ML19327C173

Summary of 891025 Meeting W/Util Re Issues Connected W/Plant Reactor Defueling.List of Attendees,Presentation Matl & Rev B to EE-DEC-0022, Fort St Vrain Neutron Source Analysis Using DIF3D, Encl

ML19327C173
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/13/1989
From:	Erickson P Office of Nuclear Reactor Regulation
To:	Weiss S Office of Nuclear Reactor Regulation
Shared Package
ML19327C174	List: ML19327C173
References
TAC-73125, NUDOCS 8911210068
Download: ML19327C173 (30)

Text

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/ga M:g'g UNITED STATES a

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I' NUCLEAR REGULATORY COMMIS$10N n

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November 13, 1989 e

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Docket No. 50-267 5

MEMORANDUM FOR: Seymour H. Weiss, Director Non-Power Reactor, Decomissioning and Environmental Project Directorate l

Division of Reactor Projects - 111, IV, Y and Special Projects FRON:

Peter B. Erickson, Project Manager Non-Power Reactor Decomissioning and i

Environmental Project Directorate i

Division of Reactor Projects - !!!,

IV, Y and Special Projects i

SUMMARY

OF MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS FORT ST. YRAIN (FSV)

REACTORDEFUELING(TACNO.73125)

Introduction i

This meeting was requested to further discuss issues connected with FSV reactor defueling. The attendees at this meeting are listed in Attachment 1.

I The material presented by PSC is provided in Attachment 2. is I

the PSC resort EE-DEC-0022 Rev B concerning the neutron count rate analysis discussed herein.

Current Status.of.Defueling.and.Decomissioning I

PSC's plans to defuel FSV are currently uncertain because their ability to ship fuel offiste in the near f uture is potentially not available. PSC only has limited (one third of core) capacity for fuel storage on-site. Long term resolution of this problem, including the options of construction of on-site storage, may take several years.

However, PSC is ready to initiate the defueling operation with the intent of placing the core in a safer configuration. The fuel handling machine (FHM) has been refurbished.

It will be modified shortly to work with a personal computer interim control system. An upgraded control system is also being developed.

PSC has on-site sufficient dunury fuel blocks and poison pin material and new metal plenum elements to start the reactor defueling.

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CONTACT:

P. Erickson, PDNP/NRR N

49-21101 s

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1 Seymour H. Weiss )

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Alternative plans for decomissioning are also being evaluated which are different from the preliminary decomissioning plan submitted to the staff.

They involve an early dismantlement option. A final decision on this option will be made during the su mer of 1990.

)

Defueling. Count Rates PSC reviewed their analysis of expected defueling count rates.

(See ). The PSC analysis is based on a complex model evaluation with the t

model being calibrated against data from earlier refuelings. The key factor in assuring an adequate count rate is the ability to add new neutron sources to the core. Additionally, boronated material between the core and the detector can be removed and replaced with equivalent non boronated material.

PSC's analysis shows an adequate count rate will be maintained at the detectors. The calculations were independently reviewed by PSC and by GA Technologies.

Other Defueling issues It was agreed that PSC could resolve the issue of a rod withdrawal accident by

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further discussion of their administrative controls. The goal of this discussion would be to demonstrate that a rod withdrawal accident was not a credible event, since strict controls were maintained over the control rod drive mechanism power sources.

i The reserve shutdown system (RSS) was also discussed.

PSC agreed to provide calculations showing how the RSS could deal with credible reactivity control i

problems arising from failures of the control rod drives.

PSC response to these issues was received on Novenber 6,1989.

Other Defueling. Sequences PSC discussed their preliminary evaluation of other defueling sequences. These l

sequences would involve removing the most reactive core segments to the fuel

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storage wells. At that point the balance of the core could be demonstrated to 4

be subcritical with k(eff) less than 0.95 with all rods withdrawn. PSC was still evaluating these alternatives at this time.

Other. Issues PSC informed the staff that they were making changes to certain qualified equipment under their equipment qualification program. Generally, this would i

mean that certain equipment no longer needed for safe operation would not have its qualifications maintained.

It was agreed that PSC would provide the staff with an information letter covering this issue.

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November 13, 1989 j

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Seymou'r H. Weiss.

