ML17317A755

Lic DPR-58 & DPR-74 Appl for Amend to Increase Spent Fuel Storage Capacity from 500 to 2.050 Fuel Assemblies.W/Encl Description & Safety Analysis for the Fuel Storage Capacity Expansion Prog

ML17317A755
Person / Time
Site:	Cook
Issue date:	11/22/1978
From:	Maloney G INDIANA MICHIGAN POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 7811270171
Download: ML17317A755 (80)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM DOCKET NBR:

-3 5 16 COOK.1 RECIPIENT:

DENTON, H.R.

. DOC DATE: 781122 ACCESSION NBR:

7811270171 ORIGINATOR MALONEY, G.P.

COPIES RECEIVED:

COMPANY IN & MI'PWR I,TR ENCL

SUBJECT:

SIZE:

41 Lic 8DPR-58

& DPR-74 Appl for Amend to increase spent fuel storage capacity from 500 to 2,050 fuel assemblies.

W/encl description, safety analysis

& environ considerations.

NOTARIZED

b?STRTRUTIDN CODE)

APPI DISTRTRUTION TITLEI f'Eh)ERAL DISTRIBUTION FOR AFTER ISSUANCE OF'PERATING L?CENSE, NAME BR CH F

NRC PAP I

L E

OELb HANAUEP CORF PFRFflRMANCE BR AD FOR SYS L F'RDJ ENGINFFRING HP REACTnp SAFFTY HR PLANT SYSTEMS BR EER EFFL'ljEhlT TREAT SYS J

MCQDUGH LPOR TERA NSIC ACRS ENCL?

'W/7 ENCI W/ENCL W/ENCL W/?

ENCL l TR ONLY W/ENCL w/ENCL w/E~CL W/ENCL W/ENCL w/ENCL W/ENCL W/ENCL w/ENCL w/ENCL w/ENCL W/ENCL W/l5 ENCL FOR ACTION 0Raf/1 aC

~q)$

TOT Al.

NUMBER QF

~ COPIES REQUIRED I LTR ENCL 40 39 NOTES-I & E 3 CYS ALL MATL

INDIANA R MICHIGAN PONIER COMP NY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 November 22, 1978 AEP:NRC:00105 Donald C.

Cook Nuclear Plant Units 1

& 2 Docket Nos.

50-315 and 50-316 License Nos.

DPR-58 and DPR-74 T Spent Fuel Storage Capacity Expansion Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

Indiana

& Michigan Power Company hereby transmits, attached to"this letter, the first part of its application to increase the Donald C.

Cook Nuclear Plant spent fuel storage capacity from 500 to 2050 fuel assemblies.

The proposed modification would remove all of the existing spent fuel racks in the spent fuel pool and replace them with high density poison spent fuel racks.

The attached application provides a description of the proposed modification, safety analysis and environmental considerations.

The infor-mation contained herein is presented in a format discussed with your staff on October 31, 1978.

The required structural and thermal analyses will be transmitted to you by December ll, 1978.

Unit 1 of the Donald C.

Cook Nuclear Plant is scheduled to refuel in the spring of 1979.

Unit 2 is scheduled to refuel in the fall of 1979.

It is necessary to perform the proposed modification during a period of time when neither of the two units is refueling since the spent fuel pool is a facility shared by both units.

In order to maintain a full core dis-charge reserve at all times during the spent fuel rack change-out

period, we plan to complete this proposed modification before the Unit 2 refueling

,which is scheduled to commence the first part of October, 1979.

Work on this modification is scheduled to commence on August 1, 1979.

This schedule will "also keep the occupational exposure as low as possible during the replacement period.

Ae P

h j~

V 4

,P

• Mr. Harold R. Denton, Director AEP:NgC;00105 Your immediate attention to this matter is requested, Very truly yours,

. Malone ice Presi'de t

<rr,

,6 Sworn,and subscribed to before this z'iz day of November, 1978 in New York County, New York Notary Public h.h.'lllLl a N D MY 740'tA~Y it.'eUC, State of Ncw'or No. 4l-4606'i92 Qualified in Clucens County'ertificate filed in Hcw York Coun+

Conumssron rxprres rrturch 30, gled cc:

R.

C. Callen G. Charnoff P.WE Steketee R.J. Vollen R. Walsh D. V. Shaller-Bridgman R.

W. Jurgensen

g>

e

~

ATTACHMENT 1':

INDIANA 5 MICHIGAN POHER COMPANY DONALD C.

COOK NUCLEAR PLANT UNIT NOS.

1 AND 2 DESCRIPTION AND SAFETY ANALYSIS FOR THE SPENT FUEL STORAGE CAPACITY EXPANSION PROGRAM

'I NOVEMBER 1978

TABLE OF CONTENTS

~Pa e

1 ~ 1 1.2 1.3 1.4 1.5 1.6 2.

2.1 2.2 2.3 2.4 2.5 2.6 2.7 3.

3'.l.

3.1.1 3'. 1:.2 3:.1:.3:

3:.1.4 3'.1. 5 INTRODUCTION History and Need for Replacement General Description Specific Needs Construction Costs Alternatives to Increasing the Storage Capacity Commitment of Material Resources RADIOLOGICAL EVALUATION Solid Waste Liquid Waste Gaseous Waste Normal Operation Dose Rates Occupational Exposure.

