enable response to a 

range of marketing demands such as: 

" cheap versus expensive cranium. 

" small versus large power ratings. 

"* significant reduction of greenhouse emissions. 

" new fuel cycles to achieve a significant response to the sustainability goals. 

Considerable R+D activity will be required to achieve these goals among which fuels, materials, and 

coolant corrosion research are the most intensive and long term. 

* Consequently it is important that while an early dialogue between designers and 

regulators occur, the dialogue be framed to encourage & promote fundamental design 

directions which inherently promote safety. Development of a new regulatory process 

using risk-based principles is an important element of this dialogue. Interactions which 

frame the dialogue around the current regulatory framework can have the undesirable 

intent of discouraging the necessary and desirable exploration of technology and design 

alternatives. M.LT. DepL of Nuclear Engineering
D. Powers, ACRS Member: One of the questions that comes to mind, especially after the 

previous speaker portrayed something of a crisis appearing, I wonder if in looking at these 

goals and looking at new systems that you compare the more modern or the existing plants against them to see if we really need all new concepts, and the 94 new concepts that were portrayed to us yesterday or, in fact, how well do the existing plants meet these various goals that you've laid out? 

N. Todreas, MIT: The answer to that is on the metrics that we're going to develop to assess 

these new concepts. We've picked a standard, and the evaluation process will measure these 

new concepts against the standard. Is it better, much better, et cetera, worse, much worse, and the standard we picked is the advanced LWR with once through fuel cycle. The rest of your question asked me what's the answer going to be, and I don't know that yet. 

D. Powers, ACRS Member: I find that a peculiar standard to pick because we don't have a 

whole lot of experience with advanced LWR. With existing machines, we have a lot of 

experience, and that experience, at least my friends at NEI certainly provide metrics that 

suggest that experience, is outstanding right now. 

N. Todreas, MIT: I can see thinking about that, but if we're going to develop advanced 

systems, I would say from the vendor community and the development community, we've got ABWR experience to an extent, and we have some degree of real respect for what the designs have accomplished in the ALWR. As a minimum you'd include both, but I certainly wouldn't go back just to the operating reactors as the standard for the future. I wouldn'tlignore the 15 years of ALWR development. 

G. Wallis, ACRS Member: I think that while you're being innovative, you should not use -- you 

seem to be here really talking about core damage frequency, and that just may get you in a 

box, and I think to be innovative, get away from these terms of the past and be more general. 

T. Kress, Chairman, Future Reactors Subcommittee: I think it is fission products we're 

worried about. 

N. Todreas, MIT: That's a reasonable point. If you are saying that we ought to get away from 

terms of the past which will lock us into certain design directions and means of dialogue, that is really my whole message, too. If you're offering me a suggestion that says what I wrote doesn't 

go that way; I should go a different way, then I'd perfectly accept it. 

J. Garrick, Chairman, ACNW: We have to be a little careful not to unduly focus on fission products because for many of the most important scenarios it is not the fission products that's 

driving the long-term performance of Yucca Mountain. It's mainly, technetium and Iodine 129 

certainly are in there, but depending on the scenario and depending on how you look at it, 

Neptunium 237 is the principal driver. 

And also, in most low level waste situations, you find that much to our surprise most of the low level waste is uranium contaminated. So, again, the fission products are not driving the long term stewardship or management of a lot of the low level waste, but rather it's actinides. The 

same thing is true in WIPP for transuranic waste. Again, it's not fission products, but it's 

plutonium.

202
N. Todreas, MIT: But that also refers back to the sustainability goal. It really doesn't obviate 

the suggestion relative to Safety and Reliability (S&R) grouping 2 relative to core damage. I 

say that because what Garrick's comment really impacts on is the waste issue, not effectively 

the immediate release through core damage. 

T. Kress, Chairman, Future Reactors Subcommittee: In particular, do you have some sort 

of criteria on what it would take to eliminate this need? And if so, does that criteria encompass 

some sort of measure of defense-in-depth also? 

N. Todreas, MIT: That's how the ACRS ought to look at it. From the point of view of a regulator 

or a group advising a regulator. These are technology goals. These are goals we want to drive 

the designers into thinking about. 

T. Kress, Chairman, Future Reactors Subcommittee: How would you know if you met that 

goal? That was my question. What is the measure that you're going to use to say, "Okay. The 

technology we have here meets that goal." Whether or not it actually comes about or not is 

another thing. 

N. Todreas, MIT: The measure has got to be release of fission products or radioactivity of a 

certain amount past the boundary. 

D. Powers, ACRS Member: I can always find a way to get fission products out. Any design 

you come up with I can find a mechanism to get the fission products out to the point that it 

violates some emergency planning guide. 

E. Lyman, Nuclear Control Institute: There are a few goals that are really missing from this 

whole formulation. First of all, under sustainability you refer to one that minimizes, that a goal is 

minimizing and managing nuclear waste, but at the same time, you really should impose a 

requirement that the routine emissions from the entire fuel cycle, as well as, occupational 

exposures are also minimized because one of the concerns with fuel cycles that involve 

reprocessing are these additional routine emissions, and you have to balance whether the 

reduced risk in a repository is justified by increased short-term emission. So that's really 

something you have to keep to minimize at the same time or it doesn't make sense. 

Second of all, under the financial goals issue, you didn't really dwell on the one that requires or 

suggests that the financial risks should be comparable to other energy projects, and I was 

wondering if in that context, you would also have a requirement then, that Price Anderson 

protection not be extended to Generation IV plants because other energy projects don't require 

that kind of protection. 

N. Todreas, MIT: Yes, on the first point you brought up, the specifics of that have been 

recognized and will come up in Safety and Reliability group 1 because there we are talking 

about across the whole fuel cycle, and those routine emissions are picked up there. They could 

be picked up either place, but that's where they come up. 

And on Price Anderson, we didn't get into the specific item within the structure of the goal that 

can be picked up and debated. It's been debated to some extent, but we didn't pin it down and 

resolve it specifically. I know that's coming up legislatively.

203
R. Barrett, NRC: My question relates to the methods that we use for estimating the likelihood 

of core damage and the likelihood of release of radioactivity. 

If NEI is correct and we have 50,000 new megawatts of capacity out there, and those are 

modular reactors -- that's 500 cores, and in an environment like that you find yourself striving 

for lower and lower core damage frequencies, and as you do that, you begin to put more and 

more stress on the current methods of estimating core damage frequency, and you begin to get 

to the point where many people think you're beyond the capability and the limitations of the 

method and the ability to have a complete model. 

In addition, as you move to different types of reactors, you find that you're depending less and 

less on highly reliable, redundant, and diverse systems and more and more on the intrinsic 

capability of the core itself to withstand these accidents, and to withstand them either 

indefinitely or for long periods of time. And, again, the methods that we have today really don't 

deal very well with this kind of intrinsic, passive capability. 

So my question to you is the stated purpose of your effort is to stimulate innovation in the 

design of the reactors, and my question is: could you also complement that with trying to 

stimulate innovation in the methods that we use for analyzing the risk associated with these 

reactors? 

N. Todreas, MIT: Yes, I would answer that two ways. First, it's a good suggestion and a fair 

suggestion. There's nothing implicit in my statement that precludes risk methods development - what's going to come out of this fundamentally it is a spectrum of concepts to focus on, but 

much more than that, an R&D road map of activities to flesh out those concepts and the 

methods associated with the concept development is certainly part and parcel of that. So we 

could do that. 

The other thing though that I'd say is if we were to develop the methods we're going to have to 

reduce core damage frequencies further to get a desired output. So that really leads you to say 

that if you go with concepts now that are clones or like -- I'm talking about 20, 30 years down 

the road -- existing concepts are like these, you're going to reach a point where the methods 

can only go so far based on the existing design approaches, and so that's a clarion call to 

change those approaches and go toward -- well, first, you go toward situations that avoid core 

melt, but that's very limited in a sense that what you really want to do is do what Dana Powers 

was talking about. It's not core melt. It's the fission products, and it's the radioactivity in the 

dose from that, and that's what you've got to get after. So I would say we certainly would 

accept and develop methods, but what we are trying to do is stimulate. I'm talking about real 

innovation, beyond that, to try to open up approaches that really change the playing field. 

L. E. Hochreiter, Penn State University: It's not clear to me why in your conclusions you have to have small versus large power ratings. It seems like you're biasing yourself already 

towards a particular class of designs. 

N. Todreas, MIT: Yes. Yesterday I presumed the whole layout of this program was 

announced or was explained as an international program with eight to nine countries now, and 

one of the goals of the program is to come up with design solutions or concepts that meet 

markets internationally, and there are some international markets. Also if you listen in the 

United States, too, depending on the grid size, there are some markets that have a priority toward low rated systems. And so you have some of those, and then you also have the

204
traditional, if you talk about Asia, Japan, Korea, Taiwan, large systems. So inherent in the 

whole program, since it's looking at worldwide markets, we're going to have this dichotomy, 

these two parts, and no one reactor thrust or direction is going to meet them both. So you're 

going to have to come up with systems in both directions. Now, your point may be fine, but 

they're not going to be sellable in the United States or the industrialized world. That's fine, but 

we'll have a product for that. We just may not use the other product. 

N. P. Kadambi, NRC: If I understand the rules by which the South Africans are trying to 

license their plant, one of their goals is that in the long term the concepts employed should be 

amenable for society to make a decision that higher levels of safety need to be obtained from 

these energy systems. 

And therefore, one of their goals, as I read it, and if I should be corrected, I'd like somebody to 

point this out; one of their goals is the design should be amenable for society to demand higher 

levels of safety at some future time if we take, you know, these systems as operating for many 

decades. 

Where does such a concept fit into the kinds of goals that you have articulated? 

N. Todreas, MIT: Okay. On this let me give you a brief answer and ask for some help 

because I am not knowledgeable about a specific or the specific South African drive that you're 

talking about. I just haven't interacted with them specifically. I would say that even though 

these are general goals, we are going to have some kind of constraint because we're going to 

come up with a set of specific metrics that go with each of these goals. They're going to be as 

we go on a year or two -- there's going to be some numbers and some specificity here. So 

there's going to be a little bit of a lock-in with your desire to accommodate future societal wishes 

for enhanced safety. The way I interpret what you're saying is you come up with a design. 

Society decides they want more safety, and so this design has somehow got to be expandable 

or have margin or a way to capture more safety. That's how I understand it. 

So I don't know the answer. These goals have been pushed in through a discussion with the 

so-called GIF countries, of which South Africa is a part, and we didn't get any effective 

comment back from them that's relevant to what you said. But if Andy Kadak or somebody else 

can speak specifically to that, that would help me. 

J. Slabber, PBMR: In the South African concept, the baseline was to use existing technology 

as far as possible, existing technology that has been qualified and tested and proven to be 

acceptable for use in the PBMRs, and with a basis that the fuel is the central point of focus. 

And within that framework, we do the system design. Imbedded in the design is the 

requirement to be fulfilled that no reliance is placed on immediate operator action to bring the 

reactor to a safe state, and I again say, in inverted commas, inherent safety and small units, 

and usable for not only producing nuclear power, but also some other usable byproducts such 

as possibly desalination specific for South Africa. 

N. Todreas, MIT: Can I build on that maybe in answer to his question? I would say with that 

focus and the ability, as you went to successive improvements in fuel fabrication and fuel 

reliability, you could actually enhance your safety profile if the key focus is fuel, and that would 

be an answer back to how you reflect the future, the fuel.

205
J. Slabber, PBMR: I think the objective of any new innovative system should be to improve, 

but there is a limit because it's also costly. So improvement, the improvement for public 

acceptance, improvement of safety, that at the boundary you do not have to shelter and 

evacuate. These are all factored in to provide a facility which is still affordable and reliable.

206
Establishing a Safety and Licensing Basis 

for Generation IV Advanced Reactors 

"License By Test" 

Andrew C. Kadak 

Massachusetts Institute of Technology 

Workshop on Advanced Reactors 

Advisory Committee on Reactor Safeguards 

June 4-5, 2001 

Background 

The regulations for nuclear power plants (NPPs), as codified in 10 CFR Part 50 

and the general design criteria (GDC), were formulated more than thirty years 

ago. The objectives were to protect public health and safety by making the 

probabilities of accidents very low and providing mitigating capabilities in case 

such accidents occurred. Since methods for quantifying these probabilities were 

not available at the time, a conservative safety philosophy emerged that was 

called defense in depth. Much has been learned since that time. The 

development of probabilistic risk assessment (PRA) has allowed the 

quantification of the probabilities of accident sequences, thus providing analytical 

tools for the evaluation of individual defense-in-depth measures. PRA 

applications have shown that many of these measures do not contribute 

significantly to safety and have been termed an "unnecessary burden." At the 

same time, the integrated PRA approach to reactor safety has identified the 

need for additional requirements in some instances that were not foreseen by 

the traditional, defense-in-depth based, regulatory system. 

To gain maximum benefit from design innovation of advanced reactors, 

engineers must utilize these advances in our ability to assess the safety of 

NPPs. Unfortunately, the current regulations have not been changed to a 

significant degree to reflect PRA insights and capabilities. One of the major 

impediments to the Electric Power Research Institute's advanced light water 

reactor (ALWR) initiatives that led to the design criteria for ALWRs was the 

objective of trying to comply with existing regulations. This effort, while making 

some improvements in design, resulted in plants that were still limited in terms of 

innovation. The ALWRs are considered to be too expensive for the US market 

due to these requirements that are arguably still too prescriptive. 

As the United States proceeds to develop a new generation of nuclear energy 

plants commonly referred to as "Generation IV," a new regulatory approach is 

needed that will allow for innovation and that captures the important features of 

safety. The existing regulations have been developed for water-based reactors 

and are not easily adaptable to other technologies such as gas cooled reactors. 

At the present time, the NRC is working on "risk informing" their regulations. This 

207
process is still largely focused on light water reactors is very difficult; and will 

take many years to complete implementation (should that ever occur). Once 

concluded, the result is expected to be a mix of deterministic, probabilistic, and 

subjective defense- in-depth criteria that will be an improvement, but still water 

based. 

There is a recognized need by the International Atomic Energy Agency1 and by 

the technical community worldwide that for true innovation in reactor design, a 

new approach to establishing the safety of advanced nuclear power plants is 

required. This new approach is required not only for innovation in design but 

also for a rational process of licensing and deployment in a timely way. 

For the next generation of advanced nuclear power plants, it is likely to be 

necessary for safety regulation to proceed under some modified version of the 

current NRC safety regulatory system. The reason for this is that replacement of 

the current system is likely to demand such time and resources that the new 

system may not become available in a timely fashion. The existing regulatory 

system may, in fact, deter deployment of innovative reactor designs. Thus, it is 

important to formulate a safety basis for a licensing approach for advanced 

reactors to be pursued under the current system. In doing this we wish to 

identify areas where current regulations do not address the safety issues of new 

reactor concepts;; to utilize our risk-based regulatory approach for development 

of proposed treatments of these latter areas; and to formulate ways of utilizing a 

demonstration plant in conjunction with a testing program in future licensing. 

Structuralist vs Rationalist Approach to Defense in Depth 

Current regulations and standards are based, in large part, on the principles of 

defense-in-depth (DID) and safety margins, which have evolved since the first 

reactors were designed in the 1940s. The NRC's Advisory Committee on 

Reactor Safeguards (ACRS)2 and Senior Fellow John Sorensen et al3. discuss 

this evolution, identify two schools of thought on DID, labeled "structuralist" and "rationalist," and recommend an approach for risk-informed regulation. 

The two schools differ in the process used to deal with uncertainty in reaching an 

acceptable level of safety. The structuralist approach has evolved from the early 

days of nuclear power with a process of accumulating DID features until a 

1 M. Gasparini, "Verification of Defense in Depth for Operating Nuclear Power Plants," draft 

September, 1999. 

2 Letter to Shirley Ann Jackson, Chairman, U.S. Nuclear Regulatory Commission, from D.A. 

Powers, Chairman, Advisory Committee on Reactor Safeguards, Subject: The Role of Defense in 

Depth in a Risk-Informed Regulatory System, May 19, 1999. 

3 Sorensen, J.N., Apostolakis, G.E., Kress, T.S., and Powers D.A., On the Role of Defense-in

Depth in Risk-Informed Regulation. Proceedings of The International Topical Meeting on 

Probabilistic Safety Assessment, Washington, DC, pp. 408-413, 1999.

208
judgment was made that sufficient protection against uncertainty in performance 

had been achieved. With the development of PRA methods, the rationalist 

approach uses these tools to quantify uncertainty and to explicitly account for 

DID features in reducing uncertainties to acceptable levels. The main difference 

is that the structuralist accepts DID as a fundamental principle, while the 

rationalist would place DID in a subsidiary role. Additionally, the structuralist 

does not deal with uncertainties in a quantitative manner, while the rationalist 

takes advantage of the fact that advances in PRA allow the quantitative 

estimation of some of these uncertainties. For new plants, the rationalist 

approach to DID, employed within the context of PRA, is preferred to more 

effectively develop a body of regulations that eliminates requirements that do not 

contribute significantly to safety. 

The rationalist relies on PRA methods to provide an integrated and systematic 

analysis of the plant that explicitly addresses sources of uncertainty. The 

process envisioned by the rationalist is: establish quantitative safety goals, such 

as health objectives, core damage frequency, and large release frequency; 

design and analyze the plant using PRA methods to establish that the safety 

goals are met; evaluate the uncertainties in the analysis, including those due to 

model inadequacies, system performance and reliability, and lack of knowledge; 

and determine what steps (i.e., DID, new design features) to take to address 

those uncertainties. The quantification of uncertainties in terms of probability 

distribution functions provides a means for determining how much redundancy 

and diversity (i.e., DID) is sufficient. 

Discussion of the Risk-Based Framework 

The framework we propose for risk-based regulation and design is illustrated in 

Figure 1. A top-down hierarchy, indicated on the left side of Figure 1, is being 

used to define the goal, establish an overall approach, and develop and 

implement appropriate strategies and tactics. The framework is based on an 

application of PRA methods and reflects a rationalist approach to DID. 

Regulations for NPPs are required to ensure adequate protection to the health 

and safety of the public. Accordingly, the goal of this effort is to provide a 

framework for developing and implementing risk-based regulations that meet this 

requirement. An approach based on evaluating risk against quantitative safety 

goals is proposed to achieve the stated goal. With respect to adequate 

protection, the NRC has established safety goals including quantitative health 

objectives (QHOs) that state the Commission expectations with respect to how 

safe is safe enough. Although the NRC safety goals are not considered 

quantitative measures of adequate protection, for new plants, we will consider 

the determination of adequate protection using increased reliance on 

comparisons of PRA results to quantitative risk measures. The safety goals we 

are using for the framework, indicated in the gray boxes in Figure 1, have been 

adapted from the NRC goals.