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'In Addition, PSC was planning to make certain char.ges to the plant's safe i

'L shutdown systems that are required under 10 CFR Part 50, Appendix R.

PSC j

proposed to meet with the staff at the plant site to discuss these changes. A I

meeting has been scheduled for November 16 1989 at fort St. Vrain.

.)

i Original signed by:

Peter B. Erickson, Project Manager Non. Power Reactor, Deconmissioning and Environmental Project Directorate

.l Division of Reactor Projects. III, IV, Y and Special Projects Attachments:

As stated j

cc w/ attachments:

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See next page

'DISTRIBUT!ON

55HETTfW"3 NRC PDR Local PDR PDNP r/f E. Jordan L

S. Weiss P. Erickson OGC ACRS(10)

P. Erickson E. Hylton J. Sniezek-Plant file J. Sharkey T. Westerman D. Fieno L. Kopp Orql c

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NAME tE, ton

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.sr:SWeiss DATE 311/p/89

11/l3/89

11/f)/89 i

r 0FFICIAL RECORD COPY Docunent Nane:

FSV 9014 l

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Seymour H. Weiss t

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In Addition, PSC was planning to make certain changes to the plant's safe shutdown systems that are required under 10 CFR Part 50 Appendix R.

PSC proposed to meet with the staff at the plant site to discuss these changes. A meeting has been scheduled for November 16, 1989 at Fort St. Vrain.

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Peter B. Erickson, Project Manager l

Non-Power Reactor, Decommissioning and i

Environmental Project Directorate i

Division of Reactor Projects - 111, IV, Y and Special Projects i

i Attachments:

i As stated i

cc w/ attachments:

See next page t

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Mr. A. Clegg Crawford Public Service Company of Colorado Fort St. Vrain l

cc' Mr, D. W. Wares 60urg Manager Robert M. Quillen, Director l

Nuclear Engineering Division Radiation control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue l

P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201-0840 Mr. David Alberstein, Manager Mr. Charles H. Fuller Fort St. Vrain Services Manager, Nuclear Production GA International Services Corporation and Station Manager Post Office Box 85608 Public Service Company of Colorado l

San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. BreyIng and Resource Mr. P. F. Tomlinson, Manager Manager Nuclear Licens Managenent Division Quality Assurance Division i

Public Service Company of Colorado Public Service Company of Colorado P. O. Box 840 16805 Weld County Road 19-1/2 Denver, Colorado 80201-0840 Platteville, Colorado 80651 j

Senior Resident inspector Mr. D. D. Hock U.S. Nuclear Regulatory Comission President and Chief Executive Officer P. O. Box 640 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Kelley, Standfield & 0'Donnell ATTN: Mr. J. K. Tirpey Comitment Control Program Public Service Company Building Coordinator l

Room 900 Public Service Company of Colorado i

550 15th Street 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80202 Denver, Colorado 80211 Regional Administrator, Region !Y U.S. Nuclear Regulatory Comission 611 Ryan Plaza Drive, Suite 1000 l

Arlington, Texas 76011

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Chairman, Board of County Comissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency l

1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413

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Attendees at PSC/NRC Meeting i

i October.25. 1989 t

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ORGANIZATION K. L. Heitner NRC/NRR l

D. Alberstein General Atories H. L. Brey PSC l

t Don Warembourg PSC

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Richard Nirschl PSC l

Russell Sherman PSC M. H. Holnes PSC r

l T. F. Westerman NRC/RIY j

l Daniel Fieno NRC/SRXB j

Larry Kopp NRC/SRXB l

Pete Erickson NRC/NRR Seymour Weiss NRC/NRR e

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i, MEETING AGENDA NRC/P50 MEETING 10 25 89 I.

INTRODUCTION

- INTRODUCTORY REMARKS, PSC

- MEETING OBJECTIVES

- INTRODUCTORY REMARKS, NRC i

CURRENT STATUS 111. COUNT RATE ENGINEERING EVALUATION OVERVIEW DISCUSSION IV. NRC QUESTIONS ON DEFUELING SAR V.

