ML20236C992

Summary of 890307 Meeting W/Util & Ga Technologies Re Plant Coastdown & Defueling.Attendance List,Agenda & Viewgraphs Encl

ML20236C992
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/13/1989
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Calvo J Office of Nuclear Reactor Regulation
References
TAC-69028, NUDOCS 8903220407
Download: ML20236C992 (76)

Text

O AfL K40g - .*

o UNITED STATES g ,

, g NUCLEAR REGULATORY COMMISSION ,

5 j WASHING TON, D. C. 20866

]

' k % ...,/ March 13, 1989 j J

Docket No. 50-267  !

l l

MEMORANDUM FOR: Jose A. Calvo, Director- )

Project Directorate - IV .

l Division of Reactor Projects - III, IV, Y and Special Projects ,

FROM: Kenneth L. Heitner, Project Manager Project Directorate - IV '

l Division of Reactor Projects - III, IV, Y and Special Projects

SUMMARY

OF MARCH 7, 1989 MEETING WITH PUBLIC SERVICE l COMPANY OF COLORADO (PSC) AND GA TECHNOLOGIES (GAT)-  !

- i ON FORT ST. VRAIN (FSV) C0ASTDOWN AND DEFUELING (TACNO.69028) i This meeting was held with PSC to discuss developing the licensee's 1 preliminary plans for Fort St. Vrain's projected end of nuclear operations.

The meeting covered reactor end-of-cycle coastdown and defueling. GAT

, representatives participated in the meeting because the detailed studies supporting this work are done by GAT. The list of attendees are in Attachment 1.

Material provided at the meeting by PSC is Attachment 2.

^

Defuel_1,n_g PSC examined a number of options to defuel the FSV reactor. This included i defueling the reactor by layer, with the use of temporary absorbers in the ,

core. However, the optinum choice selected was defueling by region. The i defueling would be accomplished in a successive' radial pattern, working from the outside of the reactor to the center of the core. The defueling would be .

adequately monitored by the plant's redundant startup detector channels which - l are in the head of the reactor.

As each region of the reactor is defueled, dumnty blocks would be used to '

maintain the core's structural integrity. The dunuty blocks would be loaded into the reactor as would fuel in a normal refueling operation. Upper reflector blocks would also be replaced to return the. core to its normal configuration. PSC was evaluating the need to replace the Region Restraint .

Devices, since these are used only in power operation. The dunsty blocks would be fabricated from HLM graphite. This material is already in use in the reactor core as permanent reflector blocks. Its structural properties were considered adequate to maintain core integrity. The dumnly blocks would be fabricated to be geometrically identical to the fuel blocks they replace. The dumrqy blocks would contain boron material fabricated in a 12 " pin" configuration. ,

The fuel blocks would also contain helium flow channels, but not as great a '

number as a normal fuel block. This would tend to force coolant flow;through-the remaining regions with fuel.

8903220407 DR 890313 ADOCK 05000267 PDC ,

m ,- - -

S , l

' Jose A. Calvo PSC and GAT have used existing codes to evaluate this defueling plant. These codes show that the core would remain subcritical by at least the normal shutdown margin (.01 delta k) throughout the defueling process. However, as defueling proceeds, the normal reactor neutrons sources would be removed.

11ence, it is the licensee's intent to place a new source'in Region 1 (the center of the core). This new source would assure an adequate count rate to  ;

, the startup channels, during defueling to assure the core condition is properly I

, monitored. The startup channel would scram the remaining " cocked" rod should 5

the count rate rise to the trip set point of 10 counts per second (cps).

J Surveillance testing at each stage o'f the defueling process would assure that i the shutdown margin is maintained, even as additional control rodscare withdrawn  !

to allow successive regions to be defueled. Once a region is defueled, the boron in the dummy blocks is sufficient to maintain that region equivalent to one with its control rod inserted. The control rod drive mechanisms for these regions would nominally be replaced, but only for storage purposes. <

l Toward the end of the defueling process, two effects occur. First, the remaining core becomes subcritical, even with ali rods removed. At this )

point it may also be difficult to monitor the core since the rate at which neutrons reach the detector is.too low to assure reliable performance'. PSC  !

and GAT are continuing to evaluate these problems and how they will be i addressed. J Coastdown The licensee had conducted en extensive evaluation of a potential coastdown after the current fuel was utilized for their normal 300 equivalent fuel power days. The calculations showed that a sustained coastdown period was possible, with no adverse affects on safety. Normal plant control systems would I acccamodate these pcher level changes.

No unreviewed safety questions or Technical Specification' changes were I involved. However, PSC would provide a summary of this unusual operation for information to the staff.

Technical. Specification Changes Technical Specification (TS) changes would be required for the FSV defueling.

• Defueling safety analyses would also have to be reviewed by the staff. PSC proposed that the defueling TS would be designed to become effective approximately 100 days after the final reactor shutdown. At this point, PSC expects that the core would be sufficiently cool so as to require a simpler set of shutdown and defueling TS. These TSs would be btsed on model TSs already extensively discussed in the TS Upgrade Program. PSC proposed to-submit these TSs by May 31, 1989.

I

e g Jose A. Calvo Other Items PSC also noted that plans for FSV decommissioning were proceeding. PSC still i intended to file a preliminary plan with the Commission by March _31,1989. {

Various schemes for decommissioning were still being evaluated. PSC also i stated that they would use the fuel storage wells for interim spent fuel  ;

storage until the long term disposal of the spent fuel could be suitably ]

arranged (and licensed as required). ]

1 The staff inquired about PSC's intentions to utilize the FAN 3D code, currently '{

being reviewed by. the staff under TAC No. 69028. PSC stated that they would l not require the use of this code for continued reactor operation or defueling i

analyses. (The code may be used for fuel accountability). The staff considers TAC No. 69028 and its associated review to be closed.