209
The strategies for developing and evaluating compliance with requirements for 

risk-based regulation and design are based on the use of PRA to quantify risk 

and uncertainties. High confidence is achieved through explicit consideration of 

uncertainties, including modeling adequacy and equipment design and 

performance. These strategies include consideration of the risk information 

available from Level 1, Level 2, and Level 3 PRA analyses. Level 1 PRA 

evaluates the potential for accident initiators and the system response to prevent 

core damage. An estimate of core damage frequency is compared to the 

corresponding goal. Level 2 PRA encompasses the response to and mitigation 

of core damage, including containment of fission products. Risk estimates here 

can be compared to goals for conditional probability of large release, both early 

and late. Level 3 PRA encompasses the response to and mitigation of 

radionuclide releases, including emergency response. These risk estimates can 

be directly compared to the QHOs or to subsidiary goals for conditional 

probability of early fatalities and latent cancer risks. 

To develop risk-based regulations, implementation of the framework is achieved 

by defining functional system characteristics, within the context of how PRA is 

performed, to determine what areas need to be regulated to assure safety. 

Implementation for design is achieved by specifying design configurations and 

using PRA to evaluate the design, then iterating with subsequent design 

changes. A master logic diagram (MLD), illustrated in Figure 2, is used to take a 

top-down approach to identify the safety functions, and systems, structures, and 

components (SSCs) that are required to maintain safety and to identify the 

accident initiators and system response failures that could compromise safety4 . 

The top event is stated in terms of risk exceeding the safety goals. The gray 

shaded events correspond to the Level 1, Level 2, and Level 3 PRA strategies, 

respectively, in Figure 1. The sixth level of the MLD defines the system 

functions that are required to assure safety. The next level down indicates that 

initiating events and failure of mitigating systems, containment, and emergency 

response can compromise safety functions. The last level of the MLD indicates 

that internal initiators for all operating modes and external initiators will be 

considered for completeness. Further development of the MLD will determine 

the "regulatory risk space" for which regulatory and design requirements are 

needed. 

Various tactics (e.g., design criteria, procedures, redundancy, emergency 

response, etc.) are applied to support the PRA strategies and implementation. 

Once the SSCs required to achieve safety have been identified, then decisions 

on appropriate tactics for regulation and design can be made. The specification 

of these tactics will be based on a systematic evaluation of the areas that need 

to be regulated for the purposes of assuring safety and will also evolve from this 

' Apostolakis, G.E., Some Issues Related to Goal Allocation and Performance Criteria. 

Proceedings of the 8' International Conference on Structural Mechanics in Reactor Technology, 

Brussels, Belgium, Paper M2 4/3, 1985

210
process.

Work on the practical implementation of the risk-based regulatory approach has 

proceeded focusing upon how to formulate design basis accidents (DBAs) within 

a risk-based system5 In this work, it was concluded that DBAs are not 

necessary within such a system, and rather are replaced by the set of risk

dominant accident event sequences as the focus of judging whether a nuclear 

power plant design is adequate, and for negotiating between the license 

applicant and regulator concerning what changes in the design application are 

required before a license can be issued. These risk dominant accident 

sequences will be used in the development of a "license by test" approach to 

safety verification and licensing. 

The thinking on how to approach establishing a safety basis and then licensing 

non-traditional nuclear technologies was advanced in 2000 by ESKOM, the 

South African utility proposing to build a pebble bed reactor at their Koeberg 

nuclear plant site in Capetown. ESKOM, through its Pebble Bed Modular 

Reactor Development Company, PBMR Pty, issued "PBMR Safety Case 

Philosophy" - PNL-001 Rev 1, (10) and the Safety Case Support Document 

PNL-009 (11), that identifies the approach to safety being proposed by ESKOM 

for eventual licensing in South Africa. This approach calls for the use of PRA to 

establish general design criteria for the pebble bed reactor. Their approach is 

consistent with South African's National Nuclear Regulator's approach to safety 

that is illustrated on Table 1 which is safety goal based and uses a combination 

of deterministic and probabilistic techniques to establish general design criteria 

and tests that need to be performed to demonstrate safety. What is proposed by 

ESKOM is whenever there is inadequate information or large uncertainty to 

bound that uncertainty and perform tests on the prototype reactor to assess 

impact. It is judged that these guidelines can form the basis of a workable risk 

informed safety basis that will allow a license by test approach. 

License by Test (LBT) 

Both this effort and the work being performed internationally are general and 

could be applied to non-traditional technologies. What is needed is an extension 

of this process to specific reactor technologies, such as the gas reactor, and to 

include the concept of "license by test." The aim would be to improve safety and 

reduce the time and effort to certify new designs while not compromising safety. 

This weakness in the generic applicability of the existing regulatory system is not 

easily addressed. An approach that historically was provided in the regulations 

' B. C. Beer, G.E. Apostolakis, and, M.W. Golay, "Methods for Formulation of Design Basis 

Accidents Within a Risk-Informed Approach to Safety Regulation of New Nuclear Power 

Plants," PSAM 5 Conference, Osaka Japan (November, 2000). [Included as Appendix YYYI

211
was "license by test." In this concept, a reactor prototype could be built and 

tested to demonstrate the safety characteristics and on that basis granted a 

license. This approach has to our knowledge never been applied to an entire 

nuclear plant but only subsystems which were tested to calibrate performance. 

With the advent of new safer plants that derive their safety from inherent 

deterministic safety features as opposed to active or passive safety systems that 

must work, we have an opportunity to apply the license by test concept on a real 

plant on an integrated basis. The challenge is to develop the test envelope to 

validate safety in a licensing sense. 

What is desired is to avoid years of paper analyses and simulations which can 

be costly and still leave doubts in the regulator's and the public's mind about the 

real safety of the plant. As an alternative to spending the huge amounts on 

paper analyses, separate effects tests, it is proposed to design and build a 

prototype of a plant that meets the fundamental safety standards relative to 

public health risk. Using a combination of deterministic and probabilistic tools as 

described above to develop subsystem and integral tests that would be 

conducted with this prototype (full scale if the plant is small enough) to 

demonstrate the safety of the plant. In this case, we would not only have a 

certified plant design and a license at the conclusion of the process, but also a 

nuclear power plant that the public could see met the safety standard and that 

could also be used to produce electricity. 

This licensing concept has been suggested as being especially applicable to the 

PBMR.The basic idea of LBT is that a set of integral tests of a full size reactor 

and associated systems be used to demonstrate the ability of the overall system 

to mitigate successfully a set of "bounding" accident cases. Should the set be 

sufficiently comprehensive (e.g., loss of coolant inventory and loss of coolant 

flow), and should they be sufficiently more demanding than expected accident 

situations. The argument goes that successful performance in these tests 

should be sufficient proof of the safety of the nuclear power plant concept. 

These tests, even if partially successful, can also serve another very important 

purpose, namely, to validate the computer codes that have been developed for 

design and accident analysis. The validation of such codes is always an issue 

when a new design, even water-based, is submitted to the NRC. For example, a 

recent ACRS letter6 recommended that the validity of predictive codes such as 

NOTRUMP that have been approved for the design certification of AP600 must 

be demonstrated before they are used in the certification process for AP1 000. 

What is hoped is that the full scale tests will be able to answer remaining 

questions for the validation of codes. This approach can only be possible for 

reactors that have significant safety margins which is a criteria for new advanced 

reactors. 

6 Report from D.A. Powers, ACRS Chairman, to R.A. Meserve, NRC Chairman. Subject: "Pre

Application Review of the AP1 000 Standard Plant Design - Phase I."

212
The Challenge of Licensing by Test

The challenge is to define a process that would identify what range of testing 

would be required for the granting of a license. For example, what design 

information needs to be available to allow for the construction of the prototype 

and the extent of review by the regulator required for the test program? What 

criteria constitute success or failure of the test program leading to the granting or 

denial of the license? 

It is our hope to work with the NRC to develop the guidance to allow such a 

process to work such that advanced non water (or water based) technologies 

could be used without spending years on inconclusive (from the standpoint of the 

public perception) analyses. 

We need to develop this concept in more detail; establish clear performance 

indicators for success; identify review requirements for the design and 

construction of the prototype facility; and specify the test program that would be 

used to certify the design for mass production. 

Reference Plant Selection 

At the present time, there are many advanced reactor projects under 

development. The three most significant are the high temperature pebble bed 

gas reactor, the light water IRIS project and several variants using lead bismuth. 

All new reactor concepts would benefit from the application of the safety 

framework approach using PRA technology in a methodology that identifies 

important safety features and requirements. 

The modular pebble bed reactor has been selected as be the reference plant 

used as a demonstration for the new safety and licensing approach since it 

provides some unique challenges to conventional wisdom about safety and 

traditional licensing. As result of the leadership and development of this 

technology for commercial application by ESKOM, the South African utility and 

MIT's three year conceptual development project on the PBMR, there is 

sufficient design and knowledge information to allow this plant to be the 

advanced reference design. Eventually, this plant will be constructed in the US 

since PECO and BNFL, investors in the South African PBMR, have indicated 

that they would like to use the South African plant as the prototype for the US 

plants. In addition, I have been proposing the construction of a "reactor research 

facility"site which can be used to test the concepts proposed by this project. In 

either case, a new approach to the safety basis of advanced plants will be 

required to allow for timely introduction of new nuclear plants in the US. The 

licensing strategy for the research facility relies upon "license by test" that needs 

a workable risk-informed regulatory framework in which to succeed. 

This is a novel approach to safety and licensing that is made possible by the 

integrated approach to reactor safety that PRA provides. Whether this approach

213
is applied to the South African prototype for US application or for a new "reactor 

research facility" built on a DOE site, it is important to develop the safety 

framework upon which to license the plant. 

Establishing a Safety and Licensing Basis for the Plant Within the Current 

Regulatory System 

In order for license by test to be successful in the short term, it must fall within 

the guidance of the current regulatory system. There are several steps that need 

to be taken to assess whether this will work. These steps are outlined below. 

1. Review literature on safety bases used by the IAEA, NRC, and the South 

African regulator. Focusing on risk-based safety goal performance criteria 

addressing defense-in-depth issues.7 , 8 The NRC regulations should be 

reviewed as would the GDC for applicability. This will also include 

understanding the relationships for normal operation, severe accidents, 

containment of fission products and public health and safety objectives in the 

context of fundamental design objectives as outlined by IAEA and others. It 

will be necessary to identify areas in the current NRC regulatory system, 

utilizing available precedents, which currently address PBMR-relevant 

regulatory questions. NRC's regulations will be reviewed to identify 

opportunities for establishing license by test approach to demonstrate safety. 

2. Develop risk-based approach to address safety basis and regulatory 

questions which are not treated by existing regulations, and then to use risk

based logic for refining the existing regulatory structure. 

3. Within the context of the existing high level regulatory design objectives, 

develop a license by test process that can demonstrate functional compliance 

to these design objectives. Identify those areas where a "Licensing by Test" 

approach could answer unresolved questions safely within the existing regulatory structure and seeking applications of it in the overall PBMR 

regulatory strategy. 

The great safety licensing problem faced by the PBMR in the United States is 

that no body of regulations exists for licensing it comparable in detail than that 

available for LWRs. Much of the current regulatory literature and policies have 

been concerned with LWR-specific problems or features (e.g., metal-water 

interactions following onset of critical heat flux; pressurized thermal shock). This 

emphasis creates the need to ensure that LWR-influenced treatments of safety 

issues will not be incorrectly applied to PBMR safety questions and 

7 Clappisson, Metcalf, Mysen"PBMR-SA Licensing Project Organization", November, 1999, 

Beijing, China. 8 Kadak,"Risk Based Regulation - The Time is Now", PSA-99, International Topical Meeting on 

Probabilistic Safety Assessment, August, 1999.

214
requirements, also. With diligence and effort this pitfall can be avoided, but 

recognizing it is the start of doing so. 

The precedents created by the Ft. St. Vrain and Peach Bottom gas-cooled 

reactors will be useful in PBMR licensing, but the overall regulatory structure has 

changed considerably since the licenses for the reactors were issued (both prior 

to the 1979 Three Mile Island reactor accident), and they are silent concerning 

such PBMR aspects as reliance upon highly reliable fuels and the stochastic 

nature of the individual fuel pebble histories. 

License by Test Approach 

1. Use of the risk-based regulatory approach to fill identified regulatory 

gaps 

Using the risk-based framework outlined above, establish the detailed steps and 

information required to use the logic of the risk based framework for developing 

proposed treatments of the regulatory gaps identified. A review of accident 

scenarios leading to release of fission products needs to be performed. 

Deterministic analysis will be performed using the VSOP code and other 

transient analysis code packages to evaluate the expected plant performance in 

accident scenarios for the pebble bed plant. 

This task is critical. On the one hand, we recognize that the risk-based 

framework that we have discussed is not part of the regulations at this time, 

although it is consistent with the NRC's Option 3 for risk-informing 10 CFR Part 

50. On the other hand, the existing regulations are LWR-based and will have to 

be modified for application to the PBMR. Appendix A to 10 CFR Part 50 states: 

'The General Design Criteria are also considered to be generally applicable to 

other types of nuclear power plants and are intended to provide guidance in 

establishing the principal design criteria for such other units." This step will 

investigate how the GDCs and other regulations could be utilized in a risk-based 

framework to actually provide a basis for design and licensing of a gas reactor. 

2. Perform a PRA on the Reference Plant 

Perform a PRA on the MIT design or obtain a PRA of the PBMR from ESKOM to 

identify major accident sequences that affect safety goal attainment to identify 

critical systems requiring test. This PRA should be at least a level 2 if not level 3 

PRA in order to support the public health and safety goal philosophy. 

3. Use of the risk-based approach for development of traditional 

deterministic elements of the current regulatory approach 

Develop a risk-based technical basis for establishing risk dominant accident 

sequences (RDAS) that could form what are traditionally referred to as Design 

Basis Accidents (DBAs). The RDAS are used to formulate acceptable essential

215
safety function unavailabilities and for judging defense in depth. Defense-in

depth is a general and somewhat subjective philosophy used by regulators and 

designers to cover unquantified uncertainties. IAEA and the ACRS have recently 

begun to address this issue again. The objective of this step is to better define 

this term given the advances in analytical methods to assess risk and to allow for 

innovative new designs may have inherent physical attributes that are 

demonstrably, rather than analytically, safe. This step will examine the basis of 

defense in depth relative to gas reactors and how it can be generically applied to 

all advanced reactor designs. Our risk-based approach offers a method for 

doing this, by evaluating acceptable risks at a conservative confidence level 

rather than in terms only of expected outcomes. 

4. Develop Test Plan for Certification 

Based on the risk assessment identifying critical safety component, systems and 

structures, develop a series of subsystem and integral tests to confirm the 

performance of the components and systems as required to validate 

performance and computer codes. These tests will form the basis of the safety 

case for the technology. This test plan is expected to include the loss of coolant 

test, air and water ingress, reactivity feedback tests, control rod withdrawal, 

turbine trips, overspeed, to name a few. What will be necessary is to 

demonstrate how a safety and licensing basis could be established. 

5. License by Test Certification 

Once the test program is completed and reviewed, a general certification for the 

design can be issued. It is recognized that not all areas of performance can be 

tested using a real reactor, but as a result of the tests, the computer codes that 

are used to demonstrate performance can be validated and by extension justified 

for broader application. This area needs to be reviewed to assure that the tests 

provide an adequate justification for application over the range of conditions 

expected. 

Conclusions: 

The license by test approach to licensing is a novel method of licensing reactors. 

It provides an opportunity to deal with innovative non-water reactors in a direct 

way on a time scale that could permit early certification based on tests of a 

demonstration reactor. The uncertainties in the design and significant 

contributors to risk would be identified in the PRA during the design. 

Deterministic analysis computer codes could be tested on a real reactor. Scaling 

effects and associated uncertainties would be minimized. License by test is an 

approach that has sufficient merit to be developed and tested.

216
References

1. M. Gasparini, "Verification of Defense in Depth for Operating Nuclear Power Plants", 

draft September, 1999. 

2. F.A. Duran, A.L. Camp, G.E. Apostolakis, M.W. Golay, "Framework for Regulatory 

Requirements and Industry Standards for New Nuclear Power Plants", PSAM 5 

Conference, Osaka, Japan, November, 2000. 

3. Letter to Shirley Ann Jackson, Chairman, U.S. Nuclear Regulatory Commission, from 

D.A. Powers, Chairman, Advisory Committee on Reactor Safeguards, Subject: The 

Role of Defense in Depth in a Risk-Informed Regulatory System, May 19, 1999 

4. Sorensen, J.N., Apostolakis, G.E., Kress, T.S., and Powers D.A., On the Role of 

Defense in Depth in Risk-Informed Regulation. Proceedings of The International 

Topical Meeting on Probabilistic Safety Assessment, Washington, DC, pp. 408-413, 

1999 

5. Apostolakis, G.E., Some Issues Related to Goal Allocation and Performance Criteria. 

Proceedings of the 8 th Intemational Conference on Structural Mechanics in Reactor 

Technology, Brussels, Belgium, Paper M2 4/3,1985 

6. B. C. Beer, G.E. Apostolakis, and, M.W. Golay, "Methods for Formulation of Design 

Basis Accidents Within a risk-Informed Approach to Safety Regulation of New Nuclear 

Power Plants", PSAM 5 Conference, Osaka Japan (November, 2000). [Included as 

Appendix YYY] 

7. Report from D.A. Powers, ACRS Chairman, to R.A. Meserve, NRC Chairman. 

Subject: "Pre-Application Review of the AP1000 Standard Plant Design - Phase I." 

8. Clappisson, Metcalf, Mysen "PBMR-SA Licensing Project Organization", November, 

1999, Beijing, China 

9. Kadak,"Risk Based Regulation - The Time is Now", PSA-99, International Topical 

Meeting on Probabilistic Safety Assessment, August, 1999 

10. ESKOM, "PBMR Safety Case Philosophy", PNL-001, Rev 1, November, 2000. 

11. ESKOM, "Safety Assessment Philosophy - Support Document", PNL-009, 

November 2000. 

12. International Nuclear Safety Advisory Group, IAEA, "Defence in Depth in Nuclear 

Safety", INSAG-10, 1996. 

13. IAEA Safety Standard Series, 'The Safety of Nuclear Power Plants: Design", 

Working draft ID NS 181, (Rev. 1), Nov. 1999.

217
Figure 1. Framework for risk-based regulation and design.

Protection 

of the Public 

I 

Evaluate Risk Again 

Safety Goals 

I 

Use PRA to Quanti 

Risk and Uncertainti S

Limit Core 

Damage Frequency 

(Level 1 PRA) GO'AL:• PrnobabIif- ofI A-- Core D 

Tactics 

Implemen tation for 

Regulation & Design

amage < W0

Mitigate Releases 

of Radionucfides 

(Level 2 PRA) 

GOAL: Conditional 

Probability of Large 

Release < 0.1

Mitigate 

Consequences 

(Level 3 PRA) 

GOAL: Conditional 

Probability of Early Fatalit 

or Latent Cancer < 0. 1

Identify Required Regulation based on 

Master Logic Diagram I 

Develop Regulatory Criteria for Desi n, 

Operation, Inspection, Maintenance, 

and Testing of Required Elementsi

218

Goal 

Approach 

Strategies
Figure 2. Example master logic diagram for framework implementation. 

