ML16162A314

Summary of 820127 Meeting W/Util & Westinghouse in Bethesda, MD Re Demonstration Spent Fuel Program Consolidation Program.Agenda,Attendee List & Viewgraphs Encl

ML16162A314
Person / Time
Site:	Oconee
Issue date:	02/02/1982
From:	Wagner P Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8202110275
Download: ML16162A314 (36)

Text

8 RE~UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Ol FEUARY 0~ 108 Dockets Nos. 50-269/270/287 LICENSEE:

DUKE POWER COMPANY (DPC)

FACILITY: OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3

SUMMARY

OF MEETING HELD ON.JANUARY 27, 1982 WITH REPRESENTATIVES OF DPC AND WESTINGHOUSE CONCERNING SPENT FUEL CONSOLIDATION A meeting was held on January 27, 1982 in our Bethesda, MD offices with DPC and Westinghouse representatives to discuss the demonstration-spent fuel consolidation program which Duke has contracted with Westinghouse to perform on Oconee spent fuel. A list of attendees to this meeting is provided in Enclosure 1.

DPC representatives introduced the subject and explained that long term spent fuel management requires evaluation of a number of options; among these are cost, logistics, storage capacity and 'onstruction require ments.

In light of the unavailability of Federal reprocessing, reposi tories or storage facilities, the only viable option at this time is some type of onsite storage. The Oconee 1 and 2 common spent fuel pool has been reracked twice and the Oconee 3 pool was reracked in 1975 and plans are underway to initiate an additional reracking to increase fuel storage. However, in order to ensure adequate spent fuel storage capacity in the long term, Duke has entered into a contract with Westing house Electric Corporation to donduct a spent fuel rod consolidation program on Oconee fuel.

It should be-noted that the present contract is only for a demonstration program to ensure the capability of the Westinghouse system to perform spent fuel consolidation safely and ef ficiently with no further commercial commitments.

A copy of the slide presentation from both DPC and Westinghouse is pro vided in Enclosure 2. Some of the Westinghouse presentation is con sidered to be proprietary and is not included.

The proposed Westinghouse demonstration program involves the use of four Oconee spent fuel assemblies which were removed from service in 1977.

All calculations and assumptions are based on the use of these assemblies which have been cooled in the pool for approximately five years.

No long-term analyses were presented for possible future use of a similar consolidation program for additional fuel assemblies since a decision on total or additional consolidation has not been made. DPC recognizes that a pro'gram.of-total.consolidation would require NRC approval and has agreed to submit the necessary documentation for review if such a decision is made.

8202110275 820202 PDR ADOCK 05000269 P

PDR

DPC The consolidation system would use special equipment which would be inserted in the spent fuel cask. loading area of the pool.

This equipment would remove all the fuel rods from one assembly, after removal of the top nozzle, and insert the rods from the assembly into one half of a center-divided spent fuel storage canister; fuel rods from a second fuel assembly would be loaded in the other half of the canister thereby placing the fuel rods from two assemblies into one canister which in turn would be inserted into one spent fuel storage pool rack location. This'operation would, therefore, double the spent fuel storage capacity in the pool. However, only two such canisters are involved in this demonstration program. The remaining components of the fuel-assemblies (inlet and outlet nozzles,. guide-tubes and spacer grids) would then be compacted by other special equipment in the ca'k loading area and placed in shipping containers - two assemblies per container.

Design and evaluation of the demonstration equipment has been completed and fabrication is in progress.

Present schedules provide for a demonstration on a dummy fuel assembly (no fuel) in the Spring of 1982, followed by con solidation of four Oconee s-pent fuel assemblies at the Oconee Station in late Summer 1982. Duke maintains that the demonstration program utilitizing four spent fuel assemblies which were removed from service in 1977 can be per formed under the provisions of 10 CFR 50.59. The justification that an unreviewed safety question is not involved in this program is contained in the Enclosure 2 sections on handling accidents, structural, seismic, criticalit.y, and thermal-hydraulic analyses and radiological considerations.

I questioned the determination that an unreviewed safety question was not involved and requested a submittal from DPC.to formally inform the NRC that they plan to proceed with the demonstration program. DPC agreed to provide a submittal, but requested a timely response from the NRC if we disagree with their determination.

Therefore, I request any comments on this proposed pro gram from any interested organization within 30 days of the date of this meeting summary so that upon receipt of the DPC submittal I can provide a timely response.

orgnlsigned W Philip C. Wagner, Project Manager Operating Reactors Branch #4 Division of Licensing.