Staff Comments The staff noted the importance of reliable and safe operation of the fuel _ .

handling machine in order to complete the defueling operation. The staff noted  ;

that in its January letter, PSC had recommended that further Inservice 1 Inspection and Testing20,(1989ISIT) program development be cancelled. The staff i noted that the ISIT for the fuel handling machine was of continued importance, 1 and that the PSC should continue its development. Furtherncre, the staff l requested the PSC consider submitting their overall ISIT program for the fuel

! handling machine for staff review. The staff noted its intent to formally request this item from PSC.

Kenneth L. Heitner, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects-Attachments:

As stated DISTRIBUTION Docket File NRC PDR Local PDR J. Snfezek' PD4 Reading J. Calvo K. Heitner OGC-Rockville E. Jordan B. Grimes F. B. Litton L. Kopp T. F. Westerman, RIV P. W. Michaud, RIV ACRS(10) T. !!artin (Region IV)

PD4 Plant File PD4/PM PD4/D /

KHeitner:bj JCalvo 03/ P/89 03/d/89

-_ _. _ _____m___.-___..____________.____.____-u-- _

. J s

l Jose A. Calvo _0ther Items PSC also noted that plans for FSV decommissioning were proceeding. PSC still

. intended to file a preliminary plan with the Comission by March 31, 1989.

Various schemes for decommissioning were still being evaluated. PSC also stated that they would use the fuel storage wells for interim spent fuel storage until the long term disposal of the spent fuel could be suitably arranged (and licensed as required). l The staff inquired about PSC's intentions to utilize the FAN 3D code, currently l I

being reviewed by the staff under TAC No. 69028. PSC stated that they would i not require.the use of this code for continued reactor operation or defueling analyses. (The code may be used for fuel accountability). The staff considers TAC No. 69028 and its associated review to be closed.

l- Staff Coments i

The staff noted the importance of reliable and safe operation of the fuel handling machine in order to complete the defueling operation. The staff noted l that in its January 20, 1989 letter, PSC had recomended that further Inservice l Inspection and Testing (ISIT) program development be cancelled. The staff noted that the ISIT for the fuel handling machine was of continued importance, and that the PSC should continue its development. Furthermore, the staff requested the PSC consider submitting their overall ISIT program for the fuel handling machine for staff review. The staff noted its intent to formally request this item from PSC.

Y.

Kenneth L. Heitner, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Attachments:

As stated l

1 1

l l l

l i

1

)

4 Attachment 1 PSC/NRC/GA TECHNOLOGIES MEETING MARCH 7, 1989 j Nuclear Regulatory Comission Public Service Company of Colorado )

K. L. Heitner, PDIV/NRR R. J. Hirschl F. B. Litton, DEST /NRR C. H. Fuller L. Kopp DEST /NRR 'M. H. Holmes l T. F. Westerman, RIV M. Niehoff i P. W. Michaud, RIV M. J. Fisher )

D. Warembourg l S. Fisher -

GA Technologies ORNL

, D. Alberstein V. Malakhof S. J. Ball l A. J. Kennedy D. L. Moses S. P. Munoz JRB Technology J. R. Brown l

l

Attachment 2

)

AGENDA FOR NRC MEETING

. TOO DISCUSS FSV DEFUELING AND COASTDOWN I. OPENING COMMENTS NRC l II. INTRODUCTION (D. Warembourg) ^

l DEFUELING PRESENTATION I I L

I. INTRODUCTION (M. Holmes) {

l.

II. DEFUELING STRATEGIES SVALUATED (M. Fisher)

{

III. OVERVIEW OF MAJOR DEFUELING CONSIDERATIONS (M. Fisher)

IV.

CORE PHYSICS ANALYSES AND DUMMY BLOCK DESIGN (S. Fisher)

V.

USE OF DUMMY BLOCKS (M. Fisher)

VI.

REACTIVITY MONITORING AND SHUTDOWN MARGIN ASSESSMENT (M.

Fisher)

VII. PHYSICAL DEFUELING ACTIVITIES (M. Fisher)

VIII. LICENSING ASSESSMENT (M. Holmes)

IX. CONCLUDING COMMENTS (D. Warembourg)

COASTDOWN PRESENTATION I. PROPOSED POWER PLAN (M. Holmes) l II. DESCRIPTION OF CYCLE 4 LIMITS AND TECHNICAL-SPECIFICATIONS (M. Holmes)

< III. COASTDOWN ANALYSES AND TESTING (S. Fisher)

IV. LICENSING ASSESSMENT (M. Holmes)

V. CONCLUDING COMMENTS (D. Warembourg)

VI. CONCLUDING COMMENTS NRC

& ~ ,- ,.

)

1 WHERE ARE WE NOW PRELIMINARY DECOMMISSIONING PLAN

-Bases i Defuel Component Removal '

Safstor w/ Fuel Storage ,

Safstor w/o Fuel Storage l

Decon/ Dismantle

-Plan Status OVERVIEW, OTHER ACTIVITES

-Segment 9/ DOE /ISFSI

-Segment 10

-Early Dismantlement >

1

-conversion OTHER ISSUES

-Fuel Handling Machine

-Financial Plan / Minimum certified Limits l

-Part 50 License, Downgrades / Extensions- i l

l 4

_ _ - - . - _ . - - . . - - . . - _ _ _ - - _ - - - _ - _ - - . _ _ _ _ _ .A

-i PSC FUTURE SCHEDULED MILESTONES l

I o Preliminary Decommissioning Plan to NRR March 31, 1989 o Defueling Plan and Associated Tech Specs to NRR May 31, 1989 l o Fueling Handling Machine Interim Upgrade ~ October 1989 o Fuel Handling Machine Final Upgrade February 1990 o Receive Dummy Blocks October 1989 o NRR Approve Defueling Plan / Tech Specs by November 1989 o ISFSI Part 72 Submittal to NRR February 1990 o ISFSI Ready to Receive Spent Fuel April 1992 o Proposed Decommissioning Plan to NRR December 1991 o Downgraded Part 50 Submittal to NRR December 1991 4

L_m _ _ _ _ _ _ _ . _ _ . _ _ _ _ _. _ _ _ _ _ _ _ _ _ . _ _ _ _

, .g s e 4

9 PURPOSE OF MEETING 1

DEFUELING' CONCEPTS / PLANS COASTDOWN PIAN / ANALYSIS l

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i SHUTDOWN MARGIN ASSESSMENT ASSUWTIONS

• THE~ HIGHEST WORTH CONTROL ROD PAIR CAPABLE OF BEING WITHDRAWN IS FULLY WITHDRAWN.

