ML19319D165

Refueling Procedure FP-203,Revision 1, Defueling & Refueling Operations

ML19319D165
Person / Time
Site:	Crystal River
Issue date:	05/09/1978
From:	Mckee P FLORIDA POWER CORP.
To:
References
FP-203, NUDOCS 8003130731
Download: ML19319D165 (47)

Text

R v. 1 5/4/78 4

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'T. 1 Document Section N FORM ATION ONL N C. "R., N uclear REFUELING PROCEDURE FP-203 FLORIDA POWER CORPOPATION CRYSTAL RIVER UNIT 3 DEFUELING AND REFUELING OPERATIONS e.cc-30 2 Ccr. rolv Ds'cM8 'IPcm:r.t:

p,gje,t-ll!ETlli7lc REVIEWED BY:

Plant Review Cec =tittee 9

IM Date Meeting No.

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APPROVED BY:

Nucicar Plant Mana:;cr b /[b t[b >

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8oo313o 7 2 /

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1.0

-PURPOSE 1.1 To describe the procedure to be followed during the defueling and subsequent refueling of the CR-3 reactor.

Procedure Section Fuel Movement in Reactor Building for Defueling 8.0 Fuel Movement in Auxiliary Building for Defueling 9.0 Fuel Movement in Auxiliary Building for Refueling 10.0 Fuel' Movement in Reactor Building for Refueling 11.0 Au1111ary Neutron Detector Movement 12.0 Inverse Count Rate Plotting 13.0 Verification of Loaded Fuel Asse=blies 14.0 Restoration of Activities and Systems 15.0 l

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2.0 DESCRIPTION

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2.1 This procedure is to be used for all defueling and refueling evolutions subsequent to CR-3 initial fuel loading.

?2.2 The loading and accompanying reactivity changes for each assembly are monitored by two or more source range neutron =onitors.

Plots of the subcritical multiplication versus the nu=ber of fuel asse=-

blies will be made during major. increases of reactivity during refueling.

2.3 The only nuclear instrumentation (NI) that is required to be operable during the defueling evolution is the two source range 9

' detectors, NI-1 and NI-2.

2.4 REFUELING SUPERVISOR The Refueling Supervisor is responsible for supervising fuel movement. He is responsible for all decisions to stop and 1

resume fuel loading. He must hold a Senior Operator (SOP)

License.

2.5 REFUELING CONSULTANT The Refueling Consultant is the individual who supplies tech-nical advice and expertise in the area of defueling and refuel-ing operations to the Refueling Supervisor.

He is also respon-Pagu 2 FP-203 Date 5/4/78 Rev. 1 m.-

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sib 13 far maintcining th3 R3fu211ng Log which will contain a list of all problems and major events' occurring during defuel-

-ing_and refueling operations.

2.6 CCOP CCOP refers to the personnel in the Control Center who are licensed operators and who are assigned to defueling and refueling operations only.

These activities will include tag board control, communication with other refueling sta-tions, and Nr manitoring.

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FP-203 Date 5/4/78 Rev. 1 V

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3.0 REFERENCES

f 3.11 CP-116, Standard Cleanlir.ess Specifications

3. 2 "

FP-601, Fuel Handling Equipment Operations 3.3 RP-106, Radiation Work Permit Procedure 3.4 SP-112, Calibration of the, Reactor Protection System 3.5 SP-220, Instrumentation Functional Tests During Refueling Operations 3.6 OP-404, Decay Heat Removal System 3.7 OP-408, Nuclear Services Cooling System 3.8 OP-704, Communications, System 3.9 OP-403, Chemical Addition System 3.10 OP-405, Reactor Building Spray System 3.11 OP-412, Waste Gas Disposal System 3.12 0P-207, Fire Protection Systems 3.13 OP-411, Instrument and House Service Air Systems 3.14 OP-410, Secondary Services closed Cycle Cooling System 3.15 OP-414, Nitrogen and Hydrogen Systems 3.16 OP-416, Domestic Water Supply System 3.17

_0P-418, Demineralized Water System 3.18 OP-419, Liquid Sampling System-Initial Valve Lineup 3.19 OP-407, Liquid Waste Disposal System 3.20 OP-604, Circulating Water System 3.21

' 0P-606, Auxiliary Steam System 3.22 OP-703, Plant Distribution System 3.23 OP-705, Emergency Power-DC System 3.24 EM-203, Classification of Emergencies and Criteria for Evacuation 3.25 Special Nuclear Materials Handling and Accountability Manual Pag. 4 FP-203 Date 5/4/78 Rev. 1

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3. 26..