Due to Change-Out of Spent Fuel Racks Nonradiological Effluents Impacts on the Community SAFETY ANALYSIS Criticality Considerations Cri.ticality Cri:teria Cal culati onal Methods Desi.gn Bas'e Fuel Assembly Descri:ption Storage Array Description Results 9

10 12 13.'3 13" 1'4 1,5 1'7

TABLE OF CONTENTS CONT'd Pacae 3.1.5.1 3.1.5.2 3.1.6 3.2 3.3 3.4 3.4.1 3.4.2 3.5 3.5.1 3.6 3.6.1 3.6.2 Fuel Assembly Reactivity Calculations Storage Array Reactivity Calculations Systematic Uncertainties and Bench-mark Calculations Fuel Handling Considerations Cask Drop Consequences Material Considerations Poison Verification Program In-Pool Surveillance Program Thermal Considerations,.

Fuel Assembly Heat Removal Mechanical Considerations Design Criteria Methods of Analysis References I

18 19

~29 29, 29 30 30 31 31 31 32 3.3.

LIST OF TABLES Pa<ac Table 1 -

1 Table 3.1-1 Table 3.1-2 Table 3.1-3 Table 3.1-4 Table 3.1-5 Table 3.1-6 Table 3.1-7 Table 3.1-8 Estimated Refueling Schedules Fuel Assembly Parameters Infinite Media Multiplication Factors H

17 x 17 Fuel Assembly Reactivity Sens itivity {CCELL)

Reactivity Calculations Boron Sensivity Reactivity Calculations Calculated KEFF Values for Cylindrical Rod Hater Critical Lattices Calculated KEFF Values for ORNL Critical Lattices Calculated KEFF Values for Bierman Critical Lattices 8

21 22 23 24 25 26 27 28 13:1

I

~ i<

~ ~

I i

LIST OF FIGURES

~Pa e

Figure 1 -

1 Figure 1

2 Figure 1 - 3 Location of Spent Fuel Pool 5

Typical High Density Poison Spent Fuel Module 6

Typical Spent Fuel Storage Cell

I

~

~

1.

Introducti on In March 1978, the Nuclear Regulatory Commission published the 0 ftG i

K i 1I S

0 t

H dli

~d5t of S ent Li ht Water Power Reactor Fuel NUREG-0404 The findings o

that report in icate t at t e tec nology of water pool storage is well developed and that the storage of LWR spent fuels in water pools has an insignificant impact on the environment.

Also, the physical security measures required for protection against sabotage of stored spent fuel are essentially the same at both reactor and away-from-reactor sites,

hence, increased spent fuel storage at either location has little relative safeguards significance.

The report also identifies at-reactor compact storage as the most favorable economic cost for providing additional time to develop a program for the final disposal of high level wastes, while allowing for economic and safe nuclear electrical'generation.

1.1 Histor and Need for Re lacement Unit No.

1 of the Donald C.

Cook Nuclear Plant achieved initial

.criticality on January'8, 1975 and Unit No.

2 on March 10, 1978.

,Unit No.

1 is currently in cycle 3 and Unit -No.

2 is in its first

'.cycle.

The spent fuel storage pool was designed under the then valid

assumption that yearly fuel cycles would be utilized requiring storage

'of a single batch of spent fuel for less than one year in the spent

'fuel pool for each unit.

'Due to the Government's decision to indefinitely defer the re-

,processing of nuclear fuel, we are forced to seek methods to solve the

spent fuel storage crisis that is before us.

Whatever alternative to

".reprocessing is chosen, our customers will be affected by higher

.electricity cost.

Replacing our current spent fuel storage racks is,

'we believe, the safest and most economical means of serving our customers.

i1..2

~General Descri tion

'The spent fuel pool is a facility shared by both Unit No.

1 and

.Unit No.

2 and is located in the Auxiliary Building between the two

'Containment Buildings.

The general location of the spent fuel pool is

.'shown in Figure l-l.

The present storage capacity is 500 assemblies.

/

g)Indiana 5

Michigan Power Company has entered into a contract with iExxon Nuclear Company, Inc. of Bellevue, Washington for the design,

'analysis, and fabrication of replacement spent fuel storage racks iw'hich will permit the storage of approximately 2050 fuel assemblies iin;the.spent -,fuel pool..

.These replacement.spent fuel storage racks

, >w'i!1!1, provide;storage capacity. and allow..fpr the continued,operation. qf.

)both.Unit'No.

1 and Unit No.

2 until approximately the first part of 0992 while still maintaining the capacity, for a full core discharge preserve

'(FCDR) of 193 locations.

~

~

I p

The replacement spent fuel storage racks are to be fabricated primarily from type 304 stainless steel.

The individual fuel assemblies will be stored in square fuel storage cells fabricated from stainless steel-clad Boral

• material.

The high density (poison) spent fuel module construction, is essentially a replica of the design used in the replacement racks for the Salem Nuclear Generating Station, which the Commission staff is currently reviewing.

The module is shown in Figure 1-2.

The design utilizes a stiffened module base and an upper box structure consisting of plate diaphragms and a top grid.

The vertical loads are carried by the module base.