• CONTROL ROD PAIRS BEING WITHDRAWN FOR REFUELING /

REPAIR, SHUTDOWN MARGIN ASSESSMENT, OR OPERABILITY TEST PURPOSES, ARE FULLY WITHDRAWN.

l

' ALL OTHER OPERABLE CONTROL R00 PAIRS ARE FULLY '

INSERTED AND INCAPABLE OF BEING WITHDRAWN.

• INOPERABLE CONTROL ROD PAIRS ARE IN THEIR KNOWN POSITION OR FULLY WITHDRAWN.

• FOR PLANNED CORE ALTERATIONS, THE CORE SHALL BE'IN .

ITS MOST REACTIVE CONFIGURATION.

• A CORE AVERAGE TEMPERATURE OF 80 DEGREES F.

• FULL DECAY OF XE-135, FULL BUILDUP OF SM-149, AND PA-233 DECAY AS A FUNCTION OF TIME AFTER SHUTDOWN.

v-

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OVERALL DEFUELING' STRATEGIES i

t A) BY REGION:

(1) DEFUEL BY AGE OF THE FUEL IN THE REGIONS (2) DEFUEL AROUND THE STARTUP DETECTORS (4) DEFUEL FROM OU'IZR RING TO INNER RING B) LAYER OPTION: INSERT SEGMENTED CONTROL RODS; DEFUEL THE CORE BY LAYERS 1

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_ _ _ _ _ _ _ _ - - _ - _ _ _ _ l

- ~ ,

1 CRITERIA FOR EVALUATION (1) LICENSABILITY METHODS.

AND COMPATIBILITY WITH EXISTING OPERATIONAL AREAS OF CONCERN INCLUDE:

A) COMPUTER MODELS (REACTOR PHYSICS, STRESS, THERMAL)

AND THEIR APPLICABILITY FOR THE DEFUELING METHODS B) REACTIVITY CONTROL C) REACTIVITY MONITORING D) ACCIDENTS AND SAFETY ANALYSIS (2) TIME / MOTION CONSEQUENCES AND THE ENGINEERING ASSOCIATED WITH A CHOSEN DEFUELING STRATEGY A) ENGINEERING DESIGN AND TESTING ASSOCIATED WITH ANY NEW REACTIVITY CONTROL EQUIPMENT B) PROCUREPINT LEAD TIMES ASSOCIATED WITH NEW HARDWARE C) PLANT STRATEGY MODIFICATIONS ASSOCIATED WITH A SELECTED D) FUEL HANDLING MACHINE IMPACTS ASSOCIATED WITH A SELECTED STRATEGY e

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1

,I CONCLUSIONS OF THE EVALUATION '

4 2

1 (1) STRATEGY FOR DEFUELING AROUND THE STARTUP CHANNELS WAS ELIMINATED.

REASON: COMPUTER MODELS ARE' QUESTIONABLE FOR ODD GEOMETRIES (2) STRATEGY FOR DEFUELING BY REGION AGE WAS ELIMINATED.

REASON: COMPUTER MODELS ARE QUESTIONABLE i FOR THE RANDOM INSERTION OF DUMMY BLOCK '

MATERIALS IN THE CORE l

(3) STRATEGY FOR INNER TO OUTER RING DEFUELING WAS f ELIMINATED.

REASON: COMPUTER MODELS ARE QUESTIONABLE FOR A 1 LARGE ANNULAR CORE AND ASSOCIATED DECOUPLING OF k FUEL i(ITH THE DUMMY BLOCKS.

. t (4) SERIOUS CONSIDERATION GIVEN TO DEFUELING BY LAYER VERSUS OUTER TO INNER RING DEFUELING BY REGION.

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, l DEFUELING BY REGION versus LAYER COMPARISON -

i (LICENSING. ISSUES)

By Reofon (Outer-Inner) By Layer

1) Computer Models Models sufficient Models sufficient' Use 3-0~ code or l R-Z model '

\

2) Reactivity Control Positive & negative moves Safety of core was prove throughout defueling. up front CRD's in active fuel-must Operators.in Control Room remain operable throughout defueling. have no means to contro reactivity

3) Reactivity Monitoring Rely.on existing SUCS Rely on existing SUCS; geometry changes as- .

core decreases in size

'i

4) Safety Analysis RWA still possible New segmented control throughout defueling rods need to be licensed Thermal Hydraulic Analysis is the same LOFC consequences

_ re-evaluated (no upper Dummy Blocks need to reflector) be licensed

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i DEFUELING BY REGION versus LAYER COMPARISON (ENGINEERING ISSUESL  !

I By Region (Outer-Inner) By layer >-

.)

1) Design ' Dummy Block Configuration is well Need to design and test j known; pattern is similar to a -new equipment for in- '

regular fuel block. '

stallation of new absorbers. ,

J Control rod procurement issues t

l

2) FHM Handle Dummy Blocks FHM more reliable due l

to handling fewer blocks {

l l

3) Defueling Time to defuel is about the same. Time to defuel is about the same.

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i CONCLUSION: EITHER DEFUELING BY REGION OR BY LAYER IS FEASIBLE. , ,

HOWEVER, THERE ARE FEWER LICENSING AND TECHNICAL' ISSUES ASSOCIATED WITH DEFUELING THE CORE BY REGION (OUTER TO INNER RING). THEREFORE WE WILL USE THIS STRATEGY.