DEFUELING/ ALTERNATIVES VI. OPEN DISCUSSION l

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NRC MEETING OCTOBER 25, 1989 o MEETING OBJECTIVES e PROVIDE DEFVELING STATUS e RESOLVE COUNT RATE ISSVE, EXISTING SAR e CLARIFY AND ESTABLISH APPROACH AND LOGISTICS NRC QUESTIONS ON DEFUELING SAR e OBTAIN NRC INPUT / THOUGHTS ON DEFUELING ALTERNATIVES

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i PRESENT STATUS i

DEFUELING ISSUES l

t e MEETING, DOE / GOVERNOR ANDRUS (SEP 9,1989) e EXPRESSION OF CONCERN ABOUT FSV FUEL e GOVERNOR ANDRUS LETTER (SEP 15,1989) i a ACCEPTANCE OF ANY FSV FUEL FOR LONG TERM l

STORAGE IN IDAHO IN JEOPARDY l

1 e PSC CE0/VP NUCLEAR OPERATIONS MEETING WITH t

COLORADO GOVERNOR SEP 29, 1989 e PSC MEETING INTERNAL (OCT 2, 1989) e DISCUSS / DEFINE STRATEGY AND ALTERNATIVES e PSC/WINCO MEETING ICPP (OCT 6, 1989) f a READINESSOFGRAPHITESTORAGEFACILITY(GSF) e MEETING 00E/ GOVERNOR ANDRUS (OCT 11,1989) e FSV FUEL SHIPMENTS TO ICPP STILL IN JEOPARDY r

a WHITE PAPER IN PREPARATION e MEETING PSC/ SECRETARY WATKINS STAFF l

a PSC CEO & VP NUCLEAR OPERATIONS TO MEET WITH UNDER SECRETARIES i

e MEETINGS WITH GOVERNOR'S STAFF AND STATE REPRESENTATIVES ON TRANSPORTATION, SHIPPING AND ISFSI g

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s PRESENT STATUS DE FUE L I N G/DEURRT55TUh7EONV ER $ 10N e PSC WORKING FUEL DECK PREPARATIONS AND MODIFICATIONS L

TO BEGIN DEFUELING ON OR ABOUT NOV 27, 1989 e FUEL HANDLING MACHINE UPGRADES IN PROGRESS o SUFFICIENT DUMMY BLOCKS ON SITE TO BEGIN DEFUELING e METAL PLENUM ELEMENTS (NON BORONATED) EXPECTED ON SITE IN EARLY NOVEMBER e DEFUELING SAR AND NECESSARY TECHNICAL SPECIFICATIONS HAVE BEEN SUBMITTED TO NRC e FIRST ROUND OF QUESTIONS ON THE PRELIMINARY DECOMMISSIONING PLAN IN WORK FOR DEC 1 RESPONSE e REQUEST FOR PROPOSAL FOR EARLY DISMANTLE AND CONVERSION IN FINAL REVISION FOR ANTICIPATED BID SOLICITATION LAST WEEK IN NOV, 1989 e FIRST ISFSI MEETING WITH NRC ON OCT 13, 1989. WORK IN PROGRESS FOR DESIGN, SAR AND ENVIRONMENTAL ASSESSMENT FOR SIX SEGMENT ISFSI WITH OPTION TO REDUCE TO ONE SEGMENT ISFSI

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MAINTAIN ADEQUATE START-UP CHANNEL COUNT RATE t

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A CREDIBLE MEANS OF ACHIEVING CORE CRITICALITY i.

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t, KEY ELEMENTS OF THE EXISTING FSV START-UP CHANNEL DETECTOR SYSTEM i

o TWO SOURCE RANGE NEUTRON DETECTORS LOCATED IN PENETRATION WELLS WITHIN THE UPPER PCRV CONCRETE.

o CF-252 NEUTRON SOURCES LOCATED IN THE TOP LAYER OF FUEL IN THE CORE, CREATING A SHUTDOWN MODE COUNT RATE OF AT LEAST 4.3 COUNTS /SECOND 7.N EACH DETECTOR.