Horizontal seismic loads are carried to the module base through the plate diaphragms.

Tipping is prevented by coupling adjacent racks through a bolted connection at the top grid level.

The detailed design of the spent fuel storage cells is slightly different from th'e design for the Salem Nuclear Generating Station.

Their basic function and construction, however, are similar.

Figure 1-3 illustrates the storage cell design for the Donald*C.

Cook Plant.'ach cell is a square cross-section formed from an inner shroud of stainless

steel, a center-sheet of aluminum clad B C, and an outer shroud of. stainless steel.

This cell acts as a stoppage space and, in addition, provides sufficient neutron absorption by the boron carbide contained in the Boral sheet to allow spacing of spent fuel in a 10.5 inch by 10.5 inch array.

The fuel weight is carried directly on the module base.

A flared guide and'ransition section is provided at the top of each storage cell.

Thig transition is designed to assure ease of entry and to preclude fuel assembly hang-up and damage.

1;.3

~Site N

d Indiana 8

Michigan Power has a contract with Allied-General Nuclear Services (AGNS) for fuel reprocessing services.

Currently, however; no spent fuel can be sent to AGNS for reprocessing due to the December 23, 1977 NRC order terminating licensing proceedings for the Barnwell facility.

Presently, there are 129 spent fuel assemblies stored in the

. spent fuel pool.

Sixty-five assemblies were discharged from Unit No. 1-in, January, 1977.

The remaining sixty-four assemblies were discharged in April, 1978.

One hundred and twelve burnable poison clusters are contained in these assemblies and an additional thirty burnable. poison cl.usters occupy.storage locations.

The total storage capacity expected to be utilized is; based.on

'maintaining a ful'1 core discharge reserve"storage capability;"

'The'stimated refueling schedules and expected number of fuel assemblies to be transferred into the spent fuel pool are given in Table 1;1;..

From this table, it can be seen that the existing 'storage capacity would'e filled.by May, 1980 with FCDR.

" Tinademark of Brooks and Perkins incorporated

The proposed modifications to the spent fuel pool do not affect the rate of spent fuel generation or the total quantity of spent fuel generated during the anticipated operating lifetime of the facility.

Due to the Government's present position on spent fuel, this

proposed expansion will not change the time period that spent fuel

assemblies would be stored on-site since we have no place to ship them at this time, and do not foresee shipping them 'off-site until the early 1990's.

The expansion will, however, allow operation of both units for an additional thirteen years without shipping spent fuel off-site.

1;4 Construction Costs

.The total cost associated with the project is expected to be

=approximately

$4.7 million.

This estimate includes the following items:

2'2.

3.

! 5.

. Project management, design, quality assurance and licensing.

Materials, tooling, and hardware fabrication.

Removal, installation, transportation, and disposal

,of the old racks.

Contingency allowance.

Allowance for funds used during construction.

!1.5

'Alternatives to Increasin the Stora e

Ca acit

,Currently, spent fuel is not being reprocessed on a commercial i,basis in the United States.

In addition, spent fuel storage at an off-ssite facility is not available at the present time nor is likely,to be aavailable before 1981 when our present storage capacity will no longer be adequate.

~Shipping spent fuel to another reactor site is not possible since

~the"American Electric Power=System has no other nuclear plants.

With

".the present situation in spent fuel storage capacity, we cannot rely

'n another utility to provide storage space for us.

'urthermore, both the Nuclear Fuel, Service's and the General EElectric Company's reprocessing plants are in a decommissioned state.

TTheir fuel storage pools are available only in a very limited capacity tto: a'ew of their original customers.

Me do not have access to this

". 0torage.

The Allied-Gulf Nuclear Service.plant is not licensed to

ooperate, and cannot be depended upon for receipt of spent fuel, due.to tthe termination of the licensing proceedings f'r the plant.

""1H'hereiare no~independent"spent fuel storage facilities available

<<at'his'ime.

And due to the uncertainty of licensing such a facility, wwe -do not foresee such a facility as being available in.the next ten

'-years.

Even if an off-site storage facility were available, we'roject it'to be more economical to store spent fuel on-site and find that there

. are no environmental benefits associated with off-site storage compared with< our proposed action..

4-If the:reactors were unable to refuel due to the existing racks being full, we would be forced to seek replacement of up to 2148 megawatts net electrical energy pr'oduction.

In the short term our system would have to increase utilization of more expensive generation means along with seeking to purchase energy from outside the system.

In most instances,.

we would not be able to purchase this large amount of energy.

Then we would have to eliminate or curtail service.

This

. would have a severe adverse socio-economic impact on the customers and communities we serve.

If it were possible to purchase this energy, we estimate that the additional cost to our customers would be approxima-tely $1.5 million in today's dollars for each day that the reactors

.were not operating.

In the long term the energy production would have to be replaced

",with new generation facilities.

Replacement generating units could not

'be built and placed into service until 1986 at the very earliest.

The installe'd cost of a generating unit to replace the idle Donald C.

Cook capacity time is projected to be.an investment of more than

$ 2. 1 billion.

.1;6 Commitment of Material Resources The proposed modification will utilize racks made of stainless

.steel, boron carbide and aluminum.

These materials are readily available in abundant supply.