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DEFUELING OVERVIEW A) CORE PHYSICS ANALYSES AND DUMMY BLOCK DESIGN SHRINKING CYLINDER (0 UTER TO INNER)

8) 80RONATED DUMMY BLOCKS SIMILAR TO REFLECTOR ELEMENT BUT BORONATED C) REACTIVITY MONITORING MAINTAIN SHUTDOWN MARGIN ASSESSMENT TESTING UNTIL ALL CRD'S CAN BE WITHDRAWN AND STILL HAVE AN ADEQUATE SDM D) DEFUELING ACTIVITIES  !

CRD STORAGE DURING DEFUELING l RCD STORAGE DURING DEFUELING FUEL TRANSFER TO THE FSW'S OR THE FLP f

5 SEGMENTS TO IDAHO l 1 SEGMENT TO FSW'S - EVENTUALLY TO ISFSI E) LICENSING ASSESSMENT SUBMITTAL OF DEFUELING PLAN ON MAY 31, 1989 l

PLAN WILL INCLUDE: l

1) DEFUELING ACTION PLAN

2) DEFUELING SAR

3) DEFUELING TECH SPEC

, 4) C0ASTDOWN SAR FOR INFORMATION ONLY

, v ._. ,

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4

.. l,

-l STATUS OF BLOCK DESIGN A PRELIMINARY DESIGN FOR THE DUMMY BLOCKS HAS BEEN COMPLETED; A FINAL DESIGN ANALYSIS FOR THE DUMMY BLOCKS IS BEING PURSUED.

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- DUMMY BLOCK DESIGN ,

'i OBJECTIVE: PROVIDE A DESIGN FOR DUMMY BLOCKS THAT SATISFIES THE REACTOR PHYSICS, THERMAL, AND OVERALL ENVIRONMENTAL REQUIREMENTS OF THE CORE.

ONE IMPORTANT ASPECT OF CRITICALITY SAFETY IS THE REACTIVITY CHANGES CORE.

ASSOCIATED WITH THE INSERTION OF PURE GRAPHITE IN THE FSV FUEL IS NEUTRONICALLY UNDERMODERATED.

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COMPUTER CODES AND HODELh.FOR THE DEFUELING ANALYSES GAUGE:

2-D NEUTRONICS RADIAL MODEL. IT IS THE WORKHORSE CODE FOR FSV REFUELING SAR ANALYSES. IT WILL BE USED FOR~THE DETERMINATION OF SHUTDOWN MARGINS. THE EXTERNAL SOURCE OPTION ALLOWS ASSESSMENT OF FLUX DISTRIBUTION.IN SUBCRITICAL CORES.

GATT: 3-D NEUTRONICS MODELS. IT IS USED FOR REFUELING SAR I CALCULATIONS FOR AXIAL POWER PEAKING. A SOURCE OPTION IS AVAILABLE.

l-DIF3D:

- 2-D OR 3-D NODAL AND FINITE DIFFERENCE NEUTRONICS MODEL.

IT HAS NOT BEEN USED FOR SAR CALCULATIONS. IT IS USED BY PSC FOR FUEL ACCOUNTABILITY CALCULATIONS, EXPLORATORY ANALYSIS, AND FOR INDEPENDENTLY CHECKING GATT CALCULATIONS. A SOURCE OPTION IS AVAILABLE.

POKE:

USED FOR STEADY STATE THERMAL-HYDRAULIC CALCULATIONS.

ALSO USED FOR ASSESSING THE THERMAL-HYDRAULIC ASPECTS OF FLUCTUATIONS.

RECA:

USED TO ANALYZE LOFC ACCIDENT CONSEQUENCES.

BLOOST: REACTOR KINETICS CODE USED TO ANALYZE THE CONSEQUENCES l ASSOCIATED WITH A ROD WITHDRAWAL ACCIDENT.

J 4

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.s SEQUENCE CONSIDERATIONS l

SEQUENCE CHANGES ARE STILL BEING MADE DUE TO THE FOLLOWING CONSIDERATIONS:  ;

I l

4 A) FHM MOVEMENTS AND DETERMINATION OF WHICH BLOCKS PSC WILL 1 STORE ON SITE

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1 1

B) SHUTDOWN MARGIN ASSESSMENT REACTIVITY CALCULATIONS J

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l C) ENHANCEMENT ' OF INCREASED COUNT RATE ON THE STARTUP -j CHANNELS l

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. PROPCSEG SRERRL DEFUELENG SEQUENCE 1 STRRTEGY cstauenct no. s) l l

l 37 20 36 35 @ 21 e

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. PROPOSED .

SPIRRL DEFUELING SEQUENCE

.STRATEG Y cstautsce no. s) cW1 2 IIEREIED 'I

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(FROM OUTSIDE TO CENTER)

RcGION NUMBcR 3@

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PROPOSED .

SPIRRL DEFUELING SEQUENCE \

STRRTEGY ;l tstauence no. sr  ;

case s l 17 LE71EL G i 20  ;

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. 1 CORE PHYSICS- j i

I THREE CASES WERE CHOSEN TO SET THE PRELIMINARY l

DESIGN OF THE DUMMY BLOCKS. 1 ASSUMPTIONS:

1) CORE AT 155EFPD (CURRENT STATUS)

2) NO SAMARIUM l

3) CORE AT 80 DEG F

4) USE DEFUELING SEQUENCE No. 5

5) REGION 1 ROD COCKED AT ALL TIMES

6) NEXT 2 RODS OUT

7) USE 7 GROUP GAUGE CODE EIGENVALUE MODEL i

_ CASE 1: 2 DEFUELED, 1 + 37 + 28 UNRODDED  !