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1 THE FSV NEUTRON SOURCE DIF3D ANALYTICAL MODEL l

o BASED UPON AN EXPLICIT 1/2 CORE 3-D

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DIFFUSION MODEL OF THE FSV ACTIVE CORE j

AND UPPER REFLECTORS t

o UTILIZES THE INHOMOGENOUS INTRINSIC l

NEUTRON SOURCE OPTION OF DIF3D l

o DETECTOR COUNT PREDICTIONS ARE BASED UPON EXPERIMENTAL BENCHMARKS h

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EE-DEC-0022 REV 8

,g-Page 12 of 39 l

Figure III.1 3-D 1/2 Core 0!F30 FSV Model for Detector / Source Analysis cm l'

367.0 337.0 Graphits "Datector" Material 1025 bt'{$

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I Boronated Reflector 1020 317.2

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30 wtl Rodded 40 wtl Rodded Reflector Reflector (Unrodded Reg.1018)

(Unrodded Reg. 1018) 247.9 237.9 unr wided Ton Reoleetar 1017 218.1 Regions i Regions 883-1014 l 846-882 2

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y Reflective Uncontrolled Controlled Boundary Fuel Regions Fuel Regions Conditions

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F#0FOSED REV 8 i:

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NEUTRON SOURCE LOCATIONS

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REY B Page 19 of 39 Table IV.1 Startup Channel Count Rate Data From the FSV Third Refueling i

1) Surveillance Data Core Configur tion SUC1 (CPS)

SUC2 (CPS)

Initial (Prior to Refueling) 31.5 7.3 Prior to Refuelin Reg 3 32.0 7.9 Prior to Refuelin og 22 31.3 7.9 Prior-to Refuelin og 18 32.2 52.6 Prior to Refuelin og 29 34.3 56.1 Prior to Refuelin og 3 34.6 56.8 Prior to Refueling Reg 33 37.5 57.0 l

Final (AfterRefueling) 35.0 56.0

2) Strip Chart Data Core Confinuration SUC1 (CPS)

SUC2 (CPS)

After source p' aced in Reg 22. Two Layers of Not Available 1600 Reflector Blocks added but l

no Plenum Blocks or CRD l

in Reg 22 L

After Source placed in Reg 22. Top Reflector and Not Available 320 t.

Plenum put back, no CRD in i

Reg 22 After Source placed in Reg 22. Top Reflector, Not Available 90 Plenum and CRD put back in Reg 22 t

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EE-DEC-0022 4

I REV B Page 20 of 39 j

Yable IV.2 Detector Reeponse Comparisons for 1/2 Core DIF3D vs Experimental Data

1) Surveillance Data: All Regions Refueled After Source in Region 22

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i Neutrons /

f Neutrons /

SUC1 SUC1 Count SUC2 SUC2 Count Measured CPS 35 56 DIF3D Total Neutrons 6.68E+2 19.1 1.28E+3 22.9

2) Strip Chart Data: Source in Region 22. No Plenum in Upper Reflector CR0 Removed from Core i Neutrons /

f Neutrons /

SUC1 SUC1 Count SUC2 SUC2 Count Measured CPS Unavailable 1600 DIF3D Total Neutrons 6.82E+2 Unavailable 3.39E+4 21.2

3) Strip Chart Data: Source in Region 22, Plenum Blocks Put Back in Core, CRD Removed from Core P

1. Neutrons /

f Neutrons /

SUC1 SUCI Count SUC2 SUC2 Count Measured CPS Unavailable 320 DIF3D Total Neu'trons 6.82E+2 Unavailable 2.50E+3 7.8

4) Strip Chart Data: Source in Region 22, Plenum and CRD Back in Core i Neutrons /

f Neutrons /

SUC1 SUCI Count SUC2 SUC2 Count Measured CPS Unavailable 90 DIF3D Total Neutrons 6.68E+2 Unavailable 1.28E+3 14.2

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i i-BASED UPON THE BENCHMARKED MODEL A SENSITIVITY STUDY WAS PERFORMED TO k

-ASSESS CORE PERFORMANCE DURING DEFUELING.