The material requirements for this one time installation are in-

.significant compared to the annual national use of these materials and ufo not represent a significant irreversible commitment of natural resources.

Based on the evaluation of these alternatives and the commitment

of resources; we have concluded that increased on-site storage must be

'provided since there is no place available to ship spent fuel and

shutting down the facility would cause grave'economic hardship.

In

'addition, in. order.not to lose FCDR capability, this modification to

.the spent fuel rack must be completed during the summer of 1979, between

the refuel.ings of-the two units.-

~4 ~>>rw at<<it<<tv<<<<I.~f&

FIGURE l-l LOCATION OF SPENT FUEL POOL r's

~.. -C----

4 I s I

I v

thar I

~t l

I lsoc7'Ia

~ 4 ~

~lsf Ta 4

~f<<ITI<<S <<1 ~ o

~tarot taott tattc l

~I<<IICC 444s4 4 at<<C Ita 1

r)I ht.t 0,&gP 84'!

CO<<ta so <<tVt ae I I

~ tsaa t ~

'"'o.

Q I I'Ig Q Q

~ co<<tarvia vvt 4 I

rat cs I+

~roa al MEZZ 4 q<L I

II(I'sf

I.

jg

3Q 4 at<<Cia<<

ITsf<<Ii)

~

g~"

~aorta OCIAI IsIat 4

~ ea V'I'IrtVoting ~

elI CT

~t <<ra vc VlI I;.~

~al ave<<<<

~ oaaliOO<<san I AC\\t I I aata fr

~

~ ~te Ita c

~

~ ra 4 oaa ~ ~. ~

~ Ia

~I

~ Iia IC<<IC<<SIO I V I ~

"I

~ 4<<

~I ~'I ts

~Iat sa go v ta ~ ~ '

~ c!<<e a

lar Il I Itl

~ la tert ~ vl

~

'I ~ Iv Vi e vl~'so.t it~

~

st la ta I ~ r Itl ~ <<

~ I I~ ~

' ato rt

~ t v

l

". ~

~

o e ~ ~

rt

~ I ~

~ 1

~.:. "..M:,

I

~are r r ~

I~,

~

~ rs

~ ~ 1 O I

~ S Shoe Ilio rase ss ~ e

)'

aa

~

~

~ ~IO I

C <<t&S

~

ooo 4 C sciv Tao<<I Avlo al ~

at f

~SC ~ Wac s

coro srl coats<<.

ecclso ot!a<<;

<<I s 'I Qt ssrel aa ll~ ~'gvovtat 4attlt

~illII

<<4 ssl

~otss floe stt tttteg+

~Ila I

I ISs-I lect sttel Svot Iiscc 4 Iat A<<TOO

~t L-r

~ I

'vs<<lI pd IassC ls,f Z

~--

i I) hg si at <<alta'

~a I,.)

be+44 I

tai taaet IIC~I as Cel I

'Lljl Aiv h

~

se r

~

I a ir<<>>r ltsoof tvea<<sl rl sattoloererae tro lt tl tloalafoo.ll tl r ~ sr o ~ I ~ /

CJ '-,,

J)'.o La%sr o

fC Tv rI;I144<<

~ 4 ]

~ Ile ~

Stt&SIr s ~

t,l I<<a<<re at ass<<a oa. ~Iar ~

Iat h<<a'IOA I I

1 ~

tl

) i yvaesa 44 6-I

/~

II l

.I'I

~tlI g,~L+2 4

4 I~@ C I C I]~taro+

44:P ~s.9 WII~~~ ~C~C<<'C~

0r 0

Os 0

Os 0t

~IS'l ~

~ If.

TII.

-IP-Ilf M'

re rem<<

~

r

~

tahe'.9 I Cit<<

~<<< I It<<II 0'h SI ~ t

~ro; SCAICV Ssaatt Ott Cl Sll I v if sf )

CC toilet ~ ~ se r>>

~'liras ~ s<< ls ssastov

~ 1 rote<<,ea<<e sa,r'a,ae<<

~

~

~

~ ~

4 I

0 I

v l

cs

~

I I

MODULE INTER-TIE II I

MODULE BASE RESTRAINT PENT FUEL CELL A l

~

1

)~

~r

~ ~

BUMPER PLATES r

y/

LEVfLINC FEEI (AIXIVSTAGLC RCMQTKLY)

COOLANT FLO>V HOLES SPENT FUEL POOL LINER MODULE BASE FIGURE 1-2 TYP,ICAL.

HIGH DENSI TY POISON SPENT FUEL MODULE.

IOX IO

\\

1

~ ~a 4 ~

SIAINI(SS Sfl(L (ILL(A ~ AS SfAINI($5 Sf((L OVIls SHAOUD (Ofs (HN)

SOAAL H(UIAON ASSOAsls (AOISON ALA(l),

$ (AINL($$ Sf((L IHN(A SHAOUO (015'(NN) o

$ (AL ff(LO SO(N (NO ~

SECTION A-A I~ ~ N(VIAON ASSOA ~(A SIAIN(lss $1((L IHS(AIION OVID(/

VIA(A (iflHSION gfI A

D.C. COOK PROjECT l0 70 FIGURE 1-3 TYPICAL SPENT FUEL STORAGE CELL

DONALD C.