PURE GF.APHITE KEFF = .9794 12 LPP KEFF = .9367 CASE 2: 11 DEFUELED, 1 + 25 + 33 UNRODDED '

PURE GRAPHITE REFF = .9976 12 LPP KEFF = .9345 ,

CASE 3: 17 DEFUELED, 1 + 22 + 13 UNROUDED PURE GRAPHITE KEFF = 1.0096 12 LPP KEFF = .9227 i

, ~

BORONATION OF DUMMY' BLOCKS OPTIONS:

(1) HOMOGENEOUS BORON DISPERSION IN HLM GRAPHITE USING THE BORON NITRIDE PROCESS l

(2) USE LUMPED POISON PINS (LPP)

A)'USE 1/2" PINS l B) PROVIDE A UNIFORM DISTRIBUTION OF PINS C) USE SELF SHIELDING FORMULA FOR LBP FROM THE GAUGE MODEL D) USE 20 WT % NATURAL BORON FOR LPP AS AN INITIAL VALUE FOR THE DESIGN

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ANALYSIS RESULTS IN TERMS OF REACTIVITY EQUIVALENT:

6 PINS IS COMPARABLE TO 200 PPM HOMOGENEOUS 12 PINS IS COMPARABLE TO 500 PPM HOMOGENEOUS 24 PINS IS COMPARABLE TO 1000 PPM HOMOGENEOUS SUGGESTED REACTIVITY CONTROL IS GREATER THAN 200 PPM NATURAL BORON (HOMOGENEOUS EQUIVALENT)

CONCLUSION: 12 LPP/ BLOCK IS OUR CURRENT DESIGN i

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. - l LUMPED POISON BLOCK' DESIGN e O O O O O O O O O O O O O  !

O O O O O O.O O O O .

O 9 0 0 O O e 0 0 O O O

O e e O l O O O of \

O O O O G O O e O O 1

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s DEFUELING ANALYSES AND EVALUATIONS ,

(1) THERMAL HYDRAULICS: SAME AS FOR SHUTDOWN CONDITIONS; HOWEVER, SOME l COOLANT H0LE TOLERANCES MAY BE SLIGHTLY LOOSENED. ANALYZE THE VERY SMALL DECREASE IN FLOW RESULTING FROM THE' DECREASED NUMBER OF COOLANT ~ HOLES. - l (2) STRESS ANALYSIS: DUMMY BLOCKS WILL PLACE LESS STRESS ON THE CORE SUPPORT FLOOR.

(3) ENVIRONMENTAL ANA'.YSIS:

A) HLM GRAPHITE IS USED AT FSV IN THE LARGE SIDE REFLECTOR ELEMENTS.

B) LBP PINS AND RSS MATERIAL ACCEPTABILITY WILL HELP FORM THE LICENSING l BASIS FOR LPP USE IN THE CORE. I C) REACTIVITY ANALYSIS'WILL PROVIDE JUSTIFICATION FOR LOOSER SPECIFICATION (AS COMPARED TO LBP) ON BORON CONTENT FOR LPP. t L I e

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.~

DUMMY BLOCKS I A) BASIC CONFIGURATION I

1) HLM GRAPHITE - MECHANICAL PROPERTIES MEET FSAR REQUIREMENTS

2) NO CRD OR RSS HOLES

3) SAME COOLANT HOLE DESIGN

4) ACCOUNTABILITY s

B) BOR0NATED GRAPHITE BLOCKS

1) BORON NITRIDE GRAPHITIZATION PROCESS I

2) TESTING OF MECHANICAL PROPERTIES l

3) LEACHING AND ENVIRONMENTAL TESTING

4) PRODUCTION TESTING AND QUALITY ASSURANCE l C) LUMP POISON PINS l 1

1) CONFIGURATION

2) BORON CARBIDE PIN INSERTION

3) ENVIRONMENTAL, CHEMICAL AND MECHANICAL PROPERTIES ARE KNOWN -- TEST PROGRAMS ARE NOT REQUIRED

4) PRODUCTION TESTING AND QUALITY ASSURANCE

5) BORON PIN INSERTION QUALITY ASSURANCE i s l

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1 deriuqeR ebyam rotceteD eht fo gnidleihS emoS )6 deriuqeR ebyam snoitacifidoM SPP )5 deriuqeR si gniniarT wen )4 deriuqeR gnitseT latnemnorivnE )3 i

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W v

. REACTIVITY MONITORING START-UP CHANNEL ADVANTAGES AND DISADVANTAGES e

i ADVANTAGES: 1 1.

1) MAINTAIN USE OF CURRENT METHODS FOR CORE MONITORING' . j

2) MAINTAIN AUTOMATIC PROTECTION WITHOUT MODIFICATIONS j

3) NO RETRAINING OF PERSONNEL IS REQUIRED L

4) SUC RESPONSES ARE KNOWN

5) NO NEW ACCIDENT CONDITIONS ARISE DISADVANTAGES:

j 1) A NEW NEUTRON SOURCE IS REQUIRED -

2)

THE BORONATED PLENUM ELEMENTS SHIELD THE SUC'S

3) INADEQUATE COUNT RATE MAY OCCUR LATE IN THE DEFUELING.

4) MAY NEED TO RETAIN 2 FUEL ELEMENTS FROM SEGMENT 10

5) MAY NEED TO REMOVE THE ABSORBER STRINGS FROM 2 SPECIFIC CRD'S 1

l l

I

_ - - - - - - _ _ - - - . _ ~ . _

e

, 1 REACTIVITY MONITORING Analysis Calculations Based on Least Reactive Core Conditions 1% Boron Loading at End of Coastdown i Ring 3 and 4 Defueled Kaff = 0.78 Reactivity Drops Off Rapidly as Ring 2 is Defueled CONCLUSION: BASED ON THE PRELIMINARY ANALYSIS THE START-UP E UTILIZED CHANNELS WILL _BE l

y 9

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, CORE REACTIVITY TRENDS 4

A) When Region 22 is Defueled Count Rate will Drop (SUC II)

B) When Regien 15 is Defueled Count Rate will' Drop (SUC I) i C) Progresses Multiplied Neutrons Available Drop Off As Defueling D) Detection Becomes More Difficult As Defueling Progresses l

E) Analysis i 1

1) Based on Worst Case - At the End of Coastdown When Reactivity is Lowest for Detection Criteria

2) Preliminary Criticality Calculations Based on Most '