AI-400, Picnt Op3rcting Qunlity-Ascurenca Manuel Control Documsnt, Section'8.0 y.

.3.27-

'SP-156, Calibration and Equipment Setup For the Auxiliary Neutron Detector System 3

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PdB0 4d' FP-203 Date 5/4/78 Rev. 1

4.0 ENCLOSURES Water Chemistry Log

' Enclosure 2-Fuel / Control Component Move Sheet (sample)

(deleted)

(deleted)

Enclosure.5 (deleted)

CR-3 Cycle 1 Core Load Map CR-3 Core Map Check Form Background Determination Random Distribution Verification and Base Count Rate Determination 0 Inverse Count Rate Calculations 1 Inverse Count Rate Plot s

.Page 5 FP-203 Date 5/4/78 Rev. 1

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50 LIMITS AND PRECAUTIONS 5.1.

'If_any of the conditions of Section 5.0 are not met, fuel handling shall cease. Action shall be initiated to correct the situation so the conditions are met.

No operation which may increase the reactivity of the core shall be made until-the conditions of Section 5.0 are =et.

5.1.1 During defueling, two source range NI monitors, each with visual indication in the Control Center and one with audible indication in the Control Center, shall be operable.

5.1.2 During refueling at least two source range NI monitors (NI-1 and NI-2), each with visua' indic,ation in the Control Center and one with audible indication in both the Control Center i

and containment, shall be operable.

If Step 13.4 of this procedure cannot be met, i.e., if average counts are less

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than 30 cpm and a random distribution of counts cannot be verified, an auxiliary neutron monitor must be used.

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auxiliary neutron monitor must have both visual and audible indication in the containment.

i 5.1.3 The reactor shall be subcritical for at least 72 hrs. prior to movement of irradiated fuel in the reactor pressure vessel.

5.1.4 The containment penetrations shall be in the following status:

The equipment door closed and~ held in place by a minimum a.

of four bolts, and b.

A minimum of one door in each air lock closed, and V

Page 6 FP-203 D.ite 3/4/78 Rev. 1

c.

Each penetration providing direct access from the contain-ment atmosphere to the'outside atmosphere shall be either:

1.

Closed by isolation valves, blind flanges, or manual valves, or 2.

Be capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve.

5.1.5 Direct communications shall be maintained between the Control Center and personnel at the defueling or refueling station.

5.1.6 All fuel handling equipment operations will be done per FP-601, fuel Handling Equipment Operations.

The fuel handling bridges

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Page 6a FP-203 Date 5/4/78 Rev. 1

shall be used for movement of control rods or fuel assemblies

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and shall be OPERABLE with:

A hoist min,imum capacity of 3,000 lbs., and a.

b. 'A hoist overload cut-off limit of 2750* lbs.

5.1.7 ammmmmmmmes 5.1 8 At least one decay heat removal loop shall be in operation.

5.1.9 The containment purge and exhaust isolation system shall be OPERABLE.

5.1.10 The water level in the refueling canal shall be maintained at the 157 ft. elevation, or higher.

5.1.11 The water level in the spent fuel (SF) pools shall be main-tained at the 157 ft. elevation, or higher.

5.1.12 The Auxiliary Building ventilation exhaust system servicing the storage pool area shall be OPERABLE when irradiated fuel is stored in the SF pools.

5.2 Access to the RB and fuel handling floor area in the Auxiliary Building will be controlled per the Radiation Work Permit (RWP).

5.3 All personnel must be logged in and out of the R3 at the Con-trol Point. The Co trol Point will be at the RB personnel hatch. This is to ensure personnel accountability in the R3.

5.4 The reactor, transfer canal, and SF pools will be Class "3"

clean (CP-ll6, Standard Cleanliness Specifications) prior to tilling.

• When handling control rods or fuel assemblics with > 23 ft. of water over the top of fuel assemblies seated in the reactor corc; otherwise, the hoist overload cut-off shall be set at 2880 lbs.

Page 7 FP-203 Date 5/./78 j

Rev. 1

l 5.5 All accessible areas and equipment in the RB and on the fuel handling floor in the Auxiliary Building will be Class "D" clean (CF-116, Standard Cleanliness Specifications).

5.6 Reactor coolant (RC) chemistry shall be maintained as follows:

a.

Fluorides ;E 0.15 ppm b.

Chlorides ;E 0.15 ppm c.