COOK NUCLEAR PLANT Unit No.

2 Re fueling Date

.Number of Fuel Assemblies Dischar ed Refuel ing Date Number of Fuel Assemblies Dischar ed Cumulative Number of Fuel Assemblies in SFP May 1979 Hay 1980 Hay 1981 May 1982 Hay 1983 Hay 1984 May 1985 Hay 1986 May 1987 Hay 1988 Hay 1989 Hay 1990 May 1991 Hay 1992 64 65 64

.65 64 64 65 64 65 65 Oct.

1979 Feb.

1981 Oct. -1982 Feb.

1984 Oct.

1985 Feb.

1987 Oct.

1988 Feb.

1990 Oct.

1991 80 96 84 88 88 88 88 88 88 129 273 338 646 711 863 10] 5 1080 1232

=

. 1384 1449 1601 1753 Storage Limit.....

1818 with FCDR TABLE l-l 2.

Radiolo ical Evaluation We have evaluated the potential radiological impact associated with possible increased releases of solid, liquid and gaseous wastes resulting from the proposed modifications to the spent fuel pit at the Donald C.

Cook Nuclear Plant and have found them to be environ-mentally insignificant.

Our findings are discussed below.

2.1 Solid Waste Installation of additional spent fuel storage capacity in the SFP will require additional reshuffling of the assemblies, which could result in additional crud (corrosion product oxides) being dislodged from the surface.

While it is unlikely, storing additional decayed spent fuel could result in some additional fission products being intro-duced into the SFP water.

At present the Donald C.

Cook Nuclear Plant Spent Fuel Pit Cooling System (SFPCS) has a maximum system flow rate equal to twice the flow rate required to maintain purity of the pool.

The details of the present SFPCS are discussed in Section 9.4 of the Plant's Final Safety Analysis Report.

Additional spent fuel could increase the amount of radioactivity accumulated in the filters and demineralizer.which are disposed of as solid waste.

This increase, if

'ny, should be mi'nor because this modification increases only the storage capacity and not the frequency or the amount of fuel to be replaced for each fuel cycle.

The amount of corrosion products re-leased into the pool during any year would be essentially the same regardless of storage capacity.

At. presen't we replace the filters in the purification ahd skimmer systems about six or seven times a year.

The demineralizer resins have not required changing to date.

The resin and filter components are changed on the basis of contact radiation levels, pressure

drop, chemical exhaustion or breakthrough.

We do not expect any increase in'he amount of solid waste generated from the spent fuel pool cleanup system due to the proposed modification.

The present spent fuel racks in the SFP will be disposed of as low activity waste.

The volume of the racks would be approximately 11~475 cubic feet.

This disposal will occur once in the lifetime of the plant.

~

. Averaged over the lifetime of the plant, this would increase the total waste volume shipped from the facility by less than l9, %.

This would not;have any significant additional environmental impact.

2!Ri

~Li id W

Normal operation of the SFP purification system generates some radi,oactive liquid wastes which" are processed by the j3ant radwaste treatment system.

In any case since the amount of radioactive liquid waste generated by the storage facility fuel is minor compared to the total volume of liquid waste generated by plant operations,'he proposed modification should add no significant amount of liquid wastes.

s ~

The effect of crud buildup on the wall was investigated by taking dose rate measurements at the middle and at the edges of the pool.

No-significant difference was observed in the readings.

Based on this, it is evident that the crud buildup on the wall has little or

. no contribution to the radiation levels.

Pool leakage is handled by the Spent.Fuel leak protection system.

Any leaking water is routed to the 587'evel of the Auxiliary Building so that observations can be made to determine any leaks.

Water is channeled to the Auxiliary Building Sump, then to the radioactive liquid waste hold up tank, and finally, to the waste evaporator.

Acceptable liquids are released.and the solids are sent off-site for disposal.

This modification should not incr'ease any, liquid wastes generated by the leak protection system.

A table showing most recent gamma isotopic analysis of SFP water identifying the principal radionuclides and their respective concen-trations is shown below:

~lsoto e

Co-57 Cs-.134 Cs-137 Cs-58 tan-54 Co-60 oui ml 2.24 E-5

1. 94.

E-.3 2.63 E-3 1;70 E-3 8.65 E-5 2.47 E-3 2:3; Gaseous Waste Data by year, for the-last two years, for KR-85 measured from the auxiliary building ventilation system are given below.

1976'.

1977'978,

last, half'irst half-last half first hal f 10'.01 Ci 161.40 Ci 13.05 Ci 124.40 Ci Kr-85 Kr.-85 Kr-85 Kr-85

'Note'hat, the only gaseous release data available are for batch releases which include.gas decay tanks and containment releases.

Given, below is a table. of-recent analyses made to determine the.'rincipal:airborne radionuclides.

and their respective concentrations-,

in>>the: SFP. area..

~

~

\\

~Isoto e:

~Ci cc.'-'131 Xe-135 Xe-133 Kr-85 Kr-85m Kr-88 Kr-87 Xe-133m

~ 9.88

~ 1.49

~ 5.73

~ 7.44 c 1.71

~ 4;89

~

~ 3.99

< 1.25 E-15'-.8 E-8 E-6 E-8 E-8 E-8 E-7 2.4 Normal 0 eration Dose Rates In 1978 the filter was changed six times with a total exposure of 0.324 man-rem.