Enclosures:

1. List of Attendees

2.

Agenda cc w/enclosures:

See next page ORB#4:

OFFICE)

SURNAME)....

ne..cb.

D A T E NRC FORM 318 (10-80) NRCM 0240 OFFICAL RECORDCOPY USGPO: 1981-335-960

ORB#4:DL MEETING

SUMMARY

DISTRIBUTION Licensee:

Duke Power Company

• Copies also sent to those people on service (cc) list for subject plant(s).

D R L PDR ORB#4 Rdg TNovak JStolz Project Manager -PWagner Licensing Assistant -RIngram OELD Heltemes, AEOD IE-3 SShowe (PWR)

Meeting Summary File-ORB#4

RFraley, ACRS-10 Program Support Branch ORAB, Rm.

542 BGrimes, DEP SSchwartz, DEP SRamos, EPDB FPagano, EPLB JBoegli WGammill JHayes CSchulten WBrooks TChan ORothberg l MWohl WPasedag LO'Reilly S3lock CBerlinger LIST OF ATTENDEES JANUARY 27, 1982 DPC MEETING Westinghouse NRC Greta Heubness Jacques S. Boegli Ed Shields W. P. Gammill Elmer.Bassler P. C. Wagner Harry Flanders John J. Hayes Barry Cooney Carl S. Schulten Walter L. Brooks Terence L. Chan Oden Rothberg Duke Milliard L. Wohl Walter Pasedag Wayne Morgan Lynne A. O'Reilly H. T. Snead Seymour Block Tom Curtis Carl H. Berlinger*

Bob Rasmussen Rober Gill

• part time AGENDA NRC FUEL CONSOLIDATION MEE=TIG

1.

Introduction

2.

Process Overview a)

General Process Description b)

Administration

.c)

Tech Spec. Limits d)

Waste

3.

Technical Review a)

Nuclear Considerations b)

Thermal Hydraulic Considerations c)

Structural/Seismic Considerations d)

Radiological Considerations e)

Accident Considerations

-f)

Handling Equipmnt

4.

Schedule

SPENT FUEL STORA(E OBJECTIVES

1.

PROVIDE ADEQUATE STORAGE CAPACITY

• 2. AVOID OR DELAY CAPITAL INTENSIVE STORAGE OPTIONS

• BASED UPON THE OPINION THAT ALL STORAGE OPT-IONS ARE ENVIRONMENTALLY EQUIVALENT

ADEQUATE CAPACITY

1. THAT CAPACITY OFFERING THE AVAILABLE LEAD TIME TO BE ABLE TO DESIGN, LICENSE, AND CONSTRUCT NEW CAPITAL FACILITIES FOR STORAGE

2. OR THAT POSITION WHERE IS IN HAND THE KNOWLEDGE THAT THIS LEAD TIME CAN BE MADE AVAILABLE THRU A COMBINATION OF STORAGE OPTIONS

3.

OR THAT CAPACITY ASSURED BY THE AVAILABILITY OF REPROCESSING, FEDERAL REPOSITORIES, OR FEDERAL STORAGE FACILITIES 14, OR THE ACTUAL AVAILABILITY OF LICENSED REACTOR DISCHARGE CAPACITY.

CASE 5 LEAD TIME COMPARISON OF SPENT FUEL STORAGE TECHNOLOGIES DECISION VAULT DECISION RERACK DECISION TRANSSHIPMENT DECISION DRYWELL DECISION CASKS DECISION ROD CONSOLIDATION DECISION POOLS.

FCR FCR FOR

DUKE/WESTINGHOUSE CONTRACT

1.

FUEL ROD CONSOLIDATION DEMONSTRATION

2. OCONEE FUEL

3.

No COMMERCIAL COMMITMENTS BEYOND THE DEMONSTRATION

ADVANTAGES OF SPENT FUEL CONSOLIDATION

1.

POTENTIAL FOR SIGNIFICANT STORAGE INCREASE 2,

POTENTIAL FOR.EASING LONG TERM LOGISTICAL PROBLEMS OF FUEL CYCLE

3.

POTENTIAL FOR COST REDUCTION IS GENERIC IN NATURE

4.

DEVELOPMENT APPEARS TO BE RELATED TO APPLICATION OF EXISTING TECHNOLOGY.

DISADVANTAGE OF CONSOLIDATION

1.