Boron > 1950 ppm 5.7 Personnel involved with fuel handling shall be limited to working a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> out of'any 24 hr. period.

5.8

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... s.

5.9 All movements of fuel assemblies and rod assemblies shall be made in accordance with the Special Nuclear Materials (SNM)

Handling and Accountability Manual.

A Control Center Operator shall follow the movement of each fuel assembly and auxiliary neutron detector on the fuel move =ent board in the Control Center.

5.10 The RB evacuation alarm shall have been tested and verified operable per SP-323, Evacuation and Fire Alarm Demonstration, within one week prior to the defueling and refueling evolutions.

The alarm shall also be tested af ter any taintenance work on the alarm equipment during these evolutions.

5.11 Any system which might decrease the boron concentration by inserting non-borated water will be isolated from the RC system-refueling canal.

Page 8 FP-203 Date 3/27/78

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'5.12 An cntry chall balmada in th2 Wstar Ch mistry Log (Enclosura 1)-

prior to, and'once each shift during, the defueling and refuel-ing evolutions.

5.13 In the event of a damaged fuel assembly or a reportable occur-rence involving a fuel assembly, it will be removed and rein-spected and approved for use by the Nuclear Plant Manager prior to insertion in the core.

a 5.14 In the event of an accident involving nuclear fuel or the sound-ing of a radiation monitor alarm, personnel shall evacuate the gff'ected area.

The area gamma monitoring subsystem which will monitor the fuel loading operations are listed below:

a.

RM-G13, Decontamination Pit Area b.

RM-G14, Fuel Storage Pool Area RM-G15, Auxiliary Building Fuel Handling Bridge c.

d-RM-G16, RB Fuel Handling Bridge t

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RM-G17, R3 Near Personnel Access Hatch f.

RM-G18, RB In-Core Instrument Arae g.

RM-Gl9, RB - Building Dome EM-203, Classification of Emergencies and Criteria for Evacua-tion, will be used to evaluate the situation and plan a course of action.

5.15 If any disagreement occurs between persons responsible for defueling and refueling (including B & W and FPC personnel)

.which cannot be resolved by mutual agreement of all parties 1

involved, the Nuclear Plant Manager shall resolve the disagree-ment.

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Page 9 FP-2 3 Date 3/27/78

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5.16 During movement of auxilisry ntutron hold 2ra, tha cscociatcd fuel handling bridge will require its low load limit bypassed.

7 This bypass must be removed when fuel moves are going to be initiated.

5.17 If refueling operations are delayed for eight hours or more, all neutron monitors being used must be response checked within eight hours prior to resuming fuel loading.

To accomplish this, pull one source-fuel assembly approximately three-fourths of the way'out and observe the response on all operable monitors.

Rein-sert the, source-fuel assembly and again observe the responses on all operable monitors.

If a response is not observed on all eperable monitors, repeat the above steps with the other source-fuel assembly.

After a response has been observed on all operable monitors, defueling or refueling may be resumed. A new base count rate should be established for each monitor and recorded on Enclosures 9 and 10.

An explanation of the occurrence, including the count rate with the source three-fourths with-drawn and reinserted plus the new "ZZ" tape reading of source-fuel asser.oly reinserted, shall be recorded on the back of.

This movement of the source should also be indi-cated by a symbol or identifying number on the SNM Move Sheets and referenced in the explanation on 2._losure 9.

5.18 Cease core alterations immediately if the neutron count rate from two neutron monitore unexpectedly doubles during any single fueling or defueling step. Withdraw the fuel assembly Page 10 FP-203 Date 5/4/78 Rev. 1

in qu2stion, parform a boron analysis of the RC, and racord on.

Core alterations shall not resu=e until the Refuel-ing Consultant.has evaluated the occurrence and the Refueling Supervisor has given permission to resume cot. alterations.

5.19 If the count rate on two or = ore neutron =onitors increases by a factor of two or more while no geo=etry changes are being made, i==ediately begin boration of the RC system until the count rate stops increasing.

The Refueling Consultant and Refue, ling Supervisor shall investigate the cause of the count rate increase.

Core alterations shall not resume until the cause is determined and rectified.

5.20 If the count rate on two or more neutron conitors decreases by a factor of two or = ore while no core geometry changes are being made, investigate to determine the cause of the count rate decrease.

Core alterations shall not resume until the cause is deter =ined and rectified.

2 5.21 All fuel assemblies should be oriented with the identification plate in the west direction.

All control rod assemblies should have been inserted with their serial numbers in the sa=e direc-tion as the fuel assecbly identification number.