(for light water reactors) 

Risk > Safety I[ Goals 

Excessive 

Health 

Effects

Excessive 

Offsite 

Release

Excessive 

Release of 

Core Material

Inadequate 

Exposure 

Mitigation

Excessive 

Release of 

Non-Core Material .

Excessive RCS Pressure 

Core Damage I Boundary 

Failure

SConditional 

1 Containment 

Failure

Insufficient Insufficient Insufficient Insufficient 

Reactivity Core-Heat-- RCS Inventory RCS Heat 

Control Removal Control Removal

Initiating Failure of Mitigating 

Events Systems/Containment/ 

Emergency Response

* Internal Initiators 

for All Operating Modes

External 

Initiators

219
Table 1

Council for Nuclear Safety Licensing Approach 

For the Pebble Bed Modular Reactor (PBMR)

SAFETY REQUIREMENTS EVENT FREQUENCY SAFETY CRITERIA 

a The design shall be such to Normal operational Individual radiation dose 

Ensure that under conditions limits 

anticipated shall be those which may per annum of 20 mSv to 

Conditions of normal occur workers 

operation with a frequency up to but and 250 p.Sv to members of 

There shall be no radiation not the 

hazard exceeding 10-2 per annum. public shall not be 

To the workforce and exceeded. 

members of +ALARA+ Defense in depth 

The public. This must be criteria 

Demonstrated by 

conservative deterministic 

analysis. 

b Design to be such to prevent Events with a frequency in Radiation doses of 500 mSv 

and mitigate potential the to 

equipment failure range 10.2 to 10-6 per annum workers and 50 mSv to 

Or withstand externally or shall be considered, members 

internally originating events of the public shall not be 

which could give exceeded. 

Rise to plant damage leading +ALARA+ Defense in depth 

to criteria 

Radiation hazards to workers 

or the public. This must be 

demonstrated 

By conservative 

deterministic 

Analysis. 

c The design shall be Consideration shall be CNS risk criteria apply. 

demonstrated given to all possible event 5X10'lndividual risk 

To respect the CNS risk sequences. 10"8 Population risk 

criteria. Bias against larger 

This must be demonstrated accidents. 

by probabilistic risk +ALARA 

assessment using 

Best estimate + uncertainty 

analysis.

(CNS is the former name of the National Nuclear Regulator)

220
221

Licensing Approach for 

Generation IV Technologies 

"License By Test" 

Andrew C. Kadak 

Massachusetts Institute of Technology 

June 5, 2001

Challenges 

m Regulations focused on water 

* Knowledge of technology lacking 

m Regulatory System Rigid 

n Infrastructure to Support New 

Technology Not Developed 

n Changes in System take a long time
222

How to Introduce New Technology in 

Less than a Lifetime ? 

"* Go Back to Basic Safety Fundamentals 

"* Work Within Existing Regulatory High 

Level Objectives 

"* Use Risk Informed - Risk Based and 

Deterministic Analysis 

"* Assess Gaps in Knowledge 

"* Prioritize (risk assess) 

"* License by Test

Establish a Safety Basis 

"* Use Public Health & Safety Goal 

"* Define Plant Risks: 

- Normal Operating Plant 

- Transients 

- Accident Scenarios 

"* Identify Safety Margins 

"* Quantify Risks 

"* Show Defense in Depth
Risk Informed Approach

Goal 

Approach 

Strategies 

Limit Core 

Damage Frequency 

(Level 1 PRA) 

GOAL. Pbabfyo. 

Tactics 

llnplentmtation tO 

ImpRemgon&.. gon Regulation & Design I•o

Protection 

of the Public 

I 

E Ealuate Risk Against 

SSafety Goals 

U Pr A to Quanti [Risk ari U.,tng

Mitigate Releases 

of Radlonuclides 

(Level 2 PRA) 

GOAL- Cordt•'Jloi 

P-abawifry R.e..'ol/) o At ag .1

Mitigate 

Consequences 

(Level 3 PRA) 

GOAL' C~Xtlr .J1 

PMb~rtanof CE.0y etaT

ty Requ tred Re uis ton" based Master Logic Diagram

uOv Oup n"OulatorY ri,-reri for ue•ign, Operation, Inspection, Maintenance, 

andTesting of Required Elements.

223

Risk Informed Approach 

"* Establish a public health and safety goal 

"* Demonstrate by a combination of 

deterministic and probabilistic 

techniques that the safety goal is met. 

"* Using risk based techniques identify 

dominant accident scenarios, critical 

systems and components that need to 

be tested as a functional system

I
E 

ucr 

fuL 

ui3

(N 

(N

I
225

Review Existing Regulatory 

Structure for Gaps 

w Based on plant specific safety basis: 

- Identify existing regulations that 

apply. 

- Use risk based regulatory approach 

to fill in gaps for areas not covered. 

- Develop implementation approach to 

General Design Criteria.

Develop Traditional Deterministic 

Regulatory Approach 

"* Establish Design Basis Accidents using 

risk based techniques 

"* Develop Defense in Depth Basis Using 

natural physical attributes of designs 

"* Establish confidence levels for analysis 

using risk assessment methods
226

License By Test 

"* Build Full Size Demonstration Plant 

"* Perform Critical Tests on components 

and systems identified using risk 

informed techniques 

"* If Successful, Certify Design

Why License By Test ? 

* Needs: 

- To validate analyses 

- To shorten time for paper reviews 

- To "prove" what is debatable 

- To reduce uncertainty 

- Show Public and NRC that plant is 

safe
227

Tests Required 

"* Traditional Performance tests of 

equipment still required for reliability 

"* Use Risk Based Techniques to identify: 

- Accident Scenarios of Importance 

- Critical Systems 

- Critical Components 

"* Conduct Integrated System Tests

Examples of Tests 

"* Loss of Coolant 

"* Reactor Depressurization 

"* Natural Circulation 

"* Rod Withdrawal 

n Reactivity Shutdown Mechanisms 

"* Reactor Cavity Heat Up and Removal 

"* Selected Component Key Component 

Failures
228

Additional Tests 

m Balance of Plant Failures - turbine 

overspeed, loss of heat sink, 

compressor failures, etc 

n Control Rod Ejection (rapid withdrawal) 

s Reactor- Cavity Heat Up 

n Validate Core Physics Models 

* Validate Safety Analysis Codes and 

Methods 

* Xenon Transients

Tests Leading Up to 

Demonstration Facility Tests 

* Fuel Performance - Irradiation, post 

accident heat up, cycling 

* Air Ingress - validate chimney model for 

air ingress potential 

m Water Ingress - assess reactivity effect 

and fuel damage
229

Reactor Research Facility 

a Pebble Bed Reactor as a prototype for 

this licensing approach. 

* Built in Idaho - Full Size w/Containment 

n Implement Structured Test Program 

m Develop Regulatory Process as Part of 

Certification of Technology using RRF. 

s Research Reactor Continues as facility 

to innovate and test new technologies 

for fleet of standard designs.

Will License By Test Be Able to 

Answer All Questions ? 

* No... 

a In combination of subtier component 

tests described and the risk informed 

analysis, it should provide high 

confidence of critical safety 

performance.
230

Will License by test instill 

public confidence ? 

* Yes, 

* By having the public and the media 

observe these tests, the confidence in 

the technology and the regulatory will 

be enhanced. 

* 10- (pick a number) is not understandable or 

effective in safety discussions. 

* It will encourage development of 

naturally safe reactors.

Traditional Regulatory 

Approach 

"* Ask General Atomics for MHTGR 

"* Ask Canadians for Candu 

" Ask W about AP-600 - 1000 

"* Costly - Time Consuming - Risky 

"* Answers Not always possible to Satisfy 

NRC staff - Ask Licensees. 

* Need An Alternative to the "Bring me a 

Rock" Process. 

n This may be it...
231

Summary 

m For Non-traditional technologies, a new 

licensing approach is needed for timely 

deployment. 

* Risk Informed Techniques with Safety 

Goals Appear to meet the Need. 

n License By Test is the most direct 

means of answering difficult questions. 

a LBT should increase public confidence.
D. Powers, ACRS Member: Why do you focus on fatalities?

A. Kadak, MIT: It's an easy measure. You could talk about injuries, if you like, as a separate 

measure. 

D. Powers, ACRS Member: If we're going to learn something from accidents that have 

occurred, the most transparent consequence of Chernobyl has been radiation injuries rather 

than fatalities. Land contamination could arguably be the other thing that we've learned. Why 

not change the measures in response to things we've learned? 

A. Kadak, MIT: We could do that. I'm not limiting it. I'm just saying establish something that 

everybody is comfortable with, and I mean societal comfort. And if it talks to land, if it talks to 

injuries or if it talks to fatalities, fatalities is the one that we now have. 

G. Wallis, ACRS Member: You seem to be applying what we do today to what we might do 

tomorrow, and did you question whether we really need design basis accidents in their present 

form? Or would it be replaced by something else which might be less plant specific and be 

more effective? 

A. Kadak, MIT: The process that I would recommend is developing dominant accident 

sequences as part of the regulatory process, and don't call them design basis accident. 

T. Kress, Chairman, Future Reactor Subcommittee: What we attribute to integral tests are 

two purposes: one, to see if there's something going on that we hadn't thought of; two, to 

validate our computerized analytical tools so that they can be used in an extrapolatory sense to 

cover the things we can't do in the test. Would that be your view of what this test might do for 

you? 

A. Kadak, MIT: The needs. Why? To validate analysis. Okay? To shorten the time for paper 

reviews; to try to prove in quotes what's debatable; to reduce uncertainty, and this is very 

important; to show the public and the NRC, and I include them as the public in this case, that 

the plan is, in fact, safe. And that's what it's all about. Can we do the -- you know, can we try 

to melt the core? If we believe that we can do it without melting the core, yes. 

G. Wallis, ACRS Member: So what you should do is you should give an operator carte 

blanche to try to melt the core, and he or she will fail. Is that your test? 

A. Kadak, MIT: Depending upon the design, yes. I mean, theoretically that would be the test, 

but I would structure it more carefully than that. 

G. Wallis, ACRS Member: Are you asking for a kind of full scale LOFT test? 

A. Kadak, MIT: Full scale LOFT test, I suppose in the sense of a LOCA. There will be others 

on a facility, and one of the things it avoids is to remember the scaling issue that you've had to 

fight over? I mean, clearly the scaling issue sort of goes away if you do a full scale plan or a 

large enough scale to be able to say scaling is not a factor. 

D. Powers, ACRS Member: Let's look at that control rod ejection because it's a fun one to 

look at. The scenario that we're now worried about is one where the fuel had extremely high 

burn-up. How are you going to do that in your test?

232
A. Kadak, MIT: That would have to be outside of the reactor. There would be a whole series of 

fuel tests as part of this program. 

D. Powers, ACRS Member: Well, and the problem that plagues the rod ejection accident is an 

argument over how it propagates within the whole course. So if you do this test at the CABRI 

facility with one rod, that doesn't answer the question. You need a whole bunch of rods. 

A. Kadak, MIT: Well, I think we could do like I said, a rapid withdrawal, and we could model it 

from the standpoint of what we expect as a reactivity transfer and to see whether those codes, 

in fact can model the event. 

D. Powers, ACRS Member: I mean, that's where the argument is, is whether the codes are 

right or not, and whether they give you the right amount of heat going into the clad and not into 

damaging fuel. 

A. Kadak, MIT: Well, the first is the reactivity. Then we can go to heat, right? 

D. Powers, ACRS Member: No. This is a time scale where those two are very coupled 

together. I think you ought to look at the experience they had at the Phoebus facility, which 

was doing an experiment, which amounted to melting down 21 fuel rods, two of which were 

fresh fuel and the rest of them were irradiated, and the public response prior to the first test 

there, and how eager they were to watch that particular test. 

A. Kadak, MIT: I'll look it up. I'm not familiar with it. 

E. Lyman, NCI: So you're proposing that the test facility go with a containment which is not 

the same containment that the pebble bed is planned to have? 

A. Kadak, MIT: Only because it's a research facility. 

E. Lyman, NCI: So I've heard the argument that the passive cooling of the pebble bed is 

incompatible with a leak tight containment and it would interfere with, for instance, the design 

basis LOCA heat removal. 

A. Kadak, MIT: Well, we'd have to look at that to see whether or not and how we could make it 

compatible for this particular facility. We'd have to look at whether, in fact, we need to make 

additional modifications to the facility to accommodate the passive cooling feature. 

T. Fabian, Nuclear Waste News: It's not as exciting as melting down the core, but I'm 

wondering if as part of your conceptual design process you've done the sort of things that the 

fusion materials program has done, is looking forward to end of plant life and looking at lower 

activation materials that are easier to dispose of, possibly easier to resmelt and reuse in a 

nuclear facility, designing the plant for decommissioning using robotics and remote technology; 

is any of this played a part in the design process? 

A. Kadak, MIT: Not at this stage, although we are following what's going on in Germany as 

they're decommissioning their AVR reactor. Clearly, one of our initial objectives was to design 

a plant with decommissioning in mind, also having a lot of personal experience about 

decommissioning the Yankee Rowe plant. So I'm very sensitive to that issue. We haven't 

really looked at it, and we're not really at that level of detail yet.

233
L.E. Hochreiter, Penn State University: As an AP600 design certification survivor, I'm familiar 

with the testing that we had done and a number of questions that we got from the NRC, which 

were large. When you structure a test program, usually you build on separate effects tests to 

try to identify and create a model that you then put into an integral code, and then you use 

integral tests for verification of that model. I think one of the problems that we have in the water 

reactor technology world is that we don't have very good integral systems tests. The loft tests, 

which are the largest integral systems tests, that we've all used for a code validation, there's a 

lot of questions on the accuracy of the instrumentation, which are really measuring versus what 

you think you're measuring, and so forth. There may be a lot of potential problems for that in 

this type of a program unless it's structured very carefully, and then if you add the 

instrumentation that you want to add, you can start to distort the things that you're trying to 

measure. 

So I think that you're -- I like the idea. I think that you really have a background of tests that 

you're going to have to provide in addition to a large, full scale test where you build the 

technology so that you can have confidence then in the code that you'll use to predict the test, 

which you'll then try to run in the facility. Otherwise you may have some unpleasant surprises. 

You'll have conflicting objectives in the design of the plant versus the measurements that you 

want to make. I mean, that's the problem that LOFT had. 

A. Kadak, MIT: I think a lot of that stuff that we're talking about, some of which at least I should 

say has been done in Germany, we don't know. I don't know, first of all, and like it's sort of the 

code of record which essentially is based on no experience in the United States. We are 

learning how to use it, and it's got a lot of models built into it and has been benchmarked 

against some of the tests that they've done in Germany. 

We would hopefully use that data, disrupt your test, but I think your point earlier is exactly right. 

This is a research facility. In order to be effective, it's got to be well instrumented, and that is 

going to cost much more money than just building a straight power plan. 

W. Hauter, Public Citizen: Who should assume liability for this test? How does Price 

Anderson play into this? What kind of radiation releases is it appropriate to expose the public 

to? Should there be a public process, public hearings and so forth to determine if this is 

something that the public would want to buy into? 

A. Kadak, MIT: Let me answer the last question first. I think clearly the public has to buy into 

this process, and relative to the public hearings, I'm not all that familiar with how that would 

occur. My sense is it would have a licensing proceeding, become a licensed and experimental 

facility, and if successful, probably another licensing facility would be ready for operation. The 

Price Anderson question, I'm not an expert on Price Anderson, but, you know, depending upon 

who ultimately ends up being the builder, whether it's DOE or some private government 

partnership, those people would obviously have to pay the insurance costs for that. 

In terms of releases, you would design the test such that it would essentially address this. 

J. Slabber, PBMR: I'm not claiming or proposing that part of the PBMR demonstration unit in 

South Africa be used as supplying all of the information to Andy Kadak, but part of our objective 

as a demo unit, and it's not a prototype; it's a demonstration unit; it will be instrumented to such 

an extent that critical parameters during transients, like load rejection, may be loss of coolant,

234
could be measured, and this is not making an open statement. We've got quite a good 

technological base for proposing something like this because in an AVR, they have done loss of 

coolant simulations, as well as reactivity excursion experiments. It is documented, and they 

found, and this is, again, coming back to the integrity and the quality of the fuel, that they did 

not observe any significant increase in releases, although the core was filled with fuel, with a 

variable degree of quality and burn-ups, and they've also substantiated the reactivity 

predictions, the temperature coefficient predictions. In fact, there is a base where we can stand 

on to claim that some of the tests that are proposed in such a reactor have got some supporting 

evidence in Germany. 

A. Kadak, MIT: To the extent that it's appropriate and doable, I think many of these tests could 

be done on the south African demonstration facility. So the concept is a generic concept 

suitable for, I believe, any type of advanced reactor that has certain characteristics. 

L.E. Hochreiter, Penn State University: One of the things that we dealt with a lot in the AP 

600 was looking at uncertainty, uncertainty in the predictions, uncertainty in the analysis. Do 

you know if they've done that with these codes for the pebble bed in Germany? 

A. Kadak, MIT: I don't know. I have not been able to get at some of the qualifications.

235
A NEW RISK-INFORMED DESIGN 

AND REGULATORY PROCESS 

by 

George E. Apostolakis and Michael W. Golay 

Massachusetts Institute of Technology 

Allen L. Camp and Felicia A. Duran 

Sandia National Laboratories 

David Finnicum and Stanley E. Ritterbusch 

Westinghouse Electric Company 

Presented at the Workshop on 

Regulatory Challenges for Future Nuclear Power 

Plants 

Advisory Committee on Reactor Safeguards 

US Nuclear Regulatory Commission 

Washington, DC

237
June 4-5, 2001 

OVERVIEW 

In a project funded by the U.S. Department of Energy (USDOE) in its Nuclear 

Energy Research Initiative Program, the authors have been involved in 

formulating a new risk-informed approach for nuclear safety regulation. We 

believe that this work is important because a new regulatory treatment is needed 

both for the licensing of new non-light water reactors (LWRs), and to rationalize 

the regulation of LWRs. It is common today for the plans for new reactor 

concepts to include proposals for how they should be licensed. The existence of 

such proposals is implicit evidence that the existing regulatory structure is 

inadequate for this purpose. Similarly, attempts to "risk inform" the regulations 

governing LWRs have made only small progress because of the complexity and 

inconsistency of the existing structure. Thus, we have concluded that a fresh 

start in formulating a regulatory structure is worth attempting. This paper 

describes the fundamental concepts of that attempt. 