BURDEN OF THE WORK IS PLACED ON THE STATION

I

• D CASE 5 LEAD TIME COMPARISON OF SPENT FUEL STORAGE TECHNOLOGIES DECISION VAULT DECISION,RERACK DECISION TRANSSHIPMENT DECISION DRYWELL DECISION CAKS DECISION ROD CONSOLIDATION DECISION POOLS 040 0o 00 T-T N

Cli c'f e

FCR FCR FCR

PROGRAM ADMINISTRATION I. PROGRAM WILL BE CONDUCTED UNDER PROCEDURES AND POLICIES OUTLINED IN THE STEAM DEPARTMENT "ADMINISTRATIVE POLICY MANUAL." WITH REGARD TO.

(A) HEALTH AND SAFETY OF THE PUBLIC (B) COMPLIANCE WITH LICENSE AND REGULATORY REQUIREMENTS (C) MINIMIZING THE IMPACT ON NORMAL STATION OPERATIONS.

(D) QUALITY ASSURANCE

REQUIREMENTS FOR STATION MODIFICATION

1. DESIGN INFORMATION NECESSARY FOR IMPLEMENTATION

2. PERFORMED SAFETY EVALUATION OF THE PROCESS 3, PERFORMED.FUEL ACCOUNTABILITY EVALUATION 4* REVIEW BY OCONEE STATION SAFETY REVIEW BOARD

5. REVIEW BY QUALITY ASSURANCE DEPARTMENT

6. REVIEW BY NUCLEAR SAFETY REVIEW BOARD

7. REGULATORY REQUIREMENTS,.IF ANY, ARE SATISFIED

8. PROCEDURES FOR IMPLEMENTATION ARE AVAILABLE

SCHEDULE FOR FUEL CONSOLIDATION PROJECT APRIL/MAY 1982 COLD DEMONSTRATION ON DuMMY FUEL JUNE/SEPTEMBER 1982 HOT DEMONSTRATION OCONEE UNIT 1/2 SPENT FUEL POOL AUGUST 1982 -

PREPARATION OF UNIT 3 POOL FEBRUARY 1984 AND INSTALLATION OF RACKS

OCONEE UNITS 1 and 2 SPENT FUEC POOL East Ca'sk-Pif Fue Consolidation Systems Stand Skeleton Compaction System ask Platfom

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SPENT FUEL CONSOLIDATION STRUCTURAL / SEISMIC CONSIDERATIONS 8

Structural Fuel Handling Tools Designed in Accordance with Westinghouse Specifications for Spent Fuel Handling Tools System Frame and Components Designed in Accordance with ASME Code Section III Article XVII-2200 0

Seismic,

- Stored Fuel Shall be Removed from the Racks a Minimum Distance of Twelve Storage Rows

-eFuel Assemblies and Consolidated Canisters will Not be Handled Over Stored Fuel Assemblies An Area.of at Least Two Fuel Assemblies in All Directions Shall be Cleared for Insertion of Storage Canister in Fuel Storage Racks

.- The Fuel Handling Tools will be Designed to Withstand an Axial Load of Seven Times the Static Load Without Fa iLre -

ACCOUNTABILITY CRITERIA ALL SPECIAL MATERIALS SHALL BE IDENTIFIABLE BY LOCATION COMPLIANCE. -

THE END CLOSURES FOR CONSOLIDATED ROD STORAGE CANISTERS WILL BE PROVIDED WITH CLEARLY VISIBLE IDENTIFICATION.

THE COMPLEMENT OF RODS FROM EACH CONSOLIDATED FUEL ASSEMBLY WILL BE CORRELATED TO THE SPECIFIC STORAGE CANISTER HALF IN WHICH THEY WILL RESIDE.

SPECIAL CANISTERS WITH CLEARLY VISIBLE IDENTIFICATION WILL BE PROVIDED TO STORE STRAY RODS (NOT EXPECTED).

STRAY.RODS WILL BE ADMINISTRATIVELY TRACEABLE FROM FUEL ASSEMBLY LOCATION TO SPECIAL CANISTER.

FASTENERS FOR END CLOSURES ON STORAGE CANISTERS WILL BE PROVIDED WITH LOCKUCPS TO.PREVENT INADVERTANT REMOVAL.

FUEL CANISTER DROP ANALYSIS CRITERIA

1.

FUEL CRITICALITY WILL NOT OCCUR

2. PERFORATION OF POOL LINER WILL NOT OCCUR ASSIPJPTIONS

1. FUEL CANISTER FREE FALLS BETWEEN RACK AND POOL WALL

2. DROP HEIGHT = 200 IN.

3.