An increase in the frequency of filter changes would increase the exposure proportionately.

No estimates for exposure during resin replacements are available since none have been performed so far.

These will be done remotely so exposures should not be significant.

A recent measurement of dose rate above the pool indicates a

reading of 1 to 3 mR/hr at one foot away from the pool surface.

!2.'5 Occu ational Ex osure Due to Chan e-Out of S ent Fuel Racks We have reviewed our plans for removal, disassembly and off-site

shipment of the old racks and installation of the new racks.

This

.operation is expected to last for a short period of time, no more

.than 20 weeks.

The occupational radiation exposure.for this operation is estimated to be about 20 Man-rem.

We consider this to be a con-

-servative estimate.

This exposure is of the same magnitude as those

expected from other special maintenance operations which are to'occur periodically during the facility lifetime.

Since this is a

.one-time exposure, it is not directly comparable to the annual doses

during normal operation of the SFP.

The increment on on-site occupational dose resulting 'from the

~proposed increase in spent fuel storage capacity has been estimated

'.by using reali'stic assumptions for occupancy times and radioactivity

of 'the spent fuel assemblies themselves.

This modification should

~have a negligible effect on the dose rates in the pool area because

'.o'f the depth of water shielding the fuel.

The occupational radiation

~exposure resulting from the proposed'ction represents a negligible

,burden.

Based on present and projected operations in the spent fuel

pool area, we estimate that the proposed modification will,'add less

than 7X -to the total annual occupational radiation exposure burden at this facility.

The small increase in radiation exposure will not

'-affect our ability to maintain individual occupational doses as low

'as 'reasonably achievable and within the limits of 10 CFR 20.

,'Thus,.we conclude that storing additiona"1 fuel in the SFP will

~not'result in any significant increase in doses received by occupational

workers..

l2.'6 ',Nonradiolo ical Effl.uents

'~There.wi.ll:be no change in the chemical or bioc'idal'effects 'from ithe 'p'lant as a result of the proposed modi'fication,

~

(

~

~

lit

~

The SFPCS is expected to keep the pool bulk water temperature at or below the design value of 120oF during normal refuelings until the modified pool is filled.

A high temperature alarm is located in the Control Room and set at 125oF.

Indicators are in the SFP Cooling System Room.

2.7 Im acts on the Communit The new storage racks will be fabricated off-site and shipped to

'the plant.

No environmental impacts on the environs outside the spent

'uel storage building are expected during removal of the existing racks and installation of the new racks.

The noise impacts generated within this building are expected to be limited to those normally

'associated with metal working activities.

No significant environmental impact on the community is expected to result from the fuel rack

.conversion or from subsequent operation with the increased storage of

.spent fuel in the SFP.

3,.0 SAFETY ANALYSIS 3.1 Criticalit Considerations

'An analysis was performed of the potential maximum reactivity of the fuel stored in the proposed fuel assembly storage facility.

This analysis considered the minimum possible spacing under normal and earthquake conditions, the maximum fuel enrichment level, the most reactive conditions of fuel density, and the most reactive water temperature.

No credit was taken for any boron present in the storage pool water under normal conditions.

3.1.1 Criticalit Criteria The spent fuel storage racks shall be designed such that Keff is limited to a value of less than 0.95 under normal circumstances when'he pool is flooded with demineralized water.

The calculated value shall be less than 0.95 by a margin sufficient to account for calculational uncertainties.

In the analysis credit may be taken for neutron absorption of the stainless steel clad and for the boron carbi'de contained within the Boral plates used in the poison cells within the rack module.

Credit shall not be taken for any burnable poison that may be contained in the fuel.

No credit shall be taken fot boron dissolved in the spent fuel pool water under normal conditions.

I The Keff calculations shall be based on a maximum fuel enrichment 1'evel in new unburned fuel of 3.5 w/o U-235.

An evaluation of all credible'bnormal fuel configuarations shall be made.

Criticality

'calculations shall consi'der any reductions in fuel bundle center-to-center spacing resulting from dimensional tolerances and clearance between the fuel bundle and its storage-cell.

The caTculation shall a~Tsoi consi'der variances i'n boron l,oadings within the Boral plates'ndI deformations under structural loads and'rom abnormal events.

3.1.2 Cal cul ational Metho'ds The methods employed by Exxon Nuclear Company in the criticality safety analysis are the same as those reviewed and approved by the NRC in prior submittals.

The following is a summary of that methodology.

The KENO IV Monte Carlo code was utilized to calculate the reactivity of the D.

C.

Cook Units 1

and 2 storage array.'ulti-group cross section data from the XSDRN 123 group data library were generated for input into KENO IV using the NITAWL and XSDRNPM codes.

Specifically, the NITAWL code was utilized to obtain cross section data adjusted to account for resonance self-shielding by the Nordheim Integral Method.

The XSDRNPM code, a

discrete ordi nates one-dimensional transport theory code,, was then used to prepare spacially cell-weighted cross section data representative of the fuel assembly for input into KENO IV.

The XSDRNPM code was also employed to perform several one-dimensional transport k

calculations to establish the relative reactivity sensitivity of the array to boron content in the storage cells.