1 THE FOLLOWING PARAMETERS WERE ANALYZED o

NEUTRON SOURCE LOCATION o

NEUTRON SOURCE STRENGTH o

EFFECT OF COCKED CONTROL RODS i

o USE OF REPLACEMENT BORONATED REFLECTOR BLOCKS o

REPLACEMENT OF UPPER PLENUM BORONATED BLOCKS o

USE OF PURE GRAPHITE BLOCKS AS SOURCE CONTAINERS 1

o AGE OF THE CORE IN EFPD l

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EE-DEC-0022 l

REV B Page 35 of 39 l

Table V.8 Near End-of-Defueling Performance for i

the 232 EFPD FSV Core e

Assumptions:

4.0E+9 Sources in Reg 3. Col 3 and Reg 6. Col 6 Rings 3+4 Defueled Minus Regions 10 and 16 250 EFPD Burnup Use of Pure Graphite Source Blocks Replace Boronated Plenum Elements in Regions 3, 10, 6, and 16

$UC 1 SUC 2 Core Predicted Predicted Configuration

1. Neutrons Count

1. Neutrons-Count All Rods In For Fueled 2.78E+2 13 3.17E+2 14 Regions All Rods Dut for 1.52E+3 69 3.16E+3 144 All Regions

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L IN CONCLUSION L

o THE MODEL AND METHODS PRESENTED ARE TECHNICALLY APPROPRIATE FOR PREDICTING FSV START-UP DETECTOR RESPONSE DURING DEFUELING.

o THE ANALYTICAL METHODS DESCRIBED FORM THE BASIS FOR PREDICTING START-UP DETECTOR COUNT RATE DURING DEFUELING.

o THE METHODS AS DESCRIBED ARE CONSERVATIVE SUCH P

THAT THE AS-MEASURED COUNT RATE EXPECTED DURING ACTUAL DEFUELING WILL BE EQUAL TO OR GREATER 1

1.

THAN THE PREDICTIONS MADE IN ADVANCE.

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THE-MODEL AND METHODS ARE FLEXIBLE ENOUGH TO BE USED FOR OTHER POSSIBLE DEFUELING SCENERIOS SHOULD THE NEED ARISE IN THE FUTURE.

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i THE PROBLEM MAINTAIN ADEQUATE START-UP CHANNEL COUNT RATE c

DURING FSV DEFUELING UNTIL THERE IS NO' LONGER A CREDIBLE MEANS OF ACHIEVING CORE CRITICALITY l

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THE SOLUTION 1

l APPLY NEUTRON SOURCE ANALYSIS TO DESIGN IN-CORE CF-252 NEUTRON SOURCES SUCH THAT THE 4.3 COUNTS l'

/SECOND TECHNICAL SPECIFICATION REOUIREMENTS ARE l

l MET THROUGHOUT THE DEFUELING PERIOD.

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ALTERNATIVES FOR DEFUELING CASE O l

e IDAHO ISSUES RESOLVED AND DEFUELING TAKES PLACE AS PLANNED o BEGIN DEFUELING NOV 27, 1989 UTILIZING FUEL STORAGE WELLS (FSW'S) AS A LAG FACILITY CONCURRENT WITH SHIPMENTS TO IDAHO e PRIMARY ISSUE - UPGRADE OF GSF/ READINESS TO RECEIVE FUEL T

e FINAL SEGMENT STILL AN ISSUE s ACCEPTANCE OF FINAL SEGMENT AT IDAHO e BUILD A ONE SEGMENT ISFSI (UTILIZE FSW'S FOR INTERIM STORAGE) e USE FSW'S FOR LONG TERM STORAGE e DEFUELING SEQUENCE /SAR a UTILIZE SEQUENCE AS DEFINED IN EXISTING SAR e UTILIZE EXISTING SAR BUT WITH A MODIFIED SEQUENCE (UNDER 10CFR50.59)

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L ALTERNATIVES FOR DEFUELING CASE I e BASIC IDAHO ISSUES RESOLVED BUT CANi40T BEGIN SHIPPING TO IDAHO UNTIL MID TO LATE JANUARY 1990 e BEGIN DEFUELING N0/ 27, 1989 UTILIZING FSW'S AS A LAG FACILITY e PRIMARY ISSUE - UPGRADE / READINESS OF GSF TO RECEIVE FUEL ON OR A3OUT JAN 17, 1990 e FINAL SEGMENT STILL AN ISSUE e ACCEPTANCE OF FINAL SEGMENT IN IDAHO e BUILDAONESEGMENTISFSI(UTILIZEFSW'S FOR INTERIM STORAGE)

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m USE FSW'S FOR LONG TERM STORAGE