The overall purpose of the new approach, termed Risk-Informed Regulation, is to 

formulate a method of regulation that is logically consistent and devised so that 

both the reactor designer and regulator can work together in obtaining systems 

able to produce economical electricity safely. In this new system the traditional 

tools (deterministic and probabilistic analyses, tests and expert judgement) and 

treatments (defense-in-depth, conservatism) of safety regulation would still be 

employed, but the logic governing their use would be reversed from the current 

treatment. In the new treatment, probabilistic risk analysis (PRA) would be used 

as the paramount decision support tool, taking advantage of its ability to 

integrate all of the elements of system performance and to represent the 

uncertainties in the results. The latter is the most important reason for this 

choice, as the most difficult part of safety regulation is the treatment of 

uncertainties, not the assurance of expected performance. 

STRUCTURE OF THE NEW REGULATORY APPROACH 

The scope of the PRA would be made as large as that of the reactor system, 

including all of its performance phenomena. The models and data of the PRA 

would be supported by deterministic analytical results, and data to the extent 

feasible. However, as in the current regulatory system, the models and data of 

the PRA would require being complemented by subjective judgements where the 

former were inadequate. All of these elements play important roles in the current 

decision-making structure; the main departure from current practice would be 

making all of these treatments explicit within the PRA, therefore, decreasing the 

frequency of sometimes arbitrary judgments. 

In the intended sense the PRA would be used as a vehicle for stating the beliefs 

of the designer and regulatory decision-maker; the foundation of their decisions. 

Thus, the PRA should be viewed as a Bayesian decision tool, and be used in 

order to take advantage of its capabilities in integration and inclusion of

238
uncertainties. In order to do this, all regulations must be formulated in terms of 

acceptable levels of unavailability of essential functions, including an acceptable 

level of uncertainty (e.g., the acceptability of system performance could be 

evaluated at a stated confidence level rather than in terms of the mean value as 

is typical currently). 

Implied in this treatment is a hierarchy of acceptable performance goals. At the 

highest level societal Safety Goals would be used, supported by subgoals 

formulated at increasingly fine levels of detail as the hierarchical level of the goal 

would decrease (see Figurel).

239
GOAL

APPROACH 

PRA STRATEGIES

Public Health & Safety 

as a Result of 

Civilian Reactor Operation I 

Evaluate Risk Against 

Safety Goals 

I 

Use PRA to Quantify 

Risk and Uncertainty

I 

Limit Core 

Damage Frequency 

(Level 1 PRA) 

"'X I 

Tactics

IMPLEMENTATION 

FOR REGULATION 

AND DESIGN

I 

Mitigate Releases 

of Radionuclides 

(Level 2 PRA) 

I

Mitigate 

Consequences 

(Level 3 PRA)

Identify Required Regulation 

Based on 

Master Logic Diagram

Develop Regulatory Criteria for 

Design, Operation, Inspection 

Maintenance, and Testing 

of Required Elements

Figure 1. Framework for Risk-Based Regulation and Design 

The differences between the proposed treatment and current practices are 

illustrated in Figure 2, which shows that the use of defense-in depth and 

requiring performance margins would remain. However, the current practice of 

permitting such features to be required without justification would be abandoned; 

rather, wherever such a requirement were to be made it would also be necessary 

for the regulator to provide evidence concerning the value of the requirement and 

to reflect that value in the master PRA (i.e., if a redundancy is to be worth 

including in a system, its safety value should also be stated in the overall system 

performance analysis).

240
Operating Plants Future Plants 

(NRC) (NERI) 

Deterministic Probabilistic

Traditional ("Structuralist") 

Approach 

khRisk-Based ("Raftionali st") Approach

"* Start with current designs and • Develop new design and 

regulatory approvals regulatory process 

"• Justify risk-informed changes - Use firm probabilistic criteria 

to assure safety 

"• Defense-in-depth remains as 

primary means of assuring • Use defense-in-depth and 

safety safety margins as needed 

Figure 2. Comparison of NRC and NERI Risk-Informed Regulatory Processes 

IMPLEMENTATION 

In the licensing of any new reactor concept the degree of detail that the 

regulatory system may require will increase with the maturity of the concept (see 

Figure 3). When viewed from this perspective, it is seen that many aspects of 

the current LWR-focused system of safety regulation (e.g., general design 

criteria, design basis accidents) may not be applicable as the body of knowledge 

and experience needed for the formulation of new concepts will likely be 

unavailable in the earlier stages of their maturation. It is important to realize this 

in order that un-critical application of current requirements (e.g., a reactor 

containment building) not lead to impaired system performance or economically 

inefficient uses of resources. We suggest that some aspects of LWR-based 

regulation should not be applied to new reactor concepts without careful study. 

As far as we can tell, the proposed regulatory approach can be applied to all 

areas of nuclear safety regulation (see Figure 4), including the "cornerstones" of 

the NRC's revised reactor oversight process. In the work of our project, we have 

focused upon the traditional areas of reactor licensing: determination of initiating 

events and requirements for mitigating systems, but nothing that we have done 

indicates an inability to extend the ideas being developed to all areas of 

regulation. 

Determination of acceptable unavailability standards for a reactor's essential 

performance functions must be done on both combined general (high level) and 

reactor concept-specific bases (see Figure 5). The Master Logic Diagram (MLD) 

of Figure 5 is developed for the example of the pebble bed modular gas-cooled

241

I
reactor (PBMR). At each level of the MLD a set of performance goals must be 

formulated which are required to be consistent with those of the MLD levels 

immediately above and below the level of interest.

Development Goals and Evaluation Tools Relevant 

Stage Acceptance Evidence 

Criteria 

Initial Concept High level - Qualitative, Experiences of 

qualitative simple, other concepts, 

deterministic deterministic 

analyses 

Initial detailed High level - Quantitative - Prior quantitative 

design quantitative probabilistic, analyses 

deterministic 

Final detailed Detailed - Detailed - Prior quantitative 

design quantitative quantitative - analyses 

(design-specific probabilistic, 

subgoals) deterministic 

N-th of a kind for Very detailed - Very detailed - Prior quantitative 

a given plant type quantitative quantitative, analyses, tests, 

(design specific probabilistic, field experience 

criteria - DBAs, deterministic, 

GDCs,....) tests 

Figure 3. Stages of Nuclear Power Plant Concept Development

242
Operational Internal Reactivity Fuel Danage 

Modes Events Control 

Full Power Frequent C Accident 

Shut Down Moderate Coolant Vessel Progression, 

Other Rare Inventory Bins 

External Pressure ontainn-ent Release 

Events Control 

Frequent 

Moeqrat emperaturej Core Moderate a Control 

Ij Rare 

Spent Fuel 

Pool? 

SWaste? 

orker Riskj Sy;em Containment issionProductl 

from Model Performance I Transport ccdns IL

Figure 4. Scope of New Regulatory Scheme

243
0

Wj C-4 SUJ ~~~~LU .. . . .. .. 

t Z 

> 0 Q > >: 

Figure 5. Illustrative Logic Diagram for Pebble Bed Modular Gas-Cooled Reactor 

244
In the regulatory example used subsequently to illustrate the practicality of the 

ideas presented here an acceptable performance goal for all loss of coolant 

accidents (LOCAs) was formulated to be that 

(0.75*CDF-50) + (0.25*CDF-95) < 7 E-7 (per reactor year), where (1) 

CDF-50 is the median core damage frequency for all LOCAs, and 

CDF-95 is the 95% confidence level value of the core damage frequency for 

all LOCAs. 

This value and its formulation are used merely for purposes of illustration. A 

method for determination of the various performance goals must be developed. 

Doing this will likely be an iterative process exploring what is feasible balancing 

ideals and practicality. 

Because new reactor regulation (i.e., licensing) must be able to address the 

performance vector of different reactor concepts and to accommodate their 

respectively differing levels of knowledge, the probabilistically-based treatment 

suggested here appears to be appropriate. For regulation of actual construction 

and operations it appears to be more feasible to utilize deterministic decision 

rules, based upon the plant's PRA, and revised as needed via use of the PRA. 

DESIGN AND LICENSING NEGOTIATIONS 

In any licensing regulatory process the plant's designer develops a design which 

he/she considers to be adequate for producing electricity safely. In areas where 

performance uncertainties are large or where potential accident consequences 

so large that risk aversion is justified, the designer would have obvious incentives 

to utilize defense in depth and performance margins in the design, and to reflect 

the effects of these tactics in the evaluated performance of the plant systems. 

When this design is submitted for regulatory approval, a negotiation follows 

which leads to any design changes required for regulatory approval. Currently, 

this negotiation is conducted focusing upon how adequately the design basis 

accidents are mitigated, with some background consideration being given to the 

important risk contributors and risk sensitivities of the plant. In our new design 

and regulatory concept, this negotiation would be conducted using the PRA as 

the primary discussion vehicle. The important questions would concern whether 

the relevant functional performance goals were satisfied with sufficient 

confidence. 

Once the goals were specified, the remaining questions would concern the 

models and data used in evaluation of the un-availabilities (including 

uncertainties) associated with performance of these functions. Disagreements 

between the licensee and regulator would be focused upon the adequacy of 

models and data used in the PRA. A response to such a disagreement could 

include further defense in depth or design conservatism, but it could also include 

defense and improvements of the relevant models and databases.

245
An additional feature of this approach is that the burden upon the regulator to 

justify his challenges to the adequacy of the design would be made explicitly. 

Any design changes that the regulator thinks necessary would also be required 

to be reflected in the PRA, and the reasons for disagreement about the 

adequacy of the design would have to be formulated in terms of the adequacy of 

the PRA. Unavoidably, some of these disagreements would involve factors of 

subjective judgement. Such judgements would be required to be integrated into 

the results of the PRA, and their bases stated explicitly. This requirement would 

be an important departure from current practice where the regulator is not 

required to justify changes demanded of a license applicant. 

For example, in the recent Design Certification licensing of the AP-600 PWR 

concept, the Certification was held up by the NRC until the designers agreed to 

add an active containment spray system which is redundant to the passive 

containment cooling system of the original design. Neither the PRA nor the 

deterministic design analysis of the plant indicated the need for the active 

system, but the regulator was able to require that it be added (presumably 

because of concern that the passive system might display unanticipated modes 

of behavior) without explicit justification (it was deemed to be the "prudent" thing 

to do). 

As an illustration of how the new negotiation process would work, the designer 

before application submission would follow the process illustrated in Figure 6. In 

this process the designer would be guided by the PRA in identifying the set of 

marginally most valuable design changes to reduce functional unavailability 

values to being lower than those specified in regulations to be acceptable. The 

method of doing this would be to search for event sequences where design 

modifications would best reduce risks and/or their associated uncertainties. 

Then, once an adequate design is developed it would be submitted for licensing 

approval. 

An illustration of this process is shown in Figure 7. In this illustration, a design 

thought to be adequate by the designer is rejected by the regulator who 

disagrees with data and models used to evaluate the risks of high pressure 

LOCA event sequences in the PRA. Rather than defend the models and data of 

that portion of the PRA the designer investigates further design changes as 

summarized in Table 1 and Figure 8. It is seen that addition of greater 

depressurization capability (used to transform the high pressure LOCA into a low 

pressure one, for which adequate mitigation systems exist in the design) is 

inadequate to meet the specified performance goal because of the remaining 

risk contributions of common cause failures in the emergency diesel generator 

and cooling water systems. Only when design changes to reduce the risks 

contributed by the common cause failures does the design become satisfactory 

to the regulator. 

In this illustration, both the designer and regulator become focused upon ways to 

reduce risks and uncertainties, all of which are stated explicitly. Both parties 

have incentives to utilize good design practices, high quality components and

246
redundancy and conservatism in order to ensure that the specified performance 

goals will be satisfied. 

From this examination it is not apparent that tools of current regulation such as 

design basis accidents and general design criteria are required. They may be 

retained in regulation for purposes of convenience, but their necessity is not 

apparent. 

Rather, the needs of the new regulatory process are more concerned with ways 

of formulating a consistent set of performance goals and sub-goals, of ensuring 

that data bases and models will be of high and uniform quality, of formulating 

methods for the reproducible integration of subjective judgments into PRAs and 

for formulation of a risk-based Standard Review Plan for use by the regulatory 

staff. The tactics for creating some of these needed elements is not obvious as 

the problems involved are complex and subtle. 

The best way of satisfying the new regulatory needs appears to be investigation 

of a set of example regulatory examples, where needed improvements in a 

general approach can be revealed via inadequacies in the application. Doing 

this is time consuming and expensive. Thus, the program for such investigations 

must be initiated well in advance of the time of anticipated license applications 

for new reactors and be sustained financially. These requirements imply the 

need for a program of risk-based regulatory development to be an essential 

component of any national effort to provide new nuclear power technology 

options. 

The question facing energy technology planners is not that of whether to include 

a regulatory research component in future nuclear technology development 

efforts, but rather is one of how to make such an element sufficiently effective 

that it will permit the creation of the logically consistent and economically efficient 

licensing process required for the success of future generations of nuclear power 

technologies. The active participation of the NRC in this process is also 

essential for its success.

247
Bare-Bones Plant Design

Deterministic analyses to 

identify failure modes 4 

PRA to identify dominant 

failure modes 

Risk Informed 

Design 

Add safety features for mitigation or prevention 

of dominant failure modes 4!

Generic Risk-Driven Design 

mnt .sati-fv .ccentahilitv . itei n 

Figure 6. Schematic Diagram of the Risk-Driven Generic Design-Builds Upon 

A Bare-Bones Design, Using an Iterative Process

248
Design submittal - thought to be 

acceptableby applicant 

Regulator dispu tes assumption 

requ iresrmw data 

Re sul t- Risk of f ailu re tohbav e 

adequ ate coo Ian t levels too great 

Cause: CDF due to highp restore 

LOCA isdo minia t con trl tor

Fkx: De signer adds sdep ressurization 

cap abi lity and revises PRA 

Result: CDF due to LOCA still too 

highdu e to the high-pte ssure LOCA 

ix. 

Fix: Designer adds indlepend ent, 

redimdan t train of dep ressurization 

cap ability 

Result- CDFremains too highda e 

to support system co mmon-cau se

4 sign is re-submitted to the regu I atar

Eva luation -1: Regul ator reviewsdes ign and PRAw ith comm -c ausef ailure 

reducticon Itisdetermined thatfurther significant improvements iner suring adeqm te 

core coo lant levels canno t be acco mplisheda t a reasmab le cos t or with an ad4p ate 

degree of certainty - throughu seofa cost-1ee fit criterion.

Evaluation-2: The regu lator co mpares the achieved level of fin cticm availability, 

includingun certainty, to apre-determined standard to determine if the design is 

acceptable.

Result: Unava ilability criteria 1haveb een met and risk metricha s decreasedby a factor 

greater tdm 3. Tie design isdetermined tobe acceptable.

249

failure s (ooo ling water-pu mp and 

diesel)

Figure 7. Example of Negotiation Between Applicant and Regulator

P-
Table 1. Following the Effect of Design Modifications Upon Important Risk Metric 

Values 

Risk 

Plant Configuration Median-CDF 5% Conf. 95% Conf. Metric* 

No Depressurization 1.528E-06 3.093E-07 4.278E-06 2.216E-06 

One Division of 

Depressurization 7.086E-07 1.226E-07 1.890E-06 1.004E-06 

Two Divisions of 

Depressurization 7.055E-07 1.445E-07 1.980E-06 1.024E-06 

Depressurization and 

reduced CW CC Failure** 4.970E-07 1.008E-07 1.432E-06 7.308E-07 

Depressurization and 

reduced Diesel CC Failure 6.120E-07 1.211E-07 1.718E-06 8.885E-07 

Depress with reduced CW 

and Diesel CC Failure 4.020E-07 7.960E-08 1.290E-06 6.24E-07

Risk metric selected = (0.75 D Median CDF) + (0.25 ** CW = Cooling Water; CC = Common Cause 

1.OO E-05 . 

Bo 1DO 2D0 2D1 

1.OOOE-06

El 95% confidence CDF)

-.- Mean %- C D F I 

-095% Confidence Level 

-*L-5% Confidence Level 

m Risk Metric

Configuration

Figure 8. Effects of Design Modifications on CDF

250

I

1.OOOE-07 

1 .OOOE-08
251

ACRS Workshop on Regulatory Challenges 

for Future Nuclear Power Plants 

NERI Project on Risk-Informed 

Regulation 

June 5, 2001 

Mr. George Davis - Westinghouse 

Professor Michael Golay - MIT 

ACRS 6-2001 Wo6kp -. wS.pp1

Presentation Breakdown 

"* Mr. George Davis 

- Purpose and Overview 

- Expectations for the Future 

"* Professor Michael Golay 

- A New Risk-Informed Design and Regulatory Process 

- Example Problem 

Q)Westinghouse.m SMassachusetts Instfitute of 

Tecnology 

t~~ft..oI Sacl Ns= Laboistbl oi 

, EGAN & ASSOCIATES, C. 

ACRS 6-2001 Wow'shW -pW,.p 2
252

Purpose of Presentation

"* Describe our project and its vision of a new design and 

regulatory process 

- provide a "work-in-progress" illustrative example 

"* Explain the need for continuing the development of a 

new design and regulatory process 

- keep pace with the development and licensing of new 

reactor design concepts. 

ACRS 6-2001 WoM6shop -PwB.i 3

Substantial Reductions in Capital Costs and 

Schedule Will be Needed for New Plants 

"* Production costs (Fuel plus O&M) for operating plants 

approaching 1 cent/KW-hr 

- not much room for further improvement 

"* Future investors likely to require payback of capital 

costs within 20 years of operation, or less 

"* Capital costs must be reduced by 35% or more 

relative to large ALWRs 

- overnight capital cost below $1,000/KWe 

- construction schedule of about 3 years (or less) 

ACRS 6-2001 Woi ksp -pw8.p2p 4
Three NERI Proposals Aimed at New 

Processes to Lower Plant Capital Costs 

Program Basic Objective

Risk-informed Assessment of 

Regulatory and Design 

Requirements 

"Smart" Equipment and Systems 

to Improve Reliability and Safety 

in Future Nuclear Power Plants 

Development of Advanced 

Technologies for Design, 

Fabrication, and Construction of 

Future Nuclear Power Plants

ACRS &-2001 WCs -m,,w

Development of methods for a 

new design and regulatory 

process. 

Development of methods for 

demonstrating improved 

component and system reliability; 

including on-line health 

monitoring systems. 

Development of methods and 

procedures for collaborative, 

internet-based engineering, 

integrated design analyses, and 

improved construction schedules.

S

253

Comparison of NRC and NERI Risk

Informed Regulatory Processes 

Operating Plants Future Plants 

Deterministic h. Probabilistic 

Traditional Starting Point Risk-Based Starting Point 

The new design and regulatory process must 

be developed further to support new plant 

license applications - including Generation IV 

design concepts. 

ACRS 6-2001 WorkshM ,p-8. PP 6
254

Risk-Informed Assessment 

Interactions With Other Programs 

"* NERI framework development activities are being 

coordinated with NEI 

- NEI will emphasize the development of regulations 

- The NERI project will address the overall risk-informed 

design and regulatory process 

- Westinghouse will be an NEI Task Force member 

"* It is anticipated that a new risk-informed design and 

regulatory process will be an input to new plant license 

applications, including Generation IV reactor concepts. 