'n addition to 'these

codes, the CCELL code was used to 'examine the effects of UO pellet density, moderator temperature, fuel temperature, and enrichment on the infinite multiplication factor of the W 17xl7 fuel assembly.

CCELL is a pin cell calculation code proprietary to Exxon Nucl.ear used primarily to obtain cell averaged multigroup cross section data for rod-water lattices.

3.1.3

.Desi n Base Fuel Assembl Descri tion The fuel storage ar ray is designed to accept fuel enriched up to 3.5 wt.X U. -,From the standpoint of fuel assembly size and'35 infinite multiplication factor, the three assembly

designs, currently in use (W 15x15 or Exxon 15xl5 in Unit.1 or W 17x17 A

in Unit 2; see Tables 3.1-1 and 3.1-2) are very similar.

Hence, differences in pool keff values for storage of different assembly design types are deemed insignificant.

The,fuel assembly specifications and the lattice cell parameters for all three fuel types are given in Table 3.1-1.

The bundle'veraged cell parameters were calculated by including the zir-conium associated with the control rods and instrument guide tube in the zirconium clad of each fuel rod.

Mater associated with each guide and instrument tube was included by increasing the unit cell dimensions (lattice pitch).

Such assumptions permit a

conservative estimation of the effect on reactivity of the extra

zirconium and water within the fuel assembly.

',The:analysis discussed herein assumes the.storage of ll 17x17

.Fuel

<design at a maximum enrichment of 3.5 wt A

U for all UO fuel 2

rods.

.3;1.4;Stora e Arra Descri tion

The D. C.

Cook Units 1'nd 2 spent fuel storage pool will accom-modate twenty specially designed storage rack modules.

Each rack

~module contains a specific number of fuel assembly locations

,(e;g.,

"110 locations for a 10 x ll module).and -installation calls

'for a 14.8 inch nominal center-to-center fuel-cell separation between adjacent modules.

!Individual fuel assembly storage cells will be manufactured out of stainless steel clad BORAL.,

Each cell guide will have,a Inominal inside =-diameter of.8.969.inches and a minimum wall

'(th'ickness of 0.,194,inches..

'IThe assumed storage cell;wall material thicknesses used 'in the

,.calculation maximize the pool reactivity by minimizing the amount of both poison and water present between adjacent fuel assemblies in the overmoderated array.

Storage cells manufactured to the minimum specified dimensions assure a minimum B loading between 10 fuel assemblies of 0.040 g/cm, assuming a

B/Bn t weight ratio

~ 2 10 of 0.180.

.From a neutronics standpoint, the arrangement of modules in the storage pool results in an essentially infinite.array of fuel assemblies in both the axial and radial directions.

The nominal storage position assumes normal conditions where each unit within the effectively infinite storage array is concentric in its respective.cell.

In addition to the nominally spaced array, the minimum spacing between fuel assemblies and the minimum water gap between adjacent storage cells has been considered.

Specifically, the minimum center-to-center separation between adjacent storage cells will be "gauged" to assure a minimum water gap between cells of 0.953

inches, compared to a nominal water gap of 1.118 inches.

The fabrication tolerances will ensure that the worst credible spacing in the pool array occurs as a cluster, of. four adjacent assemblies with other storage cells being spaced the nominal center-to-center distance from that cluster.

This arrangement also assumes that fuel"assemblies in -the cluster are in contact with the inside of each respective cell.

For the postulated accident condition of a fuel assembly lying horizontally across one or more of the storage modules, criti-cality safety is maintained through neutron isolation.

A fuel

/

assembly lying across the top of the modules would be isolated from other fissile material by greater than 20 inches of water.

This separation between fuel assemblies essentially isolates, from a neutronics standpoint, the horizontal. assembly from those in the module cells and,

hence, there is no significant con-tributionn to the overall reactivity of the array.

3.1.5 Results 3.1.5.1 Fuel Assembl Reactivit Calculations Values of k were computed for the D.

C.

Cook Units 1

and 2

design base fuel assemblies assuming both the nominal and bundle-averaged lattice cell parameters as given in Table 3.1-1.

These cases were examined to provide insight as to the reactivity effects of the instrument and control rod guide tubes within the fuel assembly.

These results (see Table 3.1-2) indicate a slight increase in k due to the increased moderator-to-fuel volume ratio inside the fuel assembly.

In order to evaluate the reactivity sensitivity to changes in enrichment, the value of k was calculated for=

U enrichments up to and including 3.9 w/o.

These values (see Table 3.1-3) would indicate an increase of approximately 6 mk per 0.1 w/o increase 'in the specified range.

To evaluate the potential effects of pool temperature on the reactivity of individual fuel assemblies, values of k were computed for temperatures ranging from 20'C to 100'C.

Calculated values of k (see Table 3.1-3) indicate that increasing the 'fuel assembly temperature results in a decrease in k of approximately 1 mk per 20'C increase.

In addition to examining the potential effects of temperature, the effect of UO density changes was also examined.

For this criticality safety analysis, the U02 density was assumed to be 945 of theoretical.

Since increasing the U02 density decreases the thermal utilization factor for the fuel, k of the assembly d'ecreases with increasing density (see Table 3.1-.3).