Reactive Core 155 EFPD 12 LPP Design

3) Rings 3 and 4 Defueled Core'Kaff = 0.95 l

Assumptions:

4

1) The Region'1 Rod Is Cocked '

2) All Other Rods in Fueled Regions'are Withdrawn for Core Kaff Values.

l

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PRCPCSEC

. SP1RRL DEFUELEMG SEQUENCE STRATEGY j cstautsce no. s) t 37 20 38 as o e @

a1 e

@ 1s g 34 is @

@ 22

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@ mc ~ @:

17 @ 23 \ suc = ,

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@ 1 3 @ {

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31 6 25

@ 13 @

14 30 @

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LEGEND SPIRAL DEFUEL (FROM OUTSIDE-TO CENTER) morm - 30 scouoce - @

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SHUTDOWN MARGIN ASSEESMENT A) ANALYSIS l

CALCULATIONS WILL BE MADE FOR EACH REGION ASSUMPTIONS:

1 GAUGE WILL BE USED FOR CORE CALCULATIONS 2 CREDIT IS TAKEN FOR CRD'S IN ACTIVE FUEL REGIONS 3 CREDIT IS TAKEN FOR BORON IN DUMMY BLOCKS 4 PA IS FULLY DECAYED 5 NO CREDIT IS TAKEN FOR CRD'S IN THE DUMMY BLOCKS B) SDM ASSESSMENT TESTING TESTING WILL CONTINUE UNTIL SUCH TIME AS WHEN ALL RODS CAN BE WITHDRAWN OUT OF CORE AND IT REMAINS SUBCRITICAL WITH CALCULATED SDM 0F 0.01 Ap TESTING METHODOLOGY - REGION ONE ROD IS C0CKED 1)

PULL THE CONTROL PODS OUT OF CORE FOR THE NEXT TWO REGIONS TO BE DEFUELED

2) RECORD A COUNT RATE FOR THIS CONFIGURATION AND LABEL CR-REF.

3) PULL THE SDM ASSESSMENT ROD (S) TO SHOW A MINIMUM OF 0.01 op AT 80'F. REINSERT THE R00(S).

4) DEFUEL THE NEXT REGION AND INSERT DUMMY BLOCKS

5) RECORD THE COUNT RATE AND LABEL CR-DEFUEL

6) COMPARE COUNT RATE VALUES CR-REF.}CR-DEFUEL C) _ FINAL TESTING DEMONSTRATE THAT THE CORE REMAINS SUBCRITICAL WHEN ALL REMAINING RODS ARE WITHDRAWN WITH A CALCULATED SDM OF 0.01 Ap.

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DEFUELING ACTIVITIES .

A) 5 segments of Fuel to Idaho

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B) 1 Segment of Fuel to FSW's - Eventually to ISFSI C) Defuel Directly to FLP or to FSW's and then to FLP D) Maintain Fuel Accountability by Location and Serial Number E) Dummy Blocks to be Loaded from the FSW's or the FLP F) CRD Storage not Required 1

G) RCD Storage H) Removable Reflector will be Reused l

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. DEFUELING TECletICAL SPECIFICATIONS (DTS) ,

GROUND RULES l

j THE DTS WILL BE BASED ON THE LATEST DRAFTS OF THE TECHNICAL SPECIFICATION ,

UPGRADE PROGRAM (TSUP).

• GROUND RULES DEFINED FOR TSUP CONTINUE TO APPLY l

- DTS WILL BE CONSISTENT WITH THE EXISTING LICENSING BASIS FOR FSV AS EMBODIED IN THE FSAR, MODIFIED AS REQUIRED BY' l DEFUELING ANALYSES. l 1

- DTS WILL INCLUDE TSUP GUIDELINES IDENTIFIED IN PSC LETTER j OF 11/16/84 (P-84498), AND NRC LETTER OF-12/20/84 (G-84473). j i

• TSUP SPECIFICATIONS WITH DEFINED APPLICABILITIES OF SHUTDOWN AND/0R REFUEL- l ING, WITH CALCULATED BULK CORE TEMPERATURE <760*F, WILL BE INCLUDED IN THE i DTS. j j

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• TSUP SPECIFICATIONS WITH DEFINED APPLICABILITIES OF PCRV PRESSURE >100 PSIA l WILL NOT BE INCLUDED.  !

)

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• ANY TECHNICAL SPECIFICATION ISSUES WHICH ARE NOT WITHIN THE ABOVE DEFINED TSUP APPLICABILITIES, AHD ARE NOT DIRECTLY RELATED TO DEFUELING, ARE BEYOND ,

THE SCOPE OF THE DTS EFFORT AND WILL BE TREATED AS SEPARATE LICENSING ISSUES. I i

• CHANGES TO THE TSUP DRAFTS PROPOSED BY PSC WILL BE JUSTIFIED IN WRITING.

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• DTS WILL BE SUBMITTED AS A PROPOSED AMENDMENT TO THE TECHNICAL SPECIFICATIONS ]

PER 10 CFR 50.91.  ;

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DEFUELING TECHNICAL SPECIFICATIONS REACTIVITY CONTROL USE OF BORONATED DUMMY BLOCKS.

WHENEVER DEFUELING OPERATIONS ARE IN PROGRESS THE CONTROL RODS IN REGION I SHALL BE FULLY WITHDRAWN,.' OPERABLE,'AND SCRAMMABLE, AND THE RSD HOPPER IN REGION 1 SHALL BE OPERABLE.

l IN FUELED REGIONS WITH CONTROL RODS ENERGIZED AND CAPABLE OF' BEING WITHDRAWN, THE CONTROL RODS SHALL BE OPERABLE.

i IN FUELED REGIONS WITH CONTROL = RODS WITHDRAWN OR CAPABLE OF BEING WITHDRAWN, THE RESERVE SHUTDOWN HOPPERS SHALL BE OPERABLE (EXCEPT IN A REGION BEING DEFUELED OR BEING PREPARED FOR DE CONTROL RODS WILL BE WITHDRAWN AND INOPERABLE IN.DEFUELED REGION (IN SHIPPING POSITION).