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e UTILIZE SEQUENCE AS DEFINED IN EXISTING SAR e UTILIZE EXISTING SAR BUT WITH A MODIFIED SEQUENCE (UNDER 10CFR50.59) l 1:

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ALTERNATIVES FOR DEFUELING CASE 2 L

o IDAHO ISSUES CAN BE RESOLVED BUT RESOLUTION IS NOT TIMELY e DEFUELING-a BEGIN DEFUELING AS SCHEDULED UTILIZING THE FSW'S AS A LAG FACILITY a DELAY DEFUELING UNTIL ISSUES ARE RESOLVED e FINAL SEGMENT STILL AN ISSUE e ACCEPTANCE OF FINAL SEGMENT IN IDAHO a BUILD A ONE SEGMENT ISFSI (UTILIZE FSW'S FOR INTERIM STORAGE) 1 e UTILIZE FSW'S FOR LONG TERM STORAGE e DEFUELING SEQUENCE /SAR l

L a BEGIN DEFUELING WITH EXISTING SAR WITH l

MODIFIED SEQUENCE (UNDER 10CFR50.59) m UTILIZE THE EXISTING SAR FOR DELAYED DEFUELING e PREPARE A NEW SAR TO ACHIEVE LONG TERM, I

l LESS REACTIVE CORE CONDITIONS e SOME COMBINATION OF THE AB0VE l

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.l ALTERNATIVES FOR DEFUELING CASE 3 e IDAHO ISSUES CANNOT BE RESOLVED e DEFUELING a BEGIN DEFUELING AS SCHEDULED UTILIZING THE FSW'S AS A LAG FACILITY a DELAY DEFUELING e PSC CONSTRUCT A SIX (6) SEGMENT ISFSI e DEFUELING SEQUENCE /SAR e BEGIN DEFUELING WITH EXISTING SAR WITH MODIFIED SEQUENCE (UNDER 10CFR50.59) e UTILIZE THE EXISTING SAR FOR DELAYED DEFUELING e PREPARE A NEW SAR TO ACHIEVE LONG TERM, LESS REACTIVE REACTOR CORE CONDITIONS a SOME COMBINATION OF AB0VE 7

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REACTOR CORE CONDITIONS FOR VARIOUS ALTERNATIVES e CASE 0/ CASE 1 s SHRINKING CORE CONCEPT AS DESCRIBED IN SAR a CONSIDER RING 4 SEQUENCE CHANGE TO DEFUEL MOST REACTIVE REGIONS FIRST e CASE 2/ CASE 3 a PRIMARY CONSIDERATION TO PLACE REACTOR IN SAFEST POSSIBLE CONDITION CONSISTENT WITH DELAYS a CONSIDER CONTROLLING THE MOST REACTIVE CORE REGIONS (SEGMENTS 8 & 9)

DISABLING CONTROL RODS

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DEFUEL MOST REACTIVE REGIONS 1

UTILIZE RESERVE SHUTDOWN MATERIAL 1

DEFUEL SOME OF THE MOST REACTIVE REGIONS L

IN COMBINATION WITH RESERVE SHUTDOWN MATERIAL VARIOUS COMBINATIONS OF THE AB0VE l

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CONSIDERATIONS FOR DELAYED DEFUELING t'

e SAFETY e PRIMARY CONSIDERATION / RESPONSIBILITY IS TO PLACE REACTOR CORE IN THE SAFEST POSSIBLE o

CONDITION CONSISTENT WITH THE' DELAY TIME e OTHER POSSIBLE BENEFITS a ELIMINATE POSSIBILITY OF INADVERTENT REACTIVITY EXCURSIONS e REDUCE DECAY HEAT e REDUCE SUPPORTING SYSTEMS AND ASSOCIATED SURVEILLANCE AND MAINTENANCE a REDUCTION OF VITAL AREAS COULD RESULT IN SECURITY REDUCTION e RADIOLOGICAL EMERGENCY RESPONSE PLAN COULD BE DOWNGRADED s DEFUELING OPERATIONS AFTER DELAY PERI O COULD BE SIMPLIFIED e POSSIBLE TO SUPPORT POSSESSION ONLY LICENSE l

e BASES FOR EXEMPTIONS TO APPENDIX R AND EQ (10CFR50.49) e BASES FOR LIABILITY INSURANCE REDUCTIONS l

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