ACRS 6-2001 Worl -p*.88 

A New Risk-Informed Design and 

Regulatory Process 

Massachusetts Institute of 

Technology 

George Apostolakis, Michael Golay 

Sandia National Laboratories 

Allen Camp, Felicia Dur.n 

Westinghouse Electric Company 

David Finnicum, Stanley Ritterbusch 

ACRS 6-2001 Woisp -pwBý. a
Overall Goal of Safety-Regulatory Reform 

* Create methods to assure consistency of nuclear 

power plant applicant and regulator in performance/ 

goals for producing safe, economical power plants 

Successful 

Electricity 

Production 

I

Economical 

Production 

Major Elements: 

- Acceptance Criteria 

- Comprehensive, consistent 

assessment methods 

- Designers, operators 

ACRS 6-2001 WMWM -P.,WB.Sp

Safe 

Production 

Major Elements: 

- Acceptance Criteria 

- Comprehensive, consistent 

assessment methods 

- Regulators, designers, operators

9

255

Risk-Informed Regulatory Approach 

Fundamental Ideas 

"* Regulatory decisions are founded upon the informed beliefs of 

decision-makers. 

"* Any regulatory belief can and should be stated in a probabilistic format. 

f(x) 

x 

dXX 

Probability (x < X < x+dx) = f(x)dx 

9 Regulatory acceptance criteria must reflect acceptable best-estimate 

performance expectations and uncertainties. 

ACRS 6-2001 Woftlop -V6.VO 10
Public Health & Safety as A Result of 

Civilian Reactor Operation 

I 

Evaluate Risk Against 

Safety Goals I 

Use PRA to Quantify 

Risk and Uncertainties 

I

I 

Limit Core 

Damage Frequency 

(Level 1 PRA)

Tactics 

Implementation for 

Regulation & Design

Mitigate Releases 

of Radionuclides 

(Level 2 PRA)

Identify Required Regulation 

based on 

Master Logic Diagram

Develop regulatory criteria for 

design, operation, inspection, 

maintenance, and testing of 

required elements.

Framework for Risk-Based Regulation and Design

ACRS 6-2001 Wc"W -Pi8.1 12

256

Risk-Informed Regulatory Approach 

Fundamental Ideas.... 

"* Regulatory questions and acceptance criteria should also 

be stated within a probabilistic framework. 

"* The probabilistic framework should be as comprehensive 

as possible: 

- utilize probabilistic and deterministic models and data where 

feasible - and use subjective treatments where not feasible, 

- state all subjective judgments probabilistically and incorporate 

into the PRA, 

- require both license applicant and regulatory staff to justify 

their decisions explicitly, and 

- initiate resolution process to resolve applicant-regulator 

disagreements. 

ACRS 6-2001 WokMM .P.8 11

Goal

Approach

PRA Strategies

Mitigate 

Consequences 

(Level 3 PRA) I t
Comparison of NRC and NERI Risk

Informed Regulatory Processes

Operating Plants 

(NRC./NEI) 

Deterministic 

Traditional ("Structuralist") 

Approach 

"* Start with current designs 

and regulatory approvals. 

"* Justify risk-informed 

changes. 

"* Defense-in-depth remains 

as primary means of 

assuring safety.

ACFS £-2001 Worisn -ow ~

Future Plants 

(NERI/New NEI Task Force) 

_ Probabilistic 

Risk-Based ("Rationalist") 

Approach

"* Develop new design 

and regulatory 

process. 

"* Use firm probabilistic 

criteria to assure 

safety. 

"* Use defense-in

depth and safety 

margins as needed.

12

257

Risk-Informed Regulatory Approach.... 

"* At all conceptual stages of development, nuclear 

power plant evaluation is performed probabilistically and is supported by deterministic 

analyses, tests, experience, and judgements. 

"* Safety results of defense-in-depth, performance 

margins, best-estimate performance, and 

subjective judgements are all incorporated into a 

comprehensive PRA 

- PRA is used as a vehicle for stating evaluator 

beliefs concerning system performance 

"* The level of detail of acceptance criteria becomes 

finer as the level of concept development 

increases 

- many LWR-based regulatory constructs (e.g., 

AJQPA.ý.§) are nolapplicable to less mature

ACRS 6.2001 WWwliop -pwe ppt 13
258

Stages of Nuclear Power Plant Concept Development 

Development Goals and Evaluation Relevant 

Stage Acceptance Tools Evidence 

Criteria 

Initial Concept High level - Qualitative, Experiences of 

qualitative simple, other concepts, 

deterministic deterministic 

analyses 

Initial detailed High level - Quantitative - Prior quantitative 

design quantitative probabilistic, analyses 

deterministic 

Final detailed Detailed - Detailed - Prior quantitative 

design quantitative quantitative - analyses 

(design-specific probabilistic, 

subgoals) deterministic 

N-th of a kind for Very detailed - Very detailed - Prior quantitative 

a given plant quantitative quantitative, analyses, tests, 

type (design specific probabilistic, field experience 

criteria - DBAs, deterministic, 

GDCs ..... tests 

ACRS 6-2=0 Wwkst=3o -PW8P is1
Master Logic Diagram Peit oirmnne* Goal Level 

CONCEPT SPECIFIC 

VI

vi' 

viuE 

x+ 

ACRS 6-2001 Womimt= -PW6-oq 18

259
ýConcept-Specific Master Logic Diagram Perfomiance GoaJ Level 

SPECIFIC FOR GAS 

.. COOLED RX 

IV. 

U~ 

V

ACRS 6-2001 W"Ww~ -pwS.ppt 20

260
261

Fundamental Interactions Between License 

Applicant (or Licensee) and Regulator 

"* Should be formulated with probabilistic methods 

"* Acceptability negotiation for new license application or 

license revision 

- currently is deterministic 

- should be risk-based; completion of procedures, tools, 

and termination criteria is needed 

"* Plant construction oversight 

- can be deterministic, subject to risk-based oversight 

"* Plant operation oversight 

- can be deterministic, subject to risk-based oversight 

ACRS r,201 Wokp-p•op W 21

Basic Design and Regulatory Process 

Employed Traditionally, Remains Valid Today 

"* Designer develops a plant design that both produces power 

reliably and operates safely 

- responsible for plant safety, using high level regulatory criteria 

and policies as inputs 

"* Regulator reviews the design 

"* Designer and regulator engage in a dialog 

- specific safety features, their performance criteria, and 

methods of design and analysis 

"* Documentation is developed throughout the process 

- designer documents the design basis 

- regulator documents the safety evaluation, policies 

established, and criteria for future reviews (e.g., Reg. Guides 

and Standard Review Plans, and possibly regulations) 

ACRS 6,2001 Workshop -PWS.ppt 22
Select Design Features and Plant Arrangements

PSA Modeling performed to 

determine the likelihood of 

specific outcomes: 

- PSA provides the basis for 

design and regulatory 

compliance assessment 

-PSA models include

Risk-Informed Design and Regulatory 

Process - PRA Decision Making

-Performance and Regulatory Requirements :>

I- -

ACRS e-2001 Wc.~h~ -ov~eo*

onsideration . OT both o. aleato: .a •al~ Safety Goal ml Copiac Compliance_ " and systemic uncertainties . N 

- PSA is not totally risk based A - margins are added to 

address uncertainties F -G cense

23

262

DeterministicUnetiy 

Design AnalysesUnetny 

1 -PL --- : ~Designer---- .... ....... ... De~innpr-+-R-egul•ator . .

Schematic Diagram of the Risk-Driven Generic Design - Builds 

Upon A Bare-Bones Design, Using an Iterative Process 

B esign 

DeAtermidniticy doanalyesto 

Design 

SAdd safety features omatfalrmoe for mitigation or prevention of 

.I& 

Generic Risk-Driven Design 

must satisfy acceptability criteria 

ACRS &20:01 Wokt=cho -M8•v.90 2

ACRS 6-20N W. kW -Pw84V 23

FI
Classification of Event Sequences Within 

the Risk-Informed DBA Approach

Response Required

Initial Sequences 

Very Small Leak 

Safety Relief Valve Stuck Open 

Small Pipe Break LOCA 

Pilot Operated Relief Valve Stuck 

Open 

RC Pump Seal Failure 

Medium Pipe Break LOCA 

Large Pipe Break LOCA 

Shared Functional Challenges 

Insufficient RCS Inventory Control 

Insufficient RCS Pressure Control 

Insufficient RCS/Core Heat 

Removal

ACRS 6-20D1 Worlý;o9 -P0S. 25

263

Normal Coolant Make-Up 

Emergency High Pressure 

Coolant Injection 

Depressurization and Emergency 

Low Pressure Coolant Injection

Classes

Apportionment of a Performance Goal Into 

Subgoals 

"* Designer proposes apportionment - then negotiates with regulator 

"* Apportionment must reflect what is feasible in the design 

"* Example shows that the reliability/availability of mitigation 

systems reflects feasibility of the design 

Initiating Event Mitigation Core Damage 

Initiating Event Frequency Unavailability Frequency 

Very Small LOCA 4E-3 Iyr 1 E-4 4E-7/yr 

Small LOCA 2E-4 /yr 1E-3 2E-7/yr 

Large LOCA 4E-5 /yr 1 E-2 4E-7/yr 

Achieved Total 

Example Acceptability Criterion: Achieved Total CDF CDF due to 

due to LOCAs must be less than or equal to 2E-6 /yr LOCAs: 

1 E-6 

ACRS 6-2001 Wor ..p,,..p 26
Example of Negotiation Between 

Applicant and Regulator

ACRS 6-2c0I1 wo*ShoD -PwS•W 28

264

Example of Designer's Initial Risk

Informed Submittal to the Regulator 

"* Two safety system divisions - each contains: 

- two active high-pressure injection trains 

- one active low-pressure injection train 

- cooling water (component cooling, service water, HVAC) 

- two diesel generators 

- DC (battery) power 

"* Shared support systems 

- chemical volume control system 

- off-site power 

"* PRA Includes: 

- deterministic analyses, data, models, 

- uncertainties, inter-dependencies, and common-cause failures 

- initiator data are from documented sources (NUREG/CR

5750) 

- component failure frequencies are estimated from existing 

PRA studies (for this LWR example problem) 

ACRS 6-2001 Woakz -".Wl 27
Example of Negotiation Between 

Applicant and Regulator.... 

reductio. It is determined that further significant improveme~nts in ensuning" 

•adequate core coolant levels cannot be accormplished at a reasonable cost or wt 

an adequate degree of certainty -through use of a cost-benefit criterion. 

Evaluation-2: The regulator compares the achieved level of function availability, 

including uncertainty, to a pre-determined standard to determine if the design is 

acceptable. 

CRS 1crt-haveýl ý been metan e decreased byWa-astor 

than 3.- The design is determined to b set able 

ACRS 6-2001 WaW~ -pwS rpt 29

265

Following the Effects of Design Modifications 

Upon Important Risk Metric Values 

Risk 

Plant Configuration Median-CDF 5% Cont. 95% Conf. Metric* 

No Depressurization 1.528E-06 3.093E-07 4.278E-06 2.216E-06 

One DMsion of 

Depressurization 7.086E-07 1.226E-07 1.890E-06 1.004E-06 

Two Divisions of 

Depressurization 7.055E-07 1.445S.07 1.980E-06 1.024E-06 

Depressunzation and reduced 

CW CC Failure** 4.970E-07 1.00BE-07 1.432E-06 7.308E-07 

Depressurization and reduced 

Diesel CC Failure 6.120E-07 1.2 IIE-07 1.718E-06 8.885E-07 

Depress with reduced CW and 

Diesel CC Failure 4.020E-07 7.960E-08 1.290E-06 6.24E-07 

* Risk metric selected = (0.75 * Median CDF) + (0.25 * 95% 

confidence CDF) 

CW = Cooling Water; CC = Common Cause 

ACRS 6-2001 Wo6kto -m•.Pw 30
B0 100 2DO 2D1 2D2 2D3

Effects of Design Modifications on CDF

1.O0OE-05 

1 .OOOE-06 

1 .OOOE-07 

1 .OGOE-08

Confloguration

ACRS 6-2001 WQC,OO ow8E 31

-0-- MPan CDF 

"-0-95% Corifpda.e Loei S5% Corndience 

-.- IRsk Memrc

266

Example Problem - Results & Questions 

E Concerns about common cause failures and large uncertainties would lead designers and regulators to 

conservative design approaches 

- defense-in-depth, safety margins 

E Guidelines are needed for consistently reflecting model weaknesses in the probabilistic database 

* Consistent acceptance criteria are needed for 

negotiation guidance and termination 

E Practical implementation requires more work 

- more trial examples 

- standardized models, methods, databases - methods for treatment of subjective judgements 

- replacements for. 

- GDCs 

- DBAs (risk-dominant event sequences) 

- Standard Review Plan 

ACRS 6-2001 Wofthcp -PwS" P 32

AORS C2001 wocltm vws-m 31
267

Summary 

"* The favored approach for a new design and regulatory 

process would: 

- use risk-based methods to the extent possible 

- use defense-in-depth when necessary to address model and 

data uncertainty. 

E A new risk-informed design and regulatory process would: 

- provide a rational method for both design activities and 

applicant-regulator negotiations 

- provide a method for an integrated assessment of 

uncertainties in design and regulation 

- provide a process that is applicable to non-LWR technologies 

"* Development of a new design and regulatory process 

should be continued to support new reactor license 

applications. 

ACRS 6-2o•w"kW -pwS.ppi 33
T. Kress, Future Reactors Subcommittee Chairman: How would you deal with the issue that 

the PRAs are traditionally very incomplete? They don't deal with shutdown conditions very well. 

They don't include fires very well, and seismic even is often not treated very well -- would you 

incorporate those kinds of missing ingredients into the uncertainty of distribution? 

M. Golay: Yes, basically the way you would incorporate them is through a statement of the 

subjective judgment of those who have to assess what practice is to be used. 

D. Powers,ACRS Member: You're going to expand the capability of PRA to carry this out. One 

of the areas you're going to expand it to carry it out is in the shutdown risk. I presume that you 

have a plant here that you say is going to have some history, and during that history it's going 

to have various kinds of shutdowns, those that it planned, to do a variety of activities that are 

going to be quite different, and it's going to have an occasional unscheduled shutdown. And 

you can prognosticate all of those things, all of the different configurations of the plant that go 

on during a shutdown, a scheduled shutdown for refueling and what not. But now we don't try 

to quantify, those times and configurations, and yet you want us to do that. How is this 

possible? 

M. Golay, MIT: I would say that your task in those areas has not changed from that people 

have today; that when you consider a license application, you try to consider the spectrum of 

conditions under which the plant will be operated, and using evidence appropriate for each 

condition, judge whether it will be operated successfully. 

The development of shutdown risk analysis provides an illustration of how you do that in, say, a 

non-power state, and when you're comparing operations between those states, you, as T. 

Kress just brought out, you inevitably come to situations where the available objective evidence 

is not sufficient for you to determine which practice is better. Do you do perform maintenance 

while you're shut down or do you do it on line, for example? Again, subjective judgment has to 

come into the process. What I'm submitting is that we use that subjective judgment today. We 

simply don't spell out loud the factors the way that we're weighing the factors. What's changed 

with the approach that we're suggesting is that we state everything in probabilistic terms and 

incorporate it into the PRA. 

T. Kress, Future Reactors Subcommittee Chairman: What I'm interested in is the risk 

associated over the full lifetime of the plant. That means shutdown number e.g; 85 is going to 

take place "n" years from now and I need to incorporate into my risk assessment. Now, since I 

don't know what that shutdown consists of, what planned maintenance they're going to have 

because it hasn't even come about yet, it may even be an unplanned shutdown. How do I know 

how to incorporate the short time during shutdown, short compared to other things? That risk, 

how do I put that risk component into my risk assessment when I don't even know what it is. 

We're dealing with a change, a variable configuration in time rather than a fixed configuration, 

which is what PRAs usually deal with. How do I deal with that in a PRA? Is that something that 

needs a new PRA methodology? 

M. Golay, MIT: I would submit not, but let me go to why. The first question that may arise is 

why do you need research on regulatory reform. Why can't you just get a few people to go off 

and think in the corner for a time and come up with some proposals and then try them out? 

My experience has been that you don't know what is a good idea until you've gone through 

some feasibility attempts. That there's an iterative process at work here, and that's the heart of

268
regulatory reform research, to find out what's feasible and then from that find a good blend of 

feasible approaches consistent with an over arching logical framework. In terms of the question 

you've asked, I would suspect, without having tried to do the analysis, that, first of all, the level 

of detail that you indicate as being required is probably not necessary; that approaching it from 

the point of view of looking at safety during shutdown and trying to anticipate a range of 

conditions that you think are reasonably plausible, which is the approach we have today, will 

likely work. What I would try and do is turn the question around and try and use a real 

probabilistic treatment of the safety, but not to try and anticipate the fine detail the history of a 

plant that might occur or might not occur. 

G. Wallis, ACRS Member: Do you have a good measure of safety margin in a probabilistic 

sense? 

M. Golay, MIT: Yes. If you're using margin on let us say concerning the approach to melting 

temperature or something of that kind, what that would translate into would be to formulate your 

acceptance criterion from the design point of view at a very, very high confidence level so that 

you ensure satisfaction. 

G. Wallis, ACRS Member: But once you start saying there's a failure point, you are making 

things deterministic, which really are not. 

M. Golay, MIT: Well, I'm trying to relate it to the current design process. 

G. Wallis, ACRS Member: That's right, but I think it would be interesting to see what you could 

do with a definition of margin which got away from these ideas of having a point or -

M. Golay, MIT: Right, and what you would do, as you're hinting, is really to use a distribution 

on all of the performance limits, and that would be a natural evolution that I think we would go to 

and probably quicker than I'm anticipating. 

G. Wallis, ACRS Member: You would look at the probabilities of all of those and the 

consequences of all of those. 

M. Golay, MIT: Right. That's right. So what you expect is that if people are using the 

approach we're suggesting well, they would have natural incentives to put defense-in-depth into 

their designs partly because they could see a benefit for doing it when they make a regulatory 

submittal. The same thing would be the case with incorporating performance margin. 

T. Kress, Advanced Reactor Subcommittee Chairman: How do I decide what confidence 

level constitutes an acceptable margin? 

M. Golay, MIT: My short answer is you have to work on it. It's partly a social policy and has to 

be worked out in an iterative manner. 

G. Wallis, ACRS Member: It's an interesting idea, but it seems to me that as you learn more 

about a plant, you might actually get less detail than any kind of plan. You might really know 

what you have to worry about and you don't need all of this detail. 