RESERVE SHUTDOWN HOPPERS WILL BE INOPERABLE AND DISCONNECTED DEFUELED REGIONS.

IN l

SHUTDOWN REGIONS.

MARGIN ASSESSMENT ANALYSIS WILL BE PERFORMEDFOR ALL SHUTDOWN MARGIN ASSESSMENT TESTS WILL BE PERFORMED FOR ALL R UNTIL ALL CONTROL RODS ARE FULLY WITHDRAWN, AT WHICH TIME FURTHER SHUTDOWN MARGIN ASSESSMENT TESTING WILL BE DISCONTINUED.  ;

CORE COUNT RATE MONITORING SYSTEM WILL REMAIN IN SERVICE UN ONE REGION IS LEFT.

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2 DEFUELING TECHNICAL SPECIFICATIONS COOLING SYSTEMS DEFUELING TECHNICAL SPECIFICATIONS WILL BECOME EFFECTIVE FOLLOWING A 100-DAY DECAY HEAT COOLING PERIOD AFTER FINAL POWER OPERATION.

ONE TRAIN OF SAFETY GRADE -FORCED CIRCULATION COOLING EQUIPMENT SHALL BE OPERABLE.

ONE PCRV LINER COOLING FLOW PATH SHALL BE OPERABLE.

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MAINTAIN CORE AVERAGE INLET TEMPERATURE BELOW 165 DEGREES F. l DURING DEFUELING OPERATIONS.

j HEET LCO 4.1.9 REQUIREMENT FOR ORIFICE VALVES AT ANY POSITION (ADJUSTED FOR NOMINAL EQUAL REGION OUTLET TEMPERATURE) WITH THE PCRV PRESSURIZED TO LESS THAN 50 PSIA THAT THE. HELIUM COOLANT TEMPERATURE RISE THROUGH ANY CORE REGION SHALL NOT EXCEED 350 DEGREES F.

- OR -

MEET LCO 4.1.9 REQUIREMENT FOR ALL ORIFICE VALVES ADJUSTED FOR EQUAL REGION COOLANT FLOWS WITH THE PCRV PRESSURIZED TO LESS THAN 50 PSIA THAT THE HELIUM COOLANT TEMPERATURE RISE THROUGH ,

ANY CORE REGION SHALL NOT EXCEED 600 DEGREES F.

WITH FUEL IN THE FUEL STORAGE WELLS, ONE WELL'C00 LING WATER COIL SHALL BE OPERATING, AND A BACK-UP COOLING COIL OR FAN SHALL BE OPERABLE. )

WITH FUEL IN THE FUEL HANDLING MACHINE, ONE COOLING WATER COIL SHALL BE OPERATING AND A BACK-UP COOLING WATER COIL SHALL BE OPERABLE.

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k DEFUELING ACCIDENT ASSESSENT PRELIMINARY SEISMIC TORNADO CORE HEATUP INADVERTENT CRITICALITY ROD WITHDRAWAL  ;

l HEAVY LOADS l

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HELB LOEP STATION BLACK 0UT FIRE SPENT FUEL SHIPPING CASK HANDLING l

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l CYCLE 4 COASTDOWN DAYS ACCUMULATIVE REACTOR AVERAGE INTO CYCLE 4- POWER FUEL TEMP COAST 00WN- EFPD  % DEGREES F 0 300 80 1373 .

75 360 80 1373

. 118 390 70 1332 l-l 176 425 60 1242 i

236 455 50 1182 l

324 490 40 1116 424 520 30 1026 p

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COASTDOWN TECHNICAL SPECIFICATION LIMITS

• MAXIMUM BURNUP 1800 EFPD-(LCO 4.1.1)'

' RADIAL REGION PEAKING FACTORS (LCO 4.1.3)

• INTERREGION PEAKING (COLUMN TILT) FACTORS (LCO 4.1.3)

• AXIAL POWER PEAKING FACTORS (LCO 4.l.3)

• MAXIMUM CONTROL ROD WORTH (INTERIM LCO 3.1.5)

• CORE SHUTDOWN MARGIN (INTERIM LCO 3.1.4)

• TEMPERATURE DEFECT (INTERIM LCO 3.1.7)

• DETECTOR DECALIBRATION (LSSS 3.3) 1 I

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COASTDOWN FSAR CRITERIA GRAPHITE STRESS LIMITS (FSAR 3.4.2.1.1)  ;

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.d EOL FUEL ELEMENT BOWING (FSAR'3.4.2.1.2) l 1

• 1 PEAK FUEL TEMPERATURE (FSAR 3.6.3.3) i FISSION PRODUCT RELEASE (FSAR 3.7) .

I FAST FLUX EXPOSURE (FSAR A.2.2) l 1

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'e REACTIVITY EFFECTS DURING COASTDOWN (1) A GAIN IN REACTIVITY DUE T0-LOWER FUEL TEMPERATURES AT LOWER POWER LEVELS (TEMPERATURE FEEDBACK)

(2) A GAIN IN REACTIVITY DUE TO LOWER POISON LEVELS-AT LOWER POWER LEVELS (PRINCIPALLY XENON AND SAMARIUM)

(3) PRODUCTION OF U233 PARTIALLY COMPENSATES FOR THE U235 DEPLETION AND INCREASED QUANTITY OF FISSION PRODUCTS 6

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CONTROL ROD SEQUENCE FOR CYCLE 4 AND COASTDOWN i

Group i Secuence Withdrawn Recions 1 2A(a) 2, 4, 6 - i 2 4F(a) 25, 31, 37 3 40 23, 29, 35 1 4

1(115"out) 1 i

. 5 48 21, 27, 33 6 2B 3,5,7 7 4E 24, 30, 36 8 4A

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20, 26, 32 9  !

4C 22, 28, 3'4-  !