CANISTER WEIGHT = 3000 LBS.

4. RESISTANCE OF WATER IS NEGLECTED RESULTS 1.KEFF < 0.95 WITH CONSOLIDATED FUEL

2. LENGTH OF CANISTER BELOW FUEL RODS ABSORBS ENERGY AT IMPACT

3.

PERFORATION OF POOL LINER DOES NOT OCCUR

STRUCTURAL & SEISMIC ANALYSIS

• LOCAL CELL ASSEMBLY STRESSES s NORMAL OPERATION a SEISMIC EVENT o POOL FLOOR LOADS a CANISTER SLID.ING/IMPACTING IN CELL

Upper Spacer Plate Upper Grid Side Plate Cell Poison Enclosure Material Poison-Material Lower Spacer Plate Lower Grid-Box Beam Lower Grid Side Plate FiSure 1 Cell Acoscbly

TABLE 3 CELL ASSEMBLY DESIGN STRESSES AND MINIMUM MARGIN OF SAFETY Normal & Upset Condition Faulted Condition sDesign

tress, psi M.S.

esign ress, psi

.S._

Component Category rspiMS.

Std Fuel :onsol Fuel Std Fuel Consol Fuel Std Fuel Consol Fuel Std Fuel Consol Fuel (1)

Enclosure Bending 4936 8933 2.34 0.85 8278 14883 2.99 1.22 Compression 211 338

> 10

> 10 317 507

> 10

>1 Bend.+Comp.

(2)

(2) 2.13 0.74 (2)

(2) 2.76 1.10 Cell Corner Weld Shear 3450 5744 5.95 3.17 5210 8616 8.21 4.57 Wrapper Cell Weld Shear 7440 7909 0.48 0.39 11300 15818 0.95 0.39 Cell-Spacer Plate Weld Shear 10650 15666 1.25 0.53 16080 23499 1.98 1.04 Spacer Plate Grid Weld Shear 14600 23077 0.64 0.03 22050 30573 1.17 0.36 (1) Local buckling of enclosure and wrapper is allowed and is assumed in calculating enclosure stresses.

(2) f a f

f

+

bx b

1 F

+

< 1.0 a

Fb.

by-

POOL FLOOR LOADS WEIGHT WITH WEIGHT WITH STD.

FA'S ARRAY STD. FA'S (LBS)

+ 2 CONSOL, CANS (LBS)

% INCREASE 8 x 11 155100 157200 1.4 8 x 12 169200 171300 1.2 NOTE:

IF, FOR EACH CONSOLIDATED CANISTER STORED IN A RACK THERE IS ONE CELL LEFT EMPTY, THE FLOOR LOADS FOR BOTH NORMAL AND SEISMIC EVENTS ARE LESS THAN THOSE EVALUATED FOR LICENSING,

CANISTER SLIDING OR IMPACTING INSIDE STORAGE CELL NOT A CONCERN DUE TO CANISTER DESIGN

CRITICALITY ANALYSIS ACCEPTANCE CRITERIA KEFF 0.95 INCLUDING UNCERTAINTIES

ASSUMPTIONS FRESH FUEL INFINITE WATER REFLECTOR PURE WATER AT 1.0 GM/CM 3 DOUBLE CONTINGENCY PRINCIPLE GEOMETRY CONTROL WHERE POSSIBLE

GEOMETRY CONTROL NOT FEASIBLE WHEN WITHDRAWING FUEL RODS WORST NoN-ACcIDENT CONDITION IS UNIFORM INCREASE IN PITCH FOR WORST PITCH, K.EFF < 0,95

KEFF vs. ROD PITCH FOR A B&W 15x15 ASSEMBLY AT 275 w/o ENRICHMENT, NO BURNUP 0.95 0.90 A

0.85 U.J 0 80 0.55 0.60 0.65 0.70 0.75 0.80 0.85 0.90 ROD PITCH (INCHES)

THERMAL HYDRAULIC EVALUATION CRITERIA TO BE CONSIDERED DESIGN HEAT LOAD OF SPENT FUEL POOL COOLING SHALL NOT BE EXCEEDED UNDER CONDITION I OCCURRENCES NO BOILING (LOCAL OR BULK) IS PERMITTED IN THE POOL UNDER NORMAL CONDITIONS

THERMAL HYDRAULIC EVALUATION ANALYTICAL BASIS & ASSUMPTIONS

• WATER LEVEL 24 FT. ABOVE TOP OF FUEL o AVERAGE DECAY HEAT RELEASE OF 961 WATTS PER ASSEMBLY o MAXIMUM INLET TEMPERATURE OF 150OF

@ CANISTERS PLACED IN INNERMOST RACK POSITIONS o PEAK TO AVERAGE ROD HEAT OUTPUT RATIO OF 1.6

• CRUD THICKNESS LAYER OF 5 MILS

• DISTANCE FROM BOTTOM OF RACK TO FLOOR OF 10 IN.