M. Golay, MIT: Conceivably, and we've seen that, for example. The evolution of the passively 

based water-cooled reactors could be an illustration of that. But one reason for putting this

269
figure together is to address this question of where do the design basis accidents and general 

design criteria come into the picture. I would say that it's a tentative conclusion, not a firm one, 

that those really play a role when you get to the detailed design and later stages of evolution 

because when you try to formulate design basis accidents, you have to have a design. You 

have to have a concept in terms of which to think about and have some seasoning in terms of 

your understanding of its weaknesses, things of that kind. if you look at what we've done with 

light water reactors, we've gone through that process. 

G. Wallis, ACRS Member: Let's try to think about this. The method of design and analysis is 

going to be in probabilistic terms. You mean that every time you put a correlation in a code, 

you have to do something probabilistic with it? 

M. Golay, MIT: Only if it propagated through into your risk evaluation. 

G. Wallis, ACRS Member: It probably does. 

M. Golay, MIT: Yes. For example, if your new correlation had a different uncertainty 

treatment, you would expect that to be propagated through. That's right. 

G. Wallis, ACRS Member: Why do you need subgoals? It seems to me that if you had a plant 

that had no LOCA probability at all because of its design, then you might trade this off and be 

allowed to have more probability somewhere else if all you care about is the total. 

M. Golay, MIT: But you care about the uncertainty associated with the total as well. 

G. Wallis, ACRS Member: Yes, you do, but the total, the bottom line is the thing, not really 

how it breaks up in all these pieces. 

M. Golay, MIT: Well, I would say that another reason why you want to do this is that in the long 

run for regulatory convenience and efficiency, you probably want to formulate risk-based 

deterministic decision rules as you reach a high stage of maturity. So there will be sort of 

natural incentives to formulate subgoals as the concept matures. And that's the reason we have 

this in here, simply to illustrate that you have to go through this iterative process. 

L.E. Hochreiter, Penn State University: You talk about using best estimate performance, 

expectations and uncertainties. And you really have two kinds of uncertainties. You can have 

the plant uncertainties, but you can have the uncertainties in the model that you use to do the 

predictions, and with a light water reactor, we've got 40 years of a database, experimental 

database so that we can quantify the models and the model uncertainty so that we have a good 

handle on that. I don't know how you address that for a new design like we've been talking 

about for these Gen. IV designs where you really don't have much of a database at all. 

Mike Golay, MIT: Yes, with any concept, regardless of its level of maturity, I'll submit that as 

you try to do a risk analysis of comparing alternatives, you ultimately end up at a point where 

the available objective data reach their limits. You can find this with plenty of light water 

examples as well, that what you're really into is a situation where you -- I think always -- that's 

too strong a word because I don't have the basis for saying "always," but my experience has 

been so -- that you end up with a combination of objectively based evidence and you have to 

supplement that by your judgment. So the only suggestion that we're making is that you should 

state that in probabilistic terms and incorporate it into the PRA so that with the new concept,

270
you reach that limit much sooner than with the mature one, but that the general structure holds 

up for both. 

Larry Parme, General Atomics: You mentioned possibly replacing the DBAs with the risk 

dominant events, and overall I'm supportive of your approach, but in the licensing risk based 

approach that we did for the MHTGR, one of the things we were looking at that sort of 

approach, and we immediately ran into the problem that when you go and say that the risk 

dominant events replace DBAs, you find that certain non-risk dominant events are the only 

challenges, if you will, to certain key equipment or safety functions, and the risk dominant 

events may not demonstrate to the regulator the various ways that your safety functions are 

done. And I hope you follow what I'm saying. My question to you is: did you think about this? 

We had thought about this in the '80s, found that risk dominant events weren't a true substitute 

for DBAs and had to also use the PRA, but had to find from our event trees events that 

challenged each of the safety functions regardless of their risk dominance. 

Mike Golay, MIT: Right. Let me try and translate it though. What I think you're really saying is 

that there's a concern about the level of uncertainty associated with your risk based analysis, 

such that if you went in and claimed that you were doing very, very well, it wouldn't be a credible 

claim, and that it was necessary to, in effect, show that you could handle something tougher, is 

in some way a defense- in-depth kind of capability.

271
Summary 

Advanced High-Temperature Reactor for Hydrogen and Electricity Production 

Dr. Charles W. Forsberg Oak Ridge National Laboratory; P.O. Box 2008; Oak Ridge TN 37830-6180 

Tel: (865) 574-6783; E-mail: forsbergcw@ornl.gov 

Historically, the production of electricity has been assumed to be the primary application of 

nuclear energy. That may change. The production of hydrogen (H2) may become a significant 

application. The technology to produce H2 using nuclear energy imposes different requirements 

on the reactor, which, in turn, may require development of new types of reactors. This 

alternative application of nuclear energy may necessitate changes in the regulatory structure. 

Alternative Applications of Nuclear Energy-H 2 Production 

World consumption of H2 for the production of chemicals (e.g., CH 3OH and NH3) and the 

refining of crude oil into transport fuels is growing rapidly. Hydrogen is added to heavy crude 

oils to (1) produce lighter fuels such as gasoline and (2) remove impurities such as sulfur. As 

resources of high-quality light crude oils are exhausted, more H2 is required to produce an 

equivalent amount of gasoline per barrel of lower-grade crude oil. Because much of the H2 is 

produced from lower-value refinery streams, an economical outside source of H2 would allow 

the conversion of these hydrocarbons into gasoline rather than require their use for H2 production. As a result, the output of liquid fuel per barrel of crude oil could significantly 

increase, thereby reducing crude oil imports. Nonfossil H2 would also substantially decrease 

the quantity of natural gas that is used to produce H2, thus reducing carbon dioxide emissions. 

Currently it is estimated that 5% of natural gas is used to manufacture H2 for chemical and 

refinery use. Hydrogen consumption is increasing rapidly, and some projections indicate that 

by 2010 the energy value of the hydrocarbons used to manufacture H2 will exceed the energy 

output of all nuclear reactors in the United States. Hydrogen has also been proposed as a 

future transport and distributed-power fuel. These advanced applications would increase the H2 

demand by one to two orders of magnitude. The development of economic nonfossil H2 would 

also protect the domestic chemical and refinery from high natural gas prices that could increase 

H2 costs sufficiently to cause parts of these industries to move offshore for lower cost sources 

of natural gas. 

Hydrogen and electricity represent the only large potential markets for nuclear energy. 

Therefore, if the uses of nuclear power are to expand, reactors must be designed to efficiently produce H2. Many direct thermochemical methods are possible for producing H2 with the input of heat and water. High temperatures (800 to 1000 0

C) are required to ensure rapid chemical 

kinetics (small plant size with low capital costs) and high conversion efficiencies (-50% thermal 

energy converted to H2). A low-pressure reactor coolant is desired to couple to the low

pressure chemical plant. The development of such a reactor would also make possible better 

methods of electricity production: indirect Brayton cycles and direct thermal-to-electric 

conversion techniques. Efficient technologies for the latter process do not exist at present. 

Advanced High-Temperature Reactor (AHTR) 

If nuclear energy is to be used for production of H2 or similar applications, reactors that can 

meet the unique high-temperature requirements (800 to 1000°C) are required. One such

273
reactor-the AHTR-is described herein. The high-temperature operations also create the 

potential for very-high efficiency methods for the production of electricity. 

The AHTR would generate up to 600 MW(t) with an outlet temperatures of >10000C. The 

reactor core contains a graphite-matrix fuel and core that has the same general characteristics 

as that developed for modular high-temperature gas cooled reactors (MHTGRs). Such fuels 

have been demonstrated at temperatures up to 12000C. The AHTR fuel cycle would be similar 

to that for the MHTGR. The liquid coolant would be a molten fluoride salt (2LiF-BeF2) developed 

for molten-salt-fueled fission reactors and proposed as a coolant for fusion reactors. The 

coolant would transfer heat from the coated-particle graphite fuel to the H2 chemical plant. This 

particular salt has a boiling point of ~1400'C. Several other candidate salts exist such as 

FLiNaK (a eutectic mixture of LiF, KF, and NaF). Fluoride salts are fully compatible with 

graphite (the aluminum industry has electrolyzed aluminum fluoride salts in graphite furnaces 

for over a century to produce aluminum metal). 

The combination of the graphite fuel form and the molten salt coolant makes possible the very 

high temperatures. The low-pressure coolant reduces the need for high-temperature, high

strength materials in the external heat exchangers, compared with those required in reactors 

that use high-pressure helium or other high-pressure fluids to transfer heat. The maximum salt 

outlet temperature can be significantly higher than that for a gas-cooled reactor with the same 

graphite fuel and same peak fuel-temperature limits. This is a consequence of the heat

transfer properties of molten salt (similar to water) compared to helium. The improved heat 

transfer lowers temperature drops between (1) fuel and coolant and (2) coolant and the H2 

plant. 

The AHTR reactor has some safety systems in common with other reactors, as well as some 

unique features. Reactor power is limited by the high-temperature Doppler effect within the 

fuel. Because the molten salt expands upon heating, an additional negative moderator 

temperature coefficient is associated with coolant expansion. The reactor physics are similar to 

those of the MHTGR. In an accident, the decay heat would be conducted directly from the 

reactor core, through the reactor vessel, and then to the environment. This is similar to the 

emergency decay-heat-removal system in an MHTGR. 

The liquid coolant lowers the potential for radionuclide release by several mechanisms: (1) 

atmospheric pressure eliminates a primary driving force for radionuclide releases, reduces the 

forces that can destroy the containment or confinement system, and simplifies isolation of the 

reactor from the environment, (2) the difference (at least 4000C) between the operating 

temperature and boiling point of the salt provides a large margin before boiling occurs, (3) the 

physical properties of the coolant allow natural circulation of the coolant to provide decay-heat 

cooling, and (4) most fission products and actinides dissolve into the coolant. Significant work is 

required before the full safety implications of this type of reactor are understood and before 

such a reactor could be built. 

Regulatory Implications 

The production of alternative products using nuclear energy encompasses different safety 

considerations involving both the reactor and the energy conversion facility. The impacts of the 

reactor on the chemical plant and the impacts of the chemical plant on the reactor must both be 

considered. It implies ownership--and possibly operation-by non-utility corporations. The 

different products (H2) may require reactors with non-traditional coolants such as molten salts.

274
275

Advanced High-Temperature Reactor 

for Hydrogen and Electricity Production 

(Joint ORNL-Sandia Activity) 

Charles Forsberg 

Oak Ridge National Laboratory 

P.O. Box 2008; Oak Ridge, TN 37831-6180 

Tel: (865) 574-6783; E-mail: forsbergcw@ornl.gov 

ACRS Workshop: Regulatory Challenges For Future Nuclear Power Plants 

Advisory Committee on Reactor Safety 

U.S. Nuclear Regulatory Commission 

Washington D. C. 

June 5, 2001 

ThazubeaO~d noncpt h boacl ataa by a eonr U.S. Goa.vnwrr urt - aooj act S0= ACO•D6-A OR2725 A-caangty. . U.S. G - t rat-n, a r . raywtrf • 

kWOt, ptýa -. raoa- pohaho f-., dp a~tocao.CW slo fo US vmo. r roa F"a ba AOTURNR 

OAK RIDGE NATIONAL LABORATORY U. S. DsPAx, T.yiE.,f OF ENFRGY LT-BATTEL.E

Outline 

"* Is a nuclear-based hydrogen economy in our 

future? 

"* The Advanced High-Temperature Reactor 

(AHTR) 

- An option for hydrogen production 

- An option for electric production 

"* Regulatory implications 

OAK RIDGE NATION.AL LABORATORY 

U. S. DE.A 'ICMEY1- OF HIaRGY UT-BATTELLE
276

Is a Hydrogen Economy 

in our Future? 

(It may already be here) 

OAK RIDGE NATIONAL LABORATORY U. S. DEP.ip.T•IF,,. oF ENFRCY UrT-BATTELL.IE

Rapid Growth Is Expected 

in Industrial Hydrogen (H2) Demand 

Rapidly growing H2 demand 

- Production uses 5% of U.S. natural gas plus refinery by-products 

- If projected rapid growth in H2 consumption continues, the energy 

value of fuel used to produce H2 will exceed the energy output of all 

nuclear power plants after 2010 

"* The chemical industry (NH 3 & CH3OH) is a large consumer 

"* Changing refinery conditions are driving up the H2 demand 

- More heavy crude oils (limited supplies of high-quality crude) 

- Demand for clean fuels (low sulfur, low nitrogen, non-toxic fuels) 

- Changing product demand (less heating oil and more gasoline) 

"• If nonfossil sources of hydrogen are used, lower-value 

refinery streams can be used to make gasoline rather than 

hydrogen-reduced oil imports 

OAK RIDGE NATIONAL LAOPATORY U. S. D£PT - OP EN1moF LfT-BATTELLE
277

Increased Use of More Abundant Heavy Crude Oils Reduces 

Refinery Yields, Unless Nonfossil Hydrogen Is Used 

Input Refinery Transport Fuel 

P as t Ligh t C H 2)n D irty (su lfur, etc. ,: 

C Sweet (CH2)n (CH2)n 

Current 

Transition 

I Heavy RI1II 

Present Sour (CH0.

8)n Clean: (CH2)n Crude Oil 

Natural Gas . Hydrogen Plant 

Future Heavy Sour (CHo.8)n Clean: (CH2)n 

Crude Oil 

Nonfossil Hydrogen I 

OAK RIDGE NATIONAL LABORATORY 

U. S. DEP.,RT.mL,.tr OF ENERGY _F-,SATTELL .E OPJýLODWG•2-0-OTR

Multiple Benefits with Economic 

Nonfossil Sources of Hydrogen 

"* Increased transport fuel yields per barrel - Lower-value oil components converted to transport fuel 

rather than to hydrogen (current practice) 

- Reduced imports of crude oil and natural gas 

"* Greater use of heavy crude oils - More abundant with lower costs 

- Western Hemisphere suppliers (Venezuela, Canada, and 

the United States) 

"• Competitive chemical and refinery industry - Natural gas price increases are increasing H2 costs 

- Risk of parts of the industry moving offshore 

"• Lower carbon dioxide emissions 

OAK RIDGE NATIONAL LABORATORY 

U. S. DFPARTr.vNT]" OF ENERGY LT.-SATTELLE'
278

The Growing Industrial Demand for Hydrogen Creates a 

Bridge to the Hydrogen Economy 

Experien - T \ .\\Deveio.-.ment 

S/ / , rDistnbuted 

Transwo'- Powe: 

Chemical Demand 

OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT O- ENERGY LT-__TTEl__Ir 

ON LDWG201-1O08

Hydrogen Can Be Produced with Heat 

from a Nuclear Reactor 

"* Heat + water 4> hydrogen (H2) + oxygen (02) 

"• Nuclear energy would compete with natural 

gas for H2 production 

- Rising natural gas prices 

- Constant (level load) H2 demand matches nuclear output 

* Characteristics of hydrogen from water - Projected efficiencies of >50% 

- High-temperature heat is required: 800 to 10002C 

- Existing commercial reactors can not produce heat at these 

high temperatures 

- An alternative reactor concept is required 

OAK RIDGE NATIONAL LABORATORY U. S. DzrAirr.r•.FTr OF ENERGY UT-BATTELLE
Chemical Processes Convert High-Temperature 

Heat and Water to Hydrogen and Oxygen 

(Example: Iodine-Sulfur Process) 

Water 

Oxygen Hydrogen 

80- so ,SO2, 

S8D041 K2S04- + S02 + 2Fý0 2.HI -

H-1⁄2o+ SO, + ' 2HI + H2 SO, 

AK RIDGE NATIONAL LABORATORY 

S. DEP.XRTmE.",T OF ENELRGcy UT-SATTELLE o;•% •G 21-•2

279

0 

U.

An Advanced High-Temperature 

Reactor (AHTR)-A Reactor 

Concept for Hydrogen Production 

(Different products may require 

different reactors) 

O.AK RIDGE NATIONAL LABORATORY U. S. DEPlwrWl"7Ž.T OF ENLEGcY UT--ATTELLE
Advanced 

Coupled to a 

Reactor

High Temperature Reactor 

Hydrogen Production Facility

Hol 

Mohen

OAK RIDGE NATIONAL LABORATORY U. S. DEPAkRT.mENT Op ENERGY UT-8ATTELL 

OR. •DWG 2001-•0OR

280

Desired Reactor Characteristics to 

Produce High-Temperature Heat 

"* Low-pressure system (atmospheric) 

- Metals become weaker at higher temperatures 

- Low pressures minimize strength 

requirements 

- Match chemical plant pressures (atmospheric) 

"* Efficient heat transfer 

- Need to minimize temperature drops between 

the nuclear fuel and application to deliver the 

highest-temperature heat 

- Liquid coolant 

OAK RIDGE NATIONAL LABORATORY U. S. DEP rkiiET o- ENERGY LT--B_ _EUE
281

The AHTR Combines Two Different 

Technologies To Create an Advanced 

High-Temperature Reactor Option 

"* Graphite-matrix fuel - Demonstrated operation at an operating limit of -12002C 

- Same fuel technology planned for modular high-temperature 

gas-cooled reactors 

- Fuel geometry/dimensions would be different for molten salt 

"• Molten salt coolant (2LiF-BeF2) 

- Very low pressure (boils at -1400 2C) 

- Efficient heat transfer (similar to that of water, except it works 

at high temperatures) 

- Proposed for fusion energy machines 

OAK RIDGE N-\TIONAIL LkBORATORY 

U. S. DEPART.EŽ-T OF ENeRGY L-T-BEATTELLE

Japanese High-Temperature Engineering Test 

Reactor Fuel for 9502C Helium Exit Temperatures 

Fuel Handling Fuel Rod Hole Dowel Pin 

Annular Coolant I 

Fuel Kernel- 1 Coated Channell High Densi, PC- Fuel Plug 

Hiih Particle Graphite 

Low De PCc naity Copc 

e ~raphite ~ Sleeve 

39 mmn 

3 4 rmm0 

Fuel Compact Fuel Rod Fuel Block 

OAK RIDGE NATIONAL LABORATORY 

U. S. Df.PrTirTxr OF ENERGY UTBATTELLE 00141 WO 201.4
282

Molten Salt Coolants Allow Low-Pressure Operations at High 

Temperatures Compared With Traditional Reactor Coolants 

Boiling Point Coolant Operating Pressure 

1400C- . .... Molten Salt ---- Atmospheric 

AHTR Operating 

Temperature-,, 

883'C . ..... Sodium - ----- Atmospheric 

High Pressure To Maintain Dense 

(Efficiency) Coolant 

100°C - Water - ----- 1000-2200 psi 

-269°C Helium - ----- 1000-2000 psi 

OAK RIDGE NATIONAL LABORATORY 

U. S. DEPAiRTm.aNT OF ENERGY LUr-BATTELLE ORNL VIG ZV01-101

The Safety Case for the AHTR 

"* Low-pressure (subatmospheric) coolant 

- Escaping pressurized fluids provide a mechanism for 

radioactivity to escape from a reactor during an accident 

- Low-pressure (<1 atm) salt coolant minimizes accident 

potential for radioactivity transport to the environment 

- Minimize chemical plant pressurization issues 

"* Good coolant characteristics provide added safety 

margins for many upset conditions 

"* Passive decay-heat-removal system similar to that 

proposed for other advanced reactors 

- Heat conducts outward from fuel to pressure vessel to 

passive vessel-cooling system 

- Power limited to -600 MW(t) 