5.22 During defueling and refueling, the sequence of fuel asse=blies to be moved, the detectors, or the source-fuel assemblies may be codified.

This change in sequence shall be handled in accordance with Section 8.0 of AI-400, Plant Operating Quality Assurance Manual Control Document.

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d 5.23-If tha Co/Ci picta from two n:utrcn monitors indicots.thtt criticality will be reached with the addition of three or less fuel assemblies, refueling will be suspended until a thorough evaluation is made by the Refueling Consultant and the Refueling Supervisor gives permissio.n to continue refueling.

NOTE: -When a' fuel assembly is-loaded directly adjacent to a neutron monitor, the count rate on this monitor is expected to rise sharply and may cause the Co/Ci pro-jection of crit 4cality to predict that the next fuel assembly will be o'ne-third or more of the additional fuel for criticality.

This is to be expected and does not require further evaluation unless one other =onitor's Co/C1 projection predicts a similar situation or exhibits an unexpectedly high count. rate.

5.24 If the refueling increment is reduced to less than one fuel

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assembly (loading by incremental insertions of a fuel assembly) because of excessive suberitical multiplications, the refueling increment shall not be increased again.

' 5.25 Inverse count rate (Co/C1) plots will be made in the Control Center from NI-l and NI-2 counts and the R3 if an auxiliary neutron monitor is in use during the refueling evolution.

5.26 Sli POINTS 5.26.1 The RC water temperature will be maintained between 75'? and 140*F.

5.26.2 The RC system and decay heat removal system boron concentra-f tion shall be maintained within Standard Technical Specifica-tion (STS) limits and greater than 1950 ppmn.

Page 12 FP-203 Date 5/4/78 Rev. 1 i

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W;t0r, chemistry chn11 ba_maintnintd by SP-710, RC, Decry Hast Removal, and RC Makeup' Systems' Chemistry Surveillance Program; and SP-713, RC_ Support Systems' Chemistry Surveillance Program.

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9 I have read-and understand the aforementioned limits and precautions and

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have discussed them with my shift.

Refueling Consultants Refueling Supervisors Duty Shift Supervisors t

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.Date 3/27/78 M

6.0 EQUIPMENT AND PERSONNEL hiQUIREMENTS

- 6.1' COMMUNICATIONS 6.1.1

. Plant Communications System 6.2 SPECIAL VISUAL AIDS 6.2.1 Underwater Lights 6.2.2 Underwater Television Camera 6.2.3 Binoculars 6.2.4 Underwater Viewing Boxes 6.3 COMPUTATIONAL AIDS 6.3.1 iHM 5100 6.3.2 Calculator 6.3.3 Straight-Edge Rulers 6.3.4 Linear Graph Paper 6.3.5 Pencils 6.4 INSTRUMENTATION s

6.4.1 Trend recorders for source range neutron detectors.

6.4.2

. Scaler Timers.(2) 6.4.3 Auxiliary Source Range Neutron Detectors (2) (FOR REFUELING ONLY) 6.5 CALIBRATION SOURCE

' 6. 5.' 1 '

Portable 5-Curie PuBe Source 6.6 PERSONNEL 6.6.1 Plant Operations Staff 6.6.2 Defueling and Refueling Shift:

a.

Refueling Supervisor b.

Refueling Consultant c.. Fuel Handling Bridge Operators

.d.

Fuel Transfer Carriage Operators c.

Data Takers f.

ChemRad Technicians Page 15 FP-$03 Date 5/4/78 Rev. 1

7.0 REQUIRED INITIAL CONDITIONS AND PLANT STATUS 7.l' DEFUELING AND REFUELING Initials /Date 7.1.1-The reactor vessel (RV) closure head has been rem 6ved and stored per FP-401, Reactor Vessel Closure Head Removal and Replacement.

7.1.2 The seal place has been lowered and installed per FP-404, Canal Seal Plate Removal and Replacement.

7.1.3 The plenum has been removed and stored in the deep end of the fuel transfer canal per FP-501, Reactor Internals Removal and Replacement.

7.1.4 All Surveillance Procedures required prior to Mode 6 operation have been performed.

7.1.5 (deleted)

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7.1.6 (deleted) 7.1.7 Connect outputs from the two source range out-of-core monitors to trend recorders.

7.1.8 Connect the output from each source range out-of-core monitor to a scaler timer.

7.1.9_

For defueling and refueling the last fuel asse=bly, audible indication of neutron flux will have to originate from a source range NI.