• RACK INLET AREA-OF 34,0 SQ, IN.

o CANISTER INLET AREA OF 20.5 SQ. IN.

• NET FLOW AREA THROUGH CANISTER OF 9.8 SQ. IN.

THERMAL HYDRAULIC EVALUATION RESULTS OF ANALYSIS (1)STANDARD (2)CONSOLIDATED FUEL FUEL

• WATER TEMPERATURE AT RACK 150 150 INLETJ OF

• MAXIMUM CELL WATER TEMP OF 183.6 178.6

• MAX. CLAD SURFACE TEMP 220 179 AVG. ROD.,

F

• MAX. CLAD SURFACE TEMP.,

241 179 PEAK ROD, OF

• CLAD TEMP INCREASE WITH 90% BLOCKAGE OF RACK INLET IS e1F e

GEOMETRY IS SUCH THAT FLOW WILL REDISTRIBUTE AT CANISTER INLET IN EVENT OF END PLATE FLOW HOLE BLOCKAGE

• INCREASE IN POOL HEAT LOAD WITH FULL STORAGE RACKS WOULD BE <0.2%

(1 )T&H ANALYSIS OCONEE 1 & 2., MARCH 1980 (2)C COMPUTER RUN AFEAHYM WITH RACKS CODE

THERMAL HYDRAULIC EVALUATION CONCLUSIONS LOCAL COOLANT OUTLET AND CLAD SURFACE TEMPERATURES IN STORAGE CANISTERS WILL BE BELOW COMPARABLE VALUES FOR STANDARD FUEL PREVIOUSLY ANALYZED, TEMPERATURE INCREASES WITHIN CANISTERS WITH MAXIMUM EXPECTED FLOW BLOCKAGES WOULD BE NEGLIGIBLE.

CONSOLIDATED ASSEMBLIES RESULT IN NO INCREASE IN EXISTING POOL HEAT LOADS, WITH FULL STORAGE RACKS HEAT LOAD INCREASE WOULD BE NEGLIGIBLE,

RADIOLOGICAL CONSIDERATIONS FSAR ROD FAILURE ANALYSIS/ASSUMPTIONS EVALUATION OF ASSUMPTIONS CONSOLIDATED FUEL FAILURE ANALYSIS/ASSUMPTIONS COMPARISON WITH CASK DROP ANALYSIS COMPARISON WITH 10CFR 100 LIMITS

RADIOLOGICAL ANALYSIS COMPARISON ANALYSIS WHOLE BODY DOSE THYROID DOSE 56 RODS 0.027 REM 0,43 REM 416 RODS 0.021 REM 0.00 REM 416+42 RODS 0.038 REM 0.33 REM 21 ASSEMBLIES 2.320 REM 36.7 REM 46 ASSEMBLIES 4.950 REM 78.7 REM 10CFR 100 LIMITS 25 REM 300 REM

RADIOLOGICAL ANALYSIS ASSUMPTIONS

1. 416 CONSOLIDATED RODS DAMAGED

2. GAP ACTIVITY BASED ON FSAR VALUES WITH 2.YEARS DECAY 3, 42 RODS DAMAGED FROM OTHER STORED ASSEMBLIES (20%)

4.

GAP ACTIVITY OF 42 RODS BASED ON FSAR VALUES 5, */o AND IODINE RETENTION FACTORS SAME AS FSAR

WASTE DISPOSAL CONSIDERATIONS, 4 ASSEMBLY SKELETONS PLUS FILTERS AND MISC, LESS THAN 1000 CURIES PER SKELETON WESTINGHOUSE COMPACTION EQUIPMENT CHEM-NUCLEAR 1-13G SHIPPING CASK TWO SHIPMENTS, TWO SKELETONS EACH BROKEN RODS SEALED AND STORED IN SPENT FUEL POOL