OAK RIDGE NATIONAL LABORATORY U. S. DPAITMsEŽi OF ENERGY UT-BATTELLE
283

High Temperatures Also Create New 

Options For Production of Electricity 

"* High-efficiency helium gas-turbine cycles - Conversion efficiency >50% at 10002C 

- Provide isolation of power cycle from the reactor using 

low-temperature-drop heat exchangers 

- Use advanced gas-turbine technology 

"* Direct thermal to electric production - No moving parts (solid-state) methods to produce 

electricity from high-temperature heat 

- Radically simplified power plant 

- Potential for major cost reductions 

- Longer-term option-solid-state technology is in an earlier 

stage of development 

OJK RIDGE NATIONAL LABORATORY 

U. S. DEP.xTruNtNT OF ENERGY LT-SATTrELL.E

Advanced High Temperature Reactor With 

Brayton Cycle For Electricity Production 

Reactor Heat Transfer Loop Electric Generation 

Prmary Secondary 

Molten Salt Control Salt Pump Coolant Salt Salt Pump 

(Examnple: oda 

2LijF BeF2) 

SExchangere 

Salt-Heliumeium HFuel 

SGraphite: Similar Heat ing 

to HTGR Fuel) Exchanger Water 

OAK RIDGE NATIONAL LABOR-ATORY 

U. S. DEPAXIUvfLNT OF ENERGY UT-SATTIELLE 

ORNLMGý o uo1-10uR
The AHTR May Enable the Longer-Term Option of 

Direct Conversion of Thermal Energy to Electricity 

Reactor Solid-State Direct 

Hot Thermal- To-Electric Converter 

l1000C" Molten Salt 

Molten Salt Ro ds 

(Example: Rods 

2UF-BeF,) Powr 

.•Cooling 

Solid-Stte 

Fuel Converter 

(Graphite: Similar to HTGR Fuel) 

DGE NATIONAL LABORATORY 

,'T OFT" ENERGY Lrr-U -'rElIP ONLOW•WG01-105

284

OA:K RI 

U. S. DLI

High Temperatures Create 

Development Challenges 

9 AHTR uses some demonstrated 

technologies 

- Fuels (modified HTGR fuel) 

- Coolant 

* AHTR requires advanced technology 

- High-temperature materials of construction 

- Optimized system design 

- Heat exchangers 

- Hydrogen and energy conversion systems 

OAK RIDGE NATIONAL LABORATORY 

U. S. D£mzT" ,'TN-r OF ENERGY UT-BATTELLE
285

Regulatory Implications of 

Hydrogen Production 

"* Different owners: oil & chemical companies 

- Larger than traditional utilities 

- Different perspectives 

"* Both chemical and nuclear safety must be 

considered (it is not clear where the primary 

hazard is) 

- Chemical plant must not impact nuclear plant 

- Nuclear plant must not impact chemical plant 

"* Non traditional (non-water, non-liquid-metal, 

non-gas) reactors may be preferred 

OAK RIDGE NATIONAL LABORATORY 

U. S. DEP.,LTMiA T op ER c¥y LFT-SATT'ELLE

Conclusions 

"* Economic methods to produce hydrogen from 

nuclear power may provide multiple benefits 

- Increased gasoline and diesel fuel yields per barrel of 

crude oil with reduced dependence on foreign oil 

- Long-term pathway to a hydrogen economy 

"* High-temperature heat allows for new, more

efficient methods to produce electricity 

"* Reactors with different characteristics may be 

preferred for such different uses 

- Very high temperatures 

- Low pressures 

OAK RIDGE NATION_-AL L.AOIRATORY 

U. S. 1)!? FEN-,T OF ENF.Rxcy UT-BATTELLE
286

Added Information 

OAK RIDGE NATIONAL LABORA,.TORY U. S. DEP.ART.mFNT OF ENERGcY _T-E3A T _I_

Hydrogen is Made From Natural Gas-If Gas Prices 

Remain High, a Significant Fraction of the Chemical 

and Refinery Industry May Move Offshore 

U.S. Natural Gas Prices are Rising 

(daily price. $11000 cu ft) $12 

10 

Q 

6 

4 

2 

'1997 '1998 '1999 2000 '2001 

OAkK RIDGE NATIONAL LABORATORY 

U. S. DPFrA-T.-x - ov " ENERGY UJT-BATTELLE
287

There Has Been Extensive Development of 

Molten Salt Technologies For High

Temperature Nuclear Applications 

" Initial development was for the Aircraft 

Nuclear Propulsion Program 

- Heat transferred from the solid-fueled reactor to 

the heat exchanger in the aircraft jet engine 

- Molten salts were chosen based on physical 

(pressure <1 atm.) and nuclear properties 

" Molten salts are being considered for cooling 

fusion reactors (both types) 

" Russian studies on molten-salt-cooled 

reactors 

OAK RIDGE NATIONAL LABORAITORY 

U. S. DEPAsT.mR"zr oF ENERGY UTr-ATTELLE

Vapor Pressure of 2LiF-BeF2 Is Low 

Compared To Other Reactor Coolants 

10 .. . . . 

1 0.01 Atmosphere 

.ý 0.1 . Molten Salt 

B e..-F2 C o o~la nt: 

0.01 *5 2LiF-BeF2 Liquidus 2 LF

on Temperature = 457C . ., Molten Salt S.-/ Reactor Fuel 

0- 0.001 (with uranium) 

o0.0001 

0.00001 Sodium Boils 

0.000001 , , i I . . . .. . .

400 500 600 700 800 900 1000 1100 

Temperature (0C) 

OAK RIDGE NATIONAL LABORATORY 

U. S. DF.PARTMEnT OF ENERGY UT-BATTELLE
Characteristics of Molten Salts 

"* For the proposed 2LiF-BeF2 salt, the temperature 

rise from the AHTR operating point to the boiling 

point is -4002C 

"* Several other fluoride salts could be used 

"* Natural circulation cooling is an option 

* Fluoride salts dissolve most fission products and 

actinides (basis for molten salt fueled reactor) 

* Freeze point is -4572C 

• Large industrial experience with other fluoride 

salts (aluminum production)

OAK RIDGE NATIONAL LABORATORY U. S. DE"P.,ýRni&'rr OF EN.•GY

U. S. DEPARThIEŽJT OF ENERGY rELLE

IT-SAT'

Reactor Energy Conversion 

Options

Advanced High-Temperature Reactor

Passive Decay 

Heat Removal

"Converson Options 

-Hydrogen from water 

Betedncity 

-Brayton Indirect 

Cycle 

- DrectThermo- l 

Eecric

Fuel "(Graphite: Simila 

to HTGR Fuel)

I

OAK RIDGE NATION.AL LABORATORY 

U. S. DEPAxrrmE.lT OF FN'oy

Air In-!tUT-BATTELLE 

0iiiLDWG=Q19S5

288

LLE

UT.-SA3"TELLF: ORNL DWG 2OO1-•

Cooling 

water
D. Powers, ACRS Member: It even goes beyond that because by taking out the aromatics you 

reduce the octane level -- octane rating of it, and so now you have to do more processing on 

the octanes. 

C. Forsberg, ORNL: Yes. This type of refinery has about 95 percent efficiency. That is for 

every 100 BTUs going in here you get 95. This type of refinery for every 100 BTUs you get 

about 80 BTUs out. So the refinery efficiency is dropping. 

D. Powers, ACRS Member: And those particular salts that you've got there, just about 

everything dissolves, even the things we think are nominally metals. 

C. Forsberg, ORNL: I know. This is an unusual coolant. But it's a different approach to safety 

also, and that's why I mention it because we normally don't think of coolants as fission product 

absorbers. And in this case the coolant is a fission product absorber. 

D. Powers, ACRS Member: We saw this in TMI, that you blow fission products through water. 

They stay in the water. And here all you're doing is magnifying that with a coolant that has a 

higher dynamic range than water does. 

C. Forsberg, ORNL: Yes. I think it's an important issue though because there are different 

approaches to safety also that you can think about when you go to these high temperatures 

and when you go to other coolants. 

D. Powers, ACRS Member: I think it has some interesting safety issues that are peculiar to 

itself. This is the classic problem of over-cooling accidents. Start-up is interesting. Start-up and 

shutdown, both are interesting events in this reactor. 

C. Forsberg, ORNL: What D. Powers means by start-up is that this material thaws, becomes 

a liquid at about 400 C., molten salt. So you have a system that is, on start-up when it turns to 

liquid, is already moderately warm. In fact, it's hotter than any light water reactor on start-up, 

which is not your normal way of thinking about things. 

J. Sieber, ACRS Member: I presume you pumped this molten salt around the surface. Are 

there pumps that can actually do that at these temperatures? 

C. Forsberg, ORNL: Yes. Well, we haven't done anything at this temperature. The molten 

salt reactor experiment at Oak Ridge operated at 700 C. Now, the difference is in that reactor 

the uranium was dissolved in the salt. There was not a solid fuel element. But that operated 

about a much lower temperature of 700 C., and of course, nobody has operated a salt system 

at these temperatures. 

J. Garrick, ACNW Chairman: Are you going to say anything about performance 

characteristics other than temperature and pressure? 

C. Forsberg, ORNL: We're very early in the game, and I wouldn't make any promises that we 

have any information that would be considered credible. It's very, very early in the game. 

J. Garrick, ACNW Chairman: Just cycle times?

289
C. Forsberg, ORNL: That's right. We started this effort about six or eight months ago, so 

we're very early in the game. Starting with the observation that there some -- maybe some 

demands for a very high temperature reactors, and if you have very high temperatures, how do 

you get there with the materials that may exist, and obviously you throw out water; you throw 

out sodium. 

J. Sieber, ACRS Member: To maintain the pressure, how does it accommodate power swings 

that could be pretty sever in some accident situations. 

C. Forsberg, ORNL: Yes. We're not at the point where we've investigated the details of how 

you're going to handle these types of events. 

D. Powers, ACRS Member: Have you thought about what your primary pressure boundary is 

going to be? 

C. Forsberg, ORNL: There are three obvious choices. One is a molybdenum alloy. Then 

there is some oxide dispersion stainless steels that may have the capability, and then there are 

also graphites. But we're very, very early. And all of those things are cases where people have 

shown in the laboratory that the materials are capable of doing something, but nobody knows 

whether or not they could be made on a large scale or whether you could fabricate them or 

whether you could convert this into a practical reactor design. 

So what we have is materials that are used -- we have -- there are a number of high 

temperature materials that are used in research applications that operate at these conditions 

normally, in a research environment, but have not been used in a production environment. So 

what you have is materials that, yes, some of them have been used for 40 years, but only in a 

research environment. There's a big difference between research and production. 

D. Powers, ACRS Member: There's a big difference between research environments and 

flowing, high velocity flows. The problem here is interesting. It's not carbon extraction, it's 

alloying-agent extraction. 

C. Forsberg, ORNL: That's right. That's exactly right. There is a fair amount of experience 

based up to about 700, 800 C. Above 800 C., the databases begin to get very sparse. 

T. Kress, Future Reactor Subcommittee Chairman: There wasn't any way to get the fission 

products out to the atmosphere or there didn't seem to be. The reason I say that is why 

wouldn't this be an attractive concept for just electricity generation? Because you don't have 

these extra hazards then of the chemical plant and so forth. And just by itself it looks like would 

be a pretty safe, inherently safe concept. 

C. Forsberg, ORNL: I think it has potential attractiveness. And that's worth considering, but I 

think an important other consideration is that in this particular case you may also have multiple 

markets. And it's those multiple markets that may make it much more attractive for a serious 

consideration as an advanced reactor concept. 

But clearly if you develop this, one will take a very hard look at it as a electric power producing 

reactor because those safety benefits apply to any other application as long as it doesn't have 

interface issues.

290
D. Carlson, NRC: Lithium 6 is a strong neutron absorber and produces copious amounts of 

tritium. 

C. Forsberg, ORNL: It's isotopically separated lithium. Lithium 7. If we're looking at several 

coolants, some with lithium and some without lithium. The ones that include lithium have 

Lithium 7 because otherwise the neutronics doesn't work. 

D. Carlson, NRC Staff: Well, even impurity levels of Lithium 6 would give you lots of tritium. 

In fact, in the pebble bed reactor work in Germany, where they were considering processed 

heat applications, the very small amounts of tritium on the order of 1,000 Curies per year were 

a concern in terms of getting the tritium into the product gas. 

C. Forsberg, ORNL: Yes. That's why one of the reasons we consider multiple coolants. Each 

coolant has particular advantages and disadvantages. Neutronically the lithium beryllium 

fluoride is a tremendous advantage. But the disadvantages include tritium and a couple of 

other issues. 

The sodium potassium, sodium potassium zirconium fluoride avoids that problem. It has a little 

more activity in the coolant, and has some other issues. So one of the issues in a molten salt 

reactor is which coolant you want. They all have the same general characteristics, but that's 

where the tradeoff comes on, coolant A versus coolant B. 

You're absolutely right. That's why the coolant decision has not been made and why several 

coolants are being considered. All fluoride salts, but they have different benefits.

291
Nuclear Energy Institute (NEI) Summary

Prepared by ACRS Staff for A. Heymer 

A. Heymer of NEI provided a brief discussion on the benefits of establishing a new 

regulatory framework. He suggested that a new paradigm in regulatory thinking is 

needed and stated that the reactor oversight process (ROP) serves as the appropriate 

basis for starting these discussions. He suggested that the ROP cornerstones of safety 

be used as the starting point for developing a new set of General Design Criteria (10 

CFR Part 50, Appendix A). He suggested that new operating criteria, generic risk

informed and performance-based regulations be developed with associated design

specific and regulation-specific regulatory guides

293
294

New Plant Regulatory 

Framework 

NRC ACRS Workshop on Advanced Reactors 

New Regulatory Framework 

Adrian Heymer, NEI 

(aph@inei.org, 202-739-8094) 

NEI

Benefits of Establishing 

New Framework 

* Helps establish a new paradigm of thinking 

- Not burdened by current requirements or 

interpretations 

- Provides a standard against which to set requirements 

* Provide a platform for agreement on 

principles and objectives 

- Ensures issues are focused on safety and are tied to 

defined safety objectives
295

Benefits of Establishing 

New Framework 

"- Provides basis for NRC & industry positions 

"* Improves regulatory consistency 

- Aligns regulations and oversight process 

" Use Reactor Oversight Framework as basis for 

starting industry & regulatory interactions 

- Avoids "re-invention" of framework already accepted 

by NRC 

- Cultural change burden eased 

NEI

New Plant Regulatory 

Framework 

"* Generic to all types of reactor 

"* Top-down approach based on NRC mission 

- Adequate protection of public health & safety 

"* Based on NRC oversight cornerstones 

"• New General Design Criteria 

"* Introduce General Operating Criteria 

"• Develop a new set of generic, risk-informed, 

performance-based regulations 

"* Develop design-specific and regulation specific 

regulatory guides NEI 

0.
296

Establishing a New Regulatory 

Framework for New Plants 

"* Concept -- Risk-Informed, Performance-Based 

Licensing and Regulatory Regime 

"* Proof-of-concept application(s) 

- Use License Renewal and Option 2 models 

- Minimizes hypothetical discussions 

- Definitive schedule to drive resolution process 

"* Industry effort consolidates lessons learned 

from proof-of-concept activities 

- Vehicle for supporting proof-of-concept positions 

NEI

NRC's Mission to Provide Adequate 

Protection of Public Health & Safety T 

Safety Areas 

t 

Cornerstones & Attributes t 

General Design and Operating Criteria T 

Risk-Informed, Performance-Based Regulations -T 

Design/Regulation Specific Regulatory Guides 

DRAFT NEI
m 

m 

0 

c 

0 

mn 

m 0

4 

0 

00 

M 

IK 

,ý I

"n

r1i

,z 

rq
DRAFT 

Cornerstones 

10 CFR Part 50

160 GDCs, Regulations & 

- Initiating Events -- 16 

- Mitigation (Systems) -- 46 

- Barriers -- 27 

- EP -- 3 

- Pub. Radiation Safety -- 9 

- Occupational Safety -- 4 

- Safeguards-- 4 

- Administrative -- 68 

- Financial -- 6 

- Operational -- 23

Appendices

NEI

298

Example of New Regulation 

XX.63 Plant configuration 

management 

Licensee shall assess and manage changes 

in risk that result from maintenance, 

modifications and operational activities 

that could degrade safety-significant 

functions. 

DRAFT NEI
299

Example of 

New Design Criteria 

Protection against natural phenomena 

Safety-significant structures, systems, and components shall 

be designed to withstand, or be protected from the effects of 

natural phenomena, such as earthquakes, tornadoes, 

hurricanes, floods, tsunami, and seiches without loss of 

capability to perform their safety functions. The design and 

protective features shall reflect the most severe natural 

phenomena that have been historically reported for the site 

and surrounding area, with sufficient margin for uncertainty 

related to the limited accuracy, quantity, and period of time 

in which the data have been accumulated. 

DRAFT NEI
G. Apostolakis, Chairman, ACRS: Everybody keeps saying risk informed performance 

based, but can licensing really be performance based? 

A. Heymer, NEI: I think in the context of purely the licensing action, no, but what follows on 

afterwards is. 

G. Apostolakis, Chairman, ACRS: Oh, the regulatory. 

A. Heymer, NEI: Yes. 

G. Apostolakis, Chairman, ACRS: The oversight, sure. We are not dealing with that now. 

You are dealing with licensing, aren't you? 

A. Heymer, NEI: Well, we think that if you put a new Part 53 in place that there should be 

some element dealing with operational aspects, and so that's where we see that coming in, and 

there's also a probability that if you look at the Part 52 process in ITAAC, that is akin to a 

performance based element to a certain extent. 

G. Apostolakis, Chairman, ACRS: What I'm saying is that you are overplaying it a little bit, 

unintentionally, the significance of the fact that this framework has been used in the oversight 

process. The fundamental issues are there. If you look at the report the staff developed on 

Option 3, essentially they follow the same approach, but they dare go beyond that, and I think 

you guys are a little cool towards the other stuff they did. If you look at what Golay did, well, it's 

buried in there. It's the same idea. So I think this is a good starting point, but I wouldn't 

overplay the connection to the oversight process. It's a very different regulatory problem. 

That's my impression. 

A. Heymer, NEI: That's good insight. It's good input. I'm going to take that. 

G. Apostolakis, Chairman, ACRS: What we're seeing now on the screen is the NRC 

oversight process. When you go to yours, you are adding a fourth element in the second tier, 

but how about the bottom? What happened to human performance, safety conscious work 

environment, and problem identification or resolution? Are you going to handle those in a 

different way? 