Any of the following methods are acceptable:

Page 16 FP-203 Date S/4/78 Rev. 1 T

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Have the Control Center' PAX phone on live circuit r.

with~the R3 Operations Floor PAX phone. Place the Control Center PAX phone next to the speaker of

'the audible source range NI and have an R3 Data Taker monitor the circuit during removal of the first fuel. assembly, b.

Place the Control Center PAX phone to the "PA" position and move it next to the source range NI speaker.

This will provide audible indication in the containment during removal of the first fuel assembly.

c.

Place either of the auxiliary detectors with audible indication in the RV atop the irradiated fuel with a fuel handling bridge.

This can only be done if the low load limit on the fuel handling

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bridge has been jumpered out.

The jumper is to be installed across relay contact LL-1 (terminal points 226 and 251).

The jumper is to be removed before any fuel is moved-by this bridge.

d.

Any other cethod which satisfies Technical Specification 3.9.2.

7.1.10 All neutron monitors have been response checked within eight hours prior to refueling or defuel-ing.

7.1.11 Current / Fuel Control Component Move Sheets are availabic.

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.Date 5/4/78 Rev. 1

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' 8. 0 FUEL MOVDfENT IN REACTOR BUILDING FOR DEFUELING

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8.1-(deleted) l l

8.2 The Main Fuel Handling Bridge Operator will move to a position above the reactor. core. location designated on the Fuel / Control Component Move Sheet. He will lower the' fuel grapple until he engages the fuel assembly, but shall not raise the assembly until the "ZZ" tape reading has been read to verify proper location of the fuel assembly. Af ter ensuring that a stable count rate has been established on at least two neutron moni-tors and permission has been obtained from the Refueling Super-t visor, the Main Fuel Handling Bridge Operator may raise the fuel assembly into the mast and move it to the upender desig-nated on the Fuel / Control Component Move Sheet.

8.3 Lower the fuel assembly into the fuel transfer carriage, release the assembly, and transfer it into the Auxiliary Building per FP-601, Fuel Handling Equipment Operations, while calling the move into the Control Center.'

8.4 Move the main fuel handling bridge to the next location described on the Fuel / Control Component Move Sheet and repeat Steps 8.1 thru 8.4 for the remaining fuel assemblies that are to be moved into the Auxiliary Building.

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Page 18 FP-203 Date 5/4/78 Rev. 1

9.0 FUEL MOVEMENT IN AUXILIARY BUILDING FOR DEFUELING 9.1 The SF Handling Bridge Operator will posit 3at. the SF bridge over the proper fuel transfer carriage as dcL=rmined by the Fuel / Control Component Move Sheet.

9.2 Lower the grapple tube over the fuel, engage the fuel assembly, and move it to its location in the SF pool per the Fuel / Control Component Move Sheet.

9.3 Lower the fuel assembly into its receptacle, confirm the "ZZ" tapereadingand" low-loddILnit",andreleasetheassembly.

9.4 Move the SF handling bcidge to the proper fuel transfer car-riage as indicated by the Fuel / Control Component Move Sheet.

9.5 Sign off the move as complete on the Fuel / Control Component Move Sheet.

NOTE:.There may be considerable delay in returning a fuel g

transfer carriage to the Auxiliary Building for the next fuel assembly.

This may be caused by detector relocation, establishing new base count races, removing a fuel assembly, etc.

9.6 Repeat procedural Steps 9.1 thru 9.5 until all fuel assemblies are seated in their proper locations established by the Fuel /

Control Component Move Sheet.

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10.0 FUEL MOVEMENT IN AUXILIARY BUILDING FOR REFUELING

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Obtain permission from the Refueling Supervisor and contact t

the Control Center to start fuel movement.

10.1 The SF Handling Bridge Operator shall move the bridge to the

' location described by the Fuel / Control Component Move Sheet.

10.2 The operator will move the fuel assembly from this location to the fuel transfer carriage designated on the approved Fuel /

Control Component Move Sheet.

10.3 The Fuel / Control Component Move Sheet shall be signed off as moves are made. provides a sample of this sheet.

Approved sheets will be provided at the time of actual fuel loading.

10.4 The fuel assembly will be transferred to the R3.

10.5 The move will be called into the Control Center.

10.6 If the next fuel transfer carriage is ready for transferring

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fuel, the Fuel Handling Bridge Operator in the. Auxiliary Build-ing may proceed to the location of the next fuel assembly listed on the Fuel / Control Component Move Sheet.