10 3C 10, 14, 18 11 3A 8, 12, 16 12 38 l 9, 13, 17 13 i 30 11, 15, 19 14 1(fu11y out) 1 ,

(a) Rod groups used for red runback.

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1 FUEL REGION IDENTIFICATION l NUMBER Identification of control red groups l

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I - _ - - _ - _ _ _ _ - _ - - _ _ _ _ . - - _ . - _

MAXIMUM BURNUP FOR COASTDOWN

- MAXIMUM BURNUP ANALYZED IN THE COASTDOWN SAR IS 1178 EFPD, WELL BELOW THE 1800 EFPD (LCO 4.1.1)

- FSAR MAXIMUM FIMA (FISSIONS PER INITIAL METAL ATOM)

FIMA - PERCENTAGE OF HEAVY METAL BURNED IN A PARTICLE. -

AS COMPARED TO HEAVY METAL. LOADED IN A PARTICLE PARTICLE COASTDOWN_ MAXIMUM FSAR__ MAXIMUM FISSILE 16% 20%

FERTILE 3.4% 7%

. 21

- FSAR MAXIMUM FAST FLUX (E>.18 MEV)' EXPOSURE - 8.2X10 NVT 21 MAXIMUM CALCULATED FOR COASTDOWN -

5'.0X10 NVT e

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POWER PEAXING DEFINITIONS:  !

REGION PEAXING FACTOR (RPF) - POWER IN'A REGION NORMALIZE 0 TO THE AVERAGE REGION POWER IN THE CORE

,4 COLUMN TILT - RATIO'0F THE POWER IN-A COLUMN TO THE AVERAGE j COLUMN POWER IN A GIVEN REGION 1 AXIAL PEAKING FACTOR - RATIOu0F THE POWER IN A LAYER  ;

(USUALLY THE BOTTOM LAYER OF FUEL) l TO THE AVERAGE LAYER POWER IN A GIVEN REGION i

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C Cl F 4 300-520 twD -

RPF-TlLT ENVELOPE UNRODDED REGIONS 1.6

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390 EFFD -

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!  ! LCO 3.1.7: TD (220 %1500') > .031 AK 0.01--~ ~ ~ ~t - . ~ . . t ~ . ~ . . . - ~

j i AT 520 EFPD, TD = .034 AK 0.00 '

b 2bc 4b0 6b0 8b0 10b0 12bo 14b0 t16b0 1800 AVERAGE CORE TEMPERATURE ( F)

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WORTH OF CONTROL ROD GROUPS'AND MAXIMUM R00 AT 300 EFPD CYCLE 4 + -m t

Group- Cumulative Max Rod. RWA Groues In Worth, Ak Worth, ok Worth, Ak Recion RR(1) 0.002 0.002 0.002 1

+3D. 0.016 i 0.018- 0.008 15

+3B 0.023 0.041 0.014 17

+3A 0.012 0.053 0.014 17

+3C 0.022 0.075 0.016 18 l

+4C 0.014 0.089 0.014 .15

+4A 0.007 0.096 -

0.017- 11

+4E(2) 0.007 0.103 0.018

13 (3) 0.112 0.215 N/A N/A l

l (1) Regulating rod 115" withdrawn.

(2) Cold criticality at 300 EFPD.

' (3) Groups 28, 48, 40, 4F, 2A and the regulating rod fully inserted to assure

subcriticality. .

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_ _ _ - - _ _ - - - - _ - - _ _ _ _ _ _ _ _ _ _ = - _ - _ - - _ _ _ _ -

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Shutdown Margins - Results Control Rod System 4

Shutdown Shutdown Margins, AK Case Teore, 'F Time,' Days 390 EFPD 520 EFPD CR'Out

_ '1 1

1 220 0 0.175 0.200 0  ;

2 220 0 0.116- 0.143 22 3- 220' O 0.086 0.114 21+22 4 80. 0 0.080 0.108 21+22 5 80 3 0.050 0.079 21+22 6 80 14 0.044 0.073 21+22 I 7 80 28 0.039 0.067 21+22 8 80 56 0.032 0.060 - 21+22 9 80 224 0.025 0.052 21+22 1

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. l EXTENDED.C,YCLE 4 CALCULATED PEAK CONDITIONS VERSUS FSAR INITIAL CORE PEAK VALUES-Parameter Peak Value Peak Value (a)

AxialStress(psi) ,

450 301 I Radial Stress (psi) 200 54.3 Axial Strain (%) (contraction) 3.0 2.0 i Radial Strain (%) (contraction) 0.8 0.8

. Fuel- Element Bowing (in.) 0.09 0.129 1 Fuel Temperature (OF) 2300 2109 (b)

(a) Values calculated using FSAR methods.

(b) Peak fuel temperature in core during Extended Cycle 4.

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COASTDOWN LICENSING ASSESSENT l

CONCLUSIONS

)

OPERATION DURING THE CYCLE 4 COASTDOWN WILL BE CONDUCTED WITHIN TECHNICAL SPECIFICATION LIMITS.

THE WORST-CASE POSTULATED FORT ST. VRAIN ACCIDENT CONDI-TIONS, FOUND TO BE ACCEPTABLE IN VHE FSAR, WOULD NOT BE EXCEEDED DURING THE CYCLE 4 C0ASTDOWN.

THE CYCLE 4 COASTDOWN PRESENTS NO 10 CFR 50.59 UNREVIEWED SAFETY QUESTIONS.

THE CYCLE 4 C0ASTDOWN REQUIRES NO TECHNICAL SPECIFICATION CHANGES.

1 THE CYCLE 4 C0ASTDOWN SAR. WILL BE SUBMITTED TO THE NRC FOR

• INFORMATION BY MAY 31, 1989.

FOLLOWING NFSC APPROVAL, THE BASE REACTIVITY CURVE CHANGES FOR THE CYCLE 4 COASTDOWN WILL BE SUBMITTED TO THE NRC PER LCO 4.1.8 90 DAYS PRIOR TO REACHING 300 EFPD.

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