A. Heymer, NEI: Problem identification and resolution is in the quality assurance element. 

G. Apostolakis, Chairman, ACRS: I thought your -- the emphasis of your talk was going to be 

on licensing of the new concepts. But yours seems to be attacking the whole thing. 

A. Heymer, NEI: It's a regulatory -

G. Apostolakis, Chairman, ACRS: Does Exelon really worry about how the NRC will regulate 

the pebble bed after they get the license? They worry about it right now? 

A. Heymer, NEI: They worry about it right now, but if you're dealing with -- and that's why I said 

when you develop the framework, you have people like Exelon moving out and testing the 

process on a pebble bed, and there's a feedback process that comes in and you can adjust.

300
M. Bonaca, ACRS Member: Although I must say that I still am confused about what's 

different in this from the previous system. I mean I could take the previous -- the existing 

system and then put it on -

A. Heymer, NEI: From a framework perspective, not much. It's when you get down to specific 

regulations you begin to see -

M. Bonaca, ACRS Member: Okay. Well, I can understand that. I don't quite understand from 

the examples where the differences may be, and I really couldn't figure it out. But I understand 

your intent. I mean, clearly you said before that it has to be risk informed and you're looking. 

The reason why I bring it up is that we saw a number of innovative processes this morning, and 

the concern I have is that you can put in a licensing framework now that may stifle, in fact, the 

credibility of some of the innovative cultures as much as the old system stifles. 

A. Heymer, NEI: Well, when you look at the framework and you see the current regulations 

and requirements, I would agree with you. If you look at the framework and say there are 

alternative regulations or a different set of regulations, a different set of design criteria, I think 

that gives you the flexibility. 

G. Apostolakis, Chairman, ACRS: No, there is a slight problem here, I think, in the sense that 

I cannot determine what is risk significant or safety significant until I have a PRA which will tell 

me when the PRA will be based on the actual design, but now I'm supposed to use the results 

of that PRA, in fact, to create the knowledge base for the PRA. 

A. Heymer, NEI: Well, it's an iterative process. 

G. Apostolakis, Chairman, ACRS: So you start with one and do it and do it again? 

A. Heymer, NEI: Yes, and there is experience. You just don't say, "Well, I'm starting with a 

new design. What have I got?" I mean, there's -

G. Apostolakis, Chairman, ACRS: I must say overall though, Adrian, maybe it's too early in 

the process, but I, frankly, thought you were going to come up with something that's a little 

more daring. You are really sticking to the existing regulations which you have blasted in the 

past. We must be doing something right. 

A. Kadak, MIT: Let me suggest something a little more daring, and it's reestablishing the 

regulatory compact between what the regulator's job is, what the licensee's job is in terms of 

how they deal with the future protection of public health and safety from a system that is quite 

prescriptive in terms of its requirements to something that more fully puts the burden on the 

operator to meet some high level goals. 

And I'm not sure what that new relationship is, but clearly if we go to 1,000 plants, let's just say, 

in trying to build on G. Apostolakis's ten times whatever the probability is and it gets to be a 

large number, that you can't continue doing it the same way, and what new regime might be 

appropriate to protect the public health and safety in the sense of a risk informed and 

performance based system. So that addresses the inspection and addresses the enforcement 

action, as well as the standards that you apply to new technology. So that's kind of the 

comment to the NEI people as well as to the rest of us, and that is how can we improve the

301
overall process not only for design and construction and operation, but also regulation. If there 

was a question on that, you can try to answer it, but it's a new regulatory paradigm. 

A. Heymer, NEI: Yes. It's thinking ahead and saying, like just challenging the NRC relative to 

how are they going to do license renewals for 80 plants in the next five years or ten years. 

They can't. Something has to change, some trust, some new relationship, and we have to 

figure out how that will work in a legal way. 

Dana Powers ACRS Member: Well, I think they came up with a fairly effective solution. I 

mean, they've gone through the catalog to a variety of data on the age degradation, a huge 

number of topical reports, they run four or five pilots, established a template, and people were 

following the template. Based on what we saw from ANO, you follow the template and you put 

out a pretty good product, and it goes very quickly. You're not going to have 80 new concepts 

in five years. We haven't got the same problem.

302
303

ACRS WORKSHOP 

Key Regulatory Challenges for Future Nuclear Power 

Plants 

Neil E. Todreas 

KEPCO Professor of Nuclear Engineering 

Massachusetts Institute of Technology 

PM June 5, 2001 

ML.Dpof Nuclear Engineering

CHALLENGES 

FUEL AND CLAD MATERIALS - TAKEN TO HIGHER BURNUPS 

AND OPERATED AT HIGHER TEMPERATURES. 

Drivers: Longer Operating Cycles. 

Higher Temperature Primary Systems. 

Particular Challenges: 1) Reductions in Waste Toxicity and Volume. 

2) Understanding and Control of Coolant Corrosion, 

particularly role of coolant impurities. 

"* 3) Qualification of Core Loads of Billions of Fuel 

Particles. 

"* 4) New Maintenance Practices. 

M.LT. Dept of Nuclear Engineering 2
304

V 3) Questions Regarding Particle Fuel Qualification 

* How many particles, if failed at the most limiting time in core life released, would be 

required to exceed the following conditions: 

"* Dose limits for plant workers? 

"• The lowest condition on the IAEA scale of plant incidents? 

"* Protective action guidelines for the general public? 

* If the fuel particle specification is product based: 

a. What are the individual particle attributes which are controlled by the 

specification, and for each, to what levels, and allowable variation to prevent 

particle failure? 

b. What is the allowable variation in related individual particle attributes which must be 

maintained to prevent particle failure? 

* If the fuel particle specification is process based: 

a. What are the individual process variables which are controlled by the 

specification, and for each, to what levels, and allowable variation to prevent 

particle failure? 

b. What are the individual allowable variations in process variables which are sufficient 

to prevent particle failure? 

c. What is the allowable variation in related individual process variations which must be 

maintained to prevent particle failure?

M.LT. Dept. of Nudear Engineering 1

Particle Fuel - Consequences of a Process Specification 

* Critical Operator Actions now become located in the fuel 

fabrication facility. The fuel fabricator is the de facto control 

room operator. 

• Innovation in particle fuel design & fabrication processing is 

likely more costly and hence inhibited. 

MNLT. Dept of Nulear Engmineeng
Why are these items Challenges? 

New Technologies - require development of 

"* NRC staff expertise 

"* NRC confirmatory research basis 

* Design Solutions are aimed at precluding historic initiators 

• Establishment of a new risk-based regulatory 

framework will be needed. 

M.LT. Dept. of Nudear Engineering

305
REGULATORY CHALLENGES 

FOR THE LICENSING OF FUTURE 

NUCLEAR PLANTS: A PUBLIC 

INTEREST PERSPECTIVE 

Edwin S. Lyman 

Scientific Director 

Nuclear Control Institute 

ACRS Advanced Reactor Workshop 

June 5, 2001 

REGULATORY CHALLENGES 

NRC licensing of advanced plants must 

ensure that these economic imperatives do 

not have adverse impacts on 

- Safety 

- Risk of radiological sabotage 

- Waste management and disposal 

- Non-proliferation 

- Full opportunity for public participation

306
EXAMPLE: PBMR 

* PBMR characteristics fundamental to its 

economic viability represent significant deviation 

from traditional "defense-in-depth" 

- Lack of pressure containment 

- Significant reduction in safety-related SSCs 

- Reduction in EPZ radius by a factor of 40 (exploits 

regulatory exemption for HTGRs) 

- Greatly increased reliance on fuel integrity under 

accident conditions for protection of public health 

* ACRS (1988): "unusually persuasive argument" 

required to justify "major safety tradeoff' 

PBMR FUEL PERFORMANCE AND 

SAFETY GOALS 

Source terms must be accurately determined for a 

full range of potential accidents 

- Pebble performance very sensitive to initial conditions 

-- relationship poorly understood 

- Robustness of PBMR fuel is being oversold 

significant fission product release (several % of Cs 

inventory) can occur at 1700-1800°C) --- hundreds of 

degrees below fuel degradation temperature 

- Quality control is paramount --- BNFL involvement in 

South African fuel fabrication plant suggests that a fuel 

quality control programmatic ITAAC is necessary

307
PBMR SAFETY GOALS 

" Safety goals need to be reexamined for advanced 

reactors 

- Current goals not conservative enough --- could still be 

met by reactors today with containments removed! 

- "Large release fraction" if EPZs are reduced 

"• Accident frequencies that could result in LR must 

be accurately calculated 

- Design-basis LOCA --- safety margin may be too small 

- Air or water ingress 

"• System upgrades may be necessary to meet goals 

- secondary coolant system (MIT vs. Eskom) 

- advanced fuel coating materials (i.e. ZrC) 

RADIOLOGICAL SABOTAGE--

THE "SHOW-STOPPER"? 

Providing adequate physical protection to defend 

plants against sabotage has proven to be a major 

challenge: 

- 50% of U.S. nuclear plants failed force-on-force 

(OSRE) testing of plant security in 2000 

- At Exelon's Quad Cities plant, "deficiencies in the 

licensee's protective strategy enabled the mock 

adversaries to challenge the ... ability to maintain core 

cooling and containment" (NRC, October 18, 2000)

308
RADIOLOGICAL SABOTAGE (cont.) 

" No nuclear system can be rendered "inherently 

safe" from radiological sabotage 

- Deliberate graphite fire in PBMR remains possible even 

if accidental fire is incredible 

- Reduction in security staffing requirements for PBMRs 

not technically justifiable 

- Systems with in-situ reprocessing plants (S-PRISM) 

would be especially attractive targets 

" ACRS (1988) recommended that NRC develop 

guidance for incorporating sabotage resistance 

into advanced designs --- need early involvement 

of Reactor Safeguards staff 

PBMR WASTE DISPOSAL 

"• Final waste disposal may be the single largest 

obstacle to nuclear power expansion 

Spent pebbles create a huge waste problem: per 

MWD, compared to spent LWR fuel: 

- Volume and weight are about 10 times greater- with 

proportionate increase in storage and transport 

requirements 

- Carbon-14 inventory is 10-20 times greater --- problem 

for unsaturated repository like Yucca Mountain

309
PUBLIC ACCEPTANCE 

* New facility siting is a great challenge: 

- Favors new plants at existing sites in areas of broad 

public support 

- Trying to greatly increase number of nuclear plant sites 

is a losing strategy --- but there is little advantage in 

modularity if available sites remain highly limited 

- Favors minimization of transport of nuclear materials 

• Public opposition may only be deterred with a clear 

commitment to maximize safety: 

- Favors "gold-plating" nuclear plants 

- Inconsistent with attempts to eliminate containment, 

reduce emergency planning, etc 

PUBLIC ACCEPTANCE (cont.) 

* Aggressive licensing schedule proposed by Exelon 

for PBMR (construction to begin in 2004, 

operation in 2007) will only antagonize 

antinuclear groups now mobilizing 

"* "License by test" is just a PR move --- unlikely to 

be adequate to resolve all safety issues to NRC 

satisfaction 

* Better to proceed more cautiously and make sure 

that full resolution of all technical concerns is 

achieved

310
THE FUNDAMENTAL DILEMMA 

OF NUCLEAR POWER EXPANSION 

* Without ratepayer or taxpayer subsidy, no new 

nuclear plants will be built unless they can 

successfully mimic the desirable economic 

features of gas turbines: 

- low capital cost 

- short construction time 

- modularity and ease of distribution 

* Can this be done safely? Or is nuclear 

technology incompatible with these objectives? 

ion 

""col 

, "

311
was higher than that of I"Cs, which was consistent with 

the previous work.10-1 3 Although the inventory is small, 

the release of "'Ag would be iroublesome in mainte10' 

10-1 

~ 2 = 10" 

0.3 

10 

U. 

U- *

10

0 50 100 150 200 250 300 

Heating time (h) 

Fig. 2. Time-dependent fractional releases of fission products 

during the ACT3 heating test at 170(0C for 270 h. ob

tained by the on-line measurements of fission gas re

lease and intermittent measurements of metallic fission 

product release. 

NUCLEAR TECHNOLOGY VOL 131 JULY 2000

ticles. To compare the irradiation performance of the in

dividual particles, activity ratios, not activities, were used 

to account for variations in kernel size and to minimize 

10 

10 

100 

1o Ag 

10' 5 100 C 5S 20 5 

ro u relaseSb 

Kr 3E9u 

0 3 

104 

0 so 100 150 200 250 

Heating time (h) 

Fig. 3. rime-dependent fractional releases of fission products 

during the ACT4 heating test at 1800T for 222 h, ob

tained by the on-line measurements of fission gas re

lease and intermittent measurements of metallic fission 

product release. 

39

312
L. E. Hochreiter, Penn State University: Some of these designs are looking at not having a 

containment, and then I think you have issues. Today in the light water area, really failed fuel is 

a utility or an operator concern, and it's a vendor concern, and you're very, very careful about it 

because obviously if you want to sell fuel, you don't want it to fail. So it's a problem that solves 

itself. But you've got a containment around the plant. In some of these designs you don't have 

a containment, and I think it could be more of a problem. 

N. Todreas, MIT: Okay. First, let me answer I'm not promoting either a process or a product 

specification. What I am doing is asking whether it is going to be a process or a product 

specification, and then developing a line of questioning along each. 

L. E. Hochreiter, Penn State University: Neil, on your process control, are you envisioning a 

control process where you can try to control each, on these particles, each layer in this 

thickness within a specified amount or the total product as it comes out? 

Because I don't see how you control each layer, and if you control on the total product that 

comes out, if it doesn't come out right, and you won't find that out probably until you operate, 

then you've got a problem. 

However, now, in addition though the way you ask the words, a process specification means 

that you control the process of every manufacturing step. So you may have a process where 

you're doing the coating, but you don't go and measure the coating or sample the coating. 

What you do.is you control the attributes of the fabrication process. How do you know you meet 

your criteria if you don't go and measure? 

N. Todreas, MIT: No, no, because what you do in the qualification stage, you take the product 

that comes out; you put it in the reactor; and you'd better make well sure it can take the burn-up 

with a failure criteria over whatever your design lifetime is. 

L. E. Hochreiter, Penn State University: At some point you're going to have to have gone 

through and verified that whatever your process is gave you the product that you wanted. 

Neil Todreas, MIT: There's a tremendous amount of radiation data on this particle fuel. If you 

can pin down the process that it was made to and link it to the data, then you can say you 

identified the process, and then you can basically duplicate it and keep going. That's the 

burden the applicant is going to have.

313
ENDNOTES

The ACRS Subcommittee on Future Reactors met on June 4-5, 2001, at 11545 

Rockville Pike. The purpose of this meeting was to discuss regulatory challenges for 

future nuclear power plants. 

The Subcommittee received no written comments or requests for time to make oral 

statements from members of the public regarding the meeting. The entire meeting was 

open to public attendance. M. Markley was the cognizant ACRS staff engineer and 

Designated Federal Official for this meeting. The meeting was convened at 9:00 a.m. 

and recessed at 7:15 p.m. on June 4. The meeting was reconvened at 8:30 a.m. and 

adjourned at 5:50 p.m. on June 5. During the course of the meeting, ACRS members 

Apostolakis, Leitch, Powers, and Sieber and ACNW member Garrick announced that 

they have conflicts with certain presentations made to the Subcommittee. 

PARTICIPANTS 

ACRS/ACNW

T. Kress, Subcommittee Chairman 

G. Apostolakis, ACRS Chairman 

M. Bonaca, ACRS Member 

P. Ford, ACRS Member 

G. Leitch, ACRS Member 

D. Powers, ACRS Member 

W. Shack, ACRS Member 

J. Sieber, ACRS Member

Principal NRC Speakers 

R. Barrett, NRR* 

E. Benner, NRR 

A. Cubbage, NRR 

J. Flack, RES* 

M. Gamberoni, NRR

R. Uhrig, ACRS Member 

G. Wallis, ACRS Member 

J. Garrick, ACNW Member 

J. Larkins, ACRS Staff 

J. Lyons, ACRS Staff 

M. Markley, ACRS Staff 

R. Savio, ACRS Staff

T. Kenyon, NRR 

A. Rae, NRR 

S. Rubin, RES 

A. Thadani, RES 

J. Wilson, NRR

Principal Presenters and Speakers

J. Slaber, PBMR Demonstration Project* 

M. Carelli, Westinghouse Science & Technology 

G. Davis, Westinghouse Electric Corporation 

C. Forsberg, ORNL* 

M. Golay, MIT*

W. Magwood, DOE 

T. Miller, DOE 

L. Parme, General Atomics 

A. Rao, GE Nuclear Energy* 

R. Simard, NEI

315
W. Hauter, Public Citizen 

A. Heymer, NEI* 

S. Johnson, DOE* 

E. Lyman, NCI*

NRR 

RES 

DOE 

GE 

MIT 

NCI 

NEI 

ORNL 

PBMR

W. Sproat, Exelon Generation 

N. Todreas, MIT 

R. Versluis, DOE

Office of Nuclear Reactor Regulation 

Office of Nuclear Regulatory Research 

U.S. Department of Energy 

General Electric 

Massachusetts Institute of Technology 

Nuclear Control Institute 

Nuclear Energy Institute 

Oak Ridge National Laboratory 

Pebble Bed Modular Reactor

316
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., 

NRCM 1102, and Addendum Numbers, if any.) 

3201. 3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) 

NUREGICP-0175 

2. TITLE AND SUBTITLE 

Proceedings of the Advisory Committee on Reactor Safeguards 3. DATE REPORT PUBLISHED 

Workshop on Future Reactors MONTH YEAR 

June 4 - 5 , 2001 December 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

Technical 

7. PERIOD COVERED (inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, 

provide name and mailing address.) 

Advisory Committee on Reactor Safeguards 

U. S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above', if contractor, provide NRC Division, Office or Region, uS,. iuciear Regulatory ,ommtsu,,n 

and mailing address.)

Same as 8. above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Because of the large amount of regulatory activity that is anticipated for licensing future reactor concepts, the ACRS decided to 

hold this workshop on "Regulatory Challenges for Future Reactor Designs." The workshop was held primarily for the benefit of 

the Committee C to acquaint the members with the various design concepts and to identify potential regulatory and policy issues 

for which ACRS may be called upon to give advice to the Commission. It was also believed that the workshop would be of 

benefit to the NRC staff as well as to the industry in getting an early dialogue started on the possible regulatory approaches to 

licensing future reactor designs. These designs are expected to be significantly different from the LWRs which are the primary 

focus of the current regulations and regulatory system. 

The primary purpose of the workshop, as indicated by its title, was to identify the regulatory challenges. A list of such 

challenges identified by the workshop was developed from the workshop notes, the various presentations, the panel 

discussions, and the question and answer sessions.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

unlimited 

AP1 00 IRIS 14. SECURITY CLASSIFICATION 

Department of Energy LWRs (This Page) 

ESBWR MHTGR 

Future reactor designs PBMR 

Generation IV (This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89) This form was electronically produced by Elite red era J ,r, , c,.
Federal Recycling Program
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300