NOTE 1: Omit all steps that do not involve fuel movement from the SF pools.

NOTE 2:

There may be considerable delay in returning a fuel transfer carriage to the Auxiliary Building for the next fuel assembly.

This may be caused by detector relocation, establishing new base count races, seating a fuel assembly, etc.

10.7 Repeat procedural Steps 10.2 thru 10.6 until all of the fuel assemblies have been removed from the SF pools.

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9 11.0 FUEL MOVEMENT IN REACTOR BUILDING FOR REFUELING

- 11.1 The fuel assembly identification number shall be read and called in to the Control Center before the main fuel handling bridge moves the fuel to the core.

NOTE: The use of underwater television cameras, binoculars, and portable lights may be used to help read the numbers at the fuel transfer carriage.

11.2 The Fuel Handling Bridge Operator in the RB will pick up the fuel assembly "and, move.it to a position above the reactor core location designated'on the Fuel / Control Component Move Sheet.

11.3 The Fuel Handling Bridge Operator in the RB may lower the assembly into the reactor but shall not disengage the fuel grapple until the "ZZ" tape reading has been read to verify the proper seating of the fuel assembly, a stable count rate is established on at least two neutron monitors, and permission to release the assembly is obtained from the Fueling Supervisor or his licensed designee.

NOTE:

If there is any difficulty seating the assembly, refer to 17-601, Fuel Handling Equipment Operations, Step 8.2.8.

11.4 The Fuel / Control Component Move Sheet will be initialed as completed and the "ZZ" tape reading recorded.

NOTE:

Omit all steps that do not involve fuel or detector movement in the reactor.

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'11.5~

Rep 3:e pr cidurcl St ps 11.1 thru 11.4 for orch fu21 ecccmbly to be moved from the fuel transfer carriages listed on the c

Fuel / Control Component Move Sheet.

NOTE: Move sequence steps 347 and 350 do not require moving fuel from the fuel transfer tubes, therefore, follow the instructions below for these steps.

11.6 (deleted)

I 11.7

' Move sequence steps 347 and 350 involve moving the source-fuel assembly.

The Fuel Handling Bridge Operator in the R3 will move the main fuel bridge to the proper location, then he will perform procedural Steps 11.1 thru 11.4.

11.8

-(deleted)

O 11.9 (deleted)

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FP-203 Date 5/4/78 Rev. 1 r-~

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12.0 AUXILIARY NEUTRON DETECTOR MOVDfENT 12.1-An. auxiliary neutron detector is to be moved by hand using a Tope.

12.2 The detector holder is to be moved by a fuel handling br'dge.

i (The low load limit switch will have to have a jumper installed (write-up) since the holder weighs less than the low load limit.]

12.3 The detector torpedo suspended from the nylon line will be lowered into the holder.

During the last two to three inches of insertion of the detector torpedo, th.e nylon line should be slowly moved in a circular motion and the detector seating verified using binoculars.

NOTE:

The seating of the detector torpedo is to avoid a large unexpected count rate iner, ease (a factor of 0.5 to 1.5) when it does seat.

This is apparently due to a plane of i

fuel assembly grid spacers in close alignment with the active center line of the auxiliary detectors and the source.

12.4 Make sure the detector cable is made stationary and placed so that it will not interfere with subsequent loading of fuel assemblies.

NOTE: When an auxiliary neutron detector is relocated, a new base count rate must be established for that detector.

12.5 The Fuel / Control Component Move Sheet will be initialed as completed at each refueling station.

12.6

'When permission is given by the Refueling Supervisor or his licensed designee to load the next fuel assembly listed on the SNM Move Sheets, repeat Steps 11.1 thru 11.4.

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/4/78 Rev. 1

12.7 In order co provide the required audible signal until the final fuel assembly is loaded, use one of the choices pro-vided by Step 7.1.9.

12.8 Remove the jumper from the low load limit switch on the auxiliary fuel handling bridge after detector holder "A" is removed from the reactor.

'12.9 Dispose / decontaminate the auxiliary detectors per ChemRad instructions.

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13.0 INVERSE COUNT RATE PLOTTING

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13.1 Determine background for each of the neutron monitors within eight hours prior to the start of defueling/ refueling evolu-tions.

The monitors shall have been in operation at least 30 min prior to background determination.

13.2 The background count rate for each monitor will be taken for five intervals of five minutes and recorded on Enclosure 8.

NOTE:

Background count rates are to be. established for each

, detector in order to verify that the background will have an ins'ignificant addition to the measured count rates for Co/Ci plots.

13.3 A base count rate (Co) for each monitor will be established after the second fuel assembly (C20) is loaded in the core.

The base count rate is determined from averaging five equal counting. intervals.

The data is to be recorded for each monitor on a Base Count Rate Determination Form (Enclosure 9).

13.4 If the average count rate is less than 30 cpm or the counts are not randomly distributed as determined in Enclosure 9, notify the Refueling Supervisor. An auxiliary neutron monitor will have to be used if the problem is not resolved.

13.5 Each time an auxiliary neutrcn detector is moved, a new base count rate (Co) will be determined for that neutron monitor 3

. system. A base count rate does not have to be determined if detector "A" is suspended over the core during Step 12.7.

13.6 Each time a source is moved, a new base count rate (Co) will be determined for all neutron monitors.

Page 25 FP-2'3 Date 5/4/78 0

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13.7 Aftcr scch fuel caccmbly is incartid into tha core, obrarva the trend recorders for the two source range neutron detec-tors.

13.8 When a stable count rate is reached (trace from trend recorders approximates a straight line), determine if the count rate has increased by 50% over the base count rate (Co/Ci = 0.67).

If not, the next fuel asse=bly can be removed; return to Step 13.7.

13.0 If th,e count rate increases by 50% over the base count rate, then Co/Ci plots shall be performed per Steps 13.10 thru 13.12.

13.10 Record three successive counts, the average count rate, and Co/C1 value on Enclosure 10 and plot the Co/Ci value on 1.

13.11 A very conservative method of criticality prediction is to use the last Co/C1 point and the previous Co/Ci point which gives the steepest slope and extrapolate to Co/Ci = 0.

13.12 If the Co/Ci plot shows no change in multiplication (10 or more points do not vary from the average Co/Ci value by more than 1 0.1), then a new base count can be established; return to Step 13.7.

13.13 If the Co/C1 plot for a neutron monitor value falls below 0.25 when a fuel assembly is moved next to a monitor or a source, the situation should be evaluated.

If it is determined that spacial effects caused the value of Co/Ci to be less than 0.25, then a new base count rate should be established for that moni-

-tor.

Page 26-FP-203 Date 5/4/73 Rev. 1

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=..

-13.14 If more.than three fuel assemblies are predicted to'go critical, inform the Refueling Supervisor or his licensed designee. After two Co/Ci plots' predict that it is safe to load the next fuel i-i i

j i

J 1

I t

i

- Page - 26'l ~

FP-203 Date 5/4/78 f

Rev. 1 i.

9mmne -

1,-r--

'f accembly, tho Refus11ng Sup3rvicor can inform th2 rsfu211ng personnel _that the next fuel assembly can be loaded.

13.15 If less than three fuel assemblies are predicted to go critical, try to determine the cause of this prediction.

(The Keff of one of the assemblies enriched to about 2.8% at cold, clean conditions with 1800 ppm boron is about 0.6 and the Keff of the new clean core for the same conditions is about 0.87.)

Consult with the Refueling Consultant about your analysis and seek his advice.

If the prediction.is due to spatial loading effects and not d'se to excessive subcritical multiplicat' ion, determine if the reading should be ignored for that monitor.

The Refueling Con-sultant should then use the data from the other monitors and determine the effect that the last fuel assembly loaded had on the subcritical multiplication.

If it can now be shown that more than three fuel assemblies are needed to go critical, inform the refueling personnel that the next fuel assembly may be loaded.

If less than three fuel assemblies are still indicated, further analysis should be done with the assistance of B & W.

NOTE: When a fuel assembly is loaded directly adjacent to a neutron monitor, the count rate on this monitor is expected to rise sharply and may cause the Co/Ci pro-jection of criticality to predict that the next fuel assembly will be one-third or more of the additional fuel for criticality.

This is to be expected and does not require further evaluatio'n unless one other monitor's

~

Co/Ci projection predicts a similar situation or exhibits an unexpectedly. hip,h count rate.

Page 27-FP-203 Date 5/4/73 Rev. 1

Whin a fusi eccambly is loadsd dircetly cdjtesnt to a

/

neutron source, the count rates-on all the neutron moni-

~

tors are expected to rise sharply. ' The Co/Ci projections of criticality from two or more neutron nonitors are not expected to predict that the next fuel asse=bly will be one-third or more of the additional fuel for. criticality.

If they do, the situation should be carefully analyzed.

13.16 If the Co/Ci plot for a neutron monitor value falls below 0.25 when a fuel assembly is moved next to a monitor or a source, the situation should be evaluated.

If it is deter =ined that spatial effects caused the value of Co/Ci r.o be less than 0.25, then a new base count race should be established for that moni-tor.

i 1

i

-\\

Page 28 FP-203 Date 5/4/78 Rev. 1 e

-+

n g

4 14.0 VERIFICATION OF LOADED FUEL ASSEMBLIES 14.1'

.Two individuals will separately record the fr=1 assembly iden-tification number and orientation for each core location af ter core loading on Enclosure 5.

All fuel elements must be facing vest.

i

'A Page 29 FP-203 Date 3/27/7S l

m o

15.0 RESTORATION OF ACTIVITIES AND SYSTEMS 15.1 Continue decay heat flow.

- 15.2 Accumulate all data and submit it to the Reactor Engineer.

15.3 Replace reactor internals according to FP-501, Reactor Internals Removal and Replacement.

15.4 Reinstall the reactor vessel closure head according to FP-401, Reactor Vessel Closure Head Removal and Replacement.

15.5 Being cleanup of fuel transfer canal per ChemRad instructions.

t

~u Page 30 FP-203 Date 3/27/78 m

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FUFL / CCh1LCL C Cr7ChihT - m3VE SHEtt - CONTR'.L ROCM C.,,Y PAGE 1

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.w 33 TP-203 I) ate 3/27/78 ENCLOSURE 2 (Page 2 of 3)

SHEET AuM LO 78-07 4

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ENCLOSURE 2 (Page 3 of 3)

ENCLOSURE 3 I'

0 3

I-9 2

(deleted) i s

3

+

a Par,e 35 Fl'-203 Date 5/4/78 i~

Rev. I 1

x 6

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3 Page 30 FP-203 Date 5 /?./ 78 4

(next page 41) gey, 1 3

I r.

k>.

_. - - ~. -

/

(deleted)

I 1

Pat;e 41

[. p_ vo3 Date 5/4/78 Rev. 1

- =... _._ w.;_ u.i T

.s CR-3 CYCI.E 1 CORE LOAD MAP 15 14 13 12 Il 10 9

8 7

6 5

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2 1

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-u-ENCLOSURE 7 CR-3 CORE MAP CHECK FORM CALLED NORTl!

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1 l'ap' 43 -

1.p-203 Date'3/27/78 l

-, - - ~

,.v.

w w

,4 m+

BACKGROUND DETERMINATION

/

Detector Time Initials Date Record counts for five counting intervals for five minutes each before loading fuel.

1.

counts per 5 min. =

cpm 2.

counts per 5 min. =

cpm 3.

counts per 5 min. =

cpm 4.

counts per 5 min. =

cpm 5.

counts per 5 min. =

cpm Sum of Counts =

cpm j

Average Background (BG) = Sum 5 cpm t

BG =

cpm l'ar,e 44 -

FI'-203 D ite 3/4/78 Rev. 1

RANDOM DISTRIBUTION VERIFICATION AND i

BASE COUNT RATE DETERMINATION i

Detector Initials Background (BG) cpm (fred Enc. 8)

Time /Date

/

NOTE: Count to a minimum of 200 counts or to a counting time of five minutes.

-(A minimum count time of 10 sec. is required.)

Take ' data for five equal counting intervals.

Ni (counts)

Ca - Ni (Ca - N1)2 N1 =

counts N2 =

counts N3 =

counts N4 =

counts N5 =

counts Sum = S1 =

counts Sum = S2 =

81 Average Count P. ate = Ca = 5 x Counting Interval in Min.

I

=

= cpm 5x

(

Random Distribution Verification 1 P1 = S2 Ca Is "P" >.75 but < 13.40?

/

(Yes or No/ Initials)

IF YES:

Two source range NI's are required for refueling.

(Ca must also be greater than 30 cpm.)

IF NO :

Two source range NI's and one auxiliary n6utron monitor are. required for refueling.

Base Count Rate Determination

1. ~Ni successive counts are Ca i 2s6Ea, 1 2s/2Ca =

2.

Ca is greater than 30 cpm.

3.

Ca is greater than 5 x BG.

5 x BG =

If all three conditions are met,'then:

Base Count Rate (Co) = Ca Co =

cpm i

IThis is known as a Chi' square test which is used to determine how well data fits a Poisson Distribution of l'robability.

1. imi t s.75<P< 11.40 represents a' good; fit.

l.

l'ar.e 45 FP-203 Date 5/4/78 Rev. 1

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