3.1.5.2 Stora e Ar ra Reactivit Calculations The KENO IV Monte Carlo code was used to compute storage pool

-reactivities for assumed worst credible conditions.

The bundle-averaged fuel assembly parameters are given in Table 3.1-1.

Reactivity calculations were performed using an effectively infinite representation of the storage array.

In evaluating the overall reactivity of the "as designed" storage

array, assumptions were made with regards to the worst credible

.conditions (from the standpoint of neutronics) that could exist in the pool.

Conditions assumed in the "worst case" reactivity

.calculations include:

1) 3.5.wt X U enriched fresh U02 fuel;

':2.)

Bundle-averaged fuel assembly parameters;

'3)

Minimum water gap thickness of 0.953 inches between adjacent

-storage cells; this accounts for limits on installation

'.tolerances and storage cell deformation due to design

.structural loads, possible earthquake disturbances, etc.

'This represents the worst case geometry for the array.

4)

'Temperature variances (20-100 C) in the pool water;

!5)

'No soluble boron in the pool water.

'For 'the nominal case reactivity calculations, only assumptions

.1:2 'and 5 are utilized, and the pool water temperature is assumed

to,be 204C.

/

fFable 3;1-'4 1'ists results of pool:reactivity calculations 'for E

cboth,the anom'anal and worst.case.conditi.ons.

All worst case (conditions given above are concurrently.considered in a s,angl.e

<oahculation.

'In 'addit'ion 'to these assumpt'ions, the non-oned'ible

<cond'ition of:assuming,the fuel assembly to have a fuel-moderator temperature 'of 20'C and the water between fuel ass'emblies to be at 100'C was made.

This assumption maximizes both the reactivity of the fuel assembly and the interaction between adjacent assemblies.

For this non-credible boundary case, the reactivity

.was calculated to be 0.923

+.004.

In evaluating the effect of storage cell boron loading on array reactivity, several reactivity calculations were performed using XSDRNPN.

A storag'e cell and associated fuel assembly were represented for the calculation by a cylinder with a geometric buckling equivalent to that of the actual fuel assembly.

The

.boron density in the infinite array of storage cells was then varied for each individual calculation.

Results as described in

'Table 3.1-5 would indicate an increase in array reactivity of

~:0.014 ak per 0.010 g/cm B decrease between fuel assemblies 2 10

.in,the specified ranges for this system.

To assure the specified minimum boron loading between fuel assemblies of 0.040 g/cm B,

.the actual average loading between assemblies will be greater than the specified minimum.

.3.1:.6

.S stematic'Uncertainties and Benchmark Calculations

'The calculational methods and computer codes used to assess the

criticality safety of the D.

C.

Cook fuel storage array have been ihenchmarked against current experimental critical experiments, and the 'results of these evaluations are discussed below.'To,veri fy the.adequacy of our cal culational model, a number of

'.t'heory-experiment, comparisons have been made.

For the KENO IY meactivity calculations, the cross section data generation and imethods of analysis employed in evaluating these.crit'ical experi-

<<ments were.the same as previousily described to evalua'te

.the

~l' weactivi.ty,of the 'Doel 3.storage arrays.

Namely, the 'XSDRNPN' end 'NITAHL codes, were used to generate 123 group, cel;1 we'ighted, (2) resonance self-shielded fuel region cross sections for input into

'KENO IY.

One set of critical experiments consists of small, water-moderated critical arrays of fuel rods as described by Grob, et al.

We have evaluated several of'hese UO rod-water lattice critical 2

experiments using the KENO codes with 18 group cross section data averaged using the CCELL code and 123 group data averaged as described above.=

Results of these calculations are shown in Table 3.1-6.

It is noted that the KENO calculated reactivities are in good agreement with previously performed DTF-IV transport (6) theory calculations within the statistical uncertainty of the Monte Carlo calculations.

In addition to these correlations, several benchmark calculations have been performed for critical experiments utilizing 4.95 wt.A 235 U metal, rod-water lattices and 2.35 wt.X U aluminum clad UO rod-water lattices Results of reactivity calculations 2

for the uranium metal rod experiments are given in Table 3.1-7.

Table 3.1-8 lists results for the aluminum clad rod data, recently reported by Bierman, et al.

The theory-experiment correlations show that the analytical methods used adequately reproduce the experimental results.

TABLE 3;1-'1 DESIGN BASE D; C.

COOK UNITS 1

AHO 2 FUEL ASSEMBLY PARAMETERS Nomiiial Bundle-

~Avera ed Westin house 15xl5 Nominal Bundle-

~Avera ed Exxon 15xl5 Westin house 17x17 Bundle-Nominal

~Avera ed Lattice Pitch; iB:

Clad OD Clad Material Clad Thickness; iii; U02 Pellet OD; in:

Pellet Density; X p-Enrichment*; w't;X 235 Active Fuel Rods 0.563 0.422 Zr-4 0.024 0.3659 95 3.5 204 0.5913 0.4256 Zr-4

'0.0261 0.3659 95 3.5 204 0.563 0.422 Zr-4 0.030

'.3565 94 3.5 (same as 1l 15xl5) 0.5913 0.4285 Zr-4 0.033 0.3565 94 3.5 (same as W 15x15) 0.496 0.374 Zr-4 0.0225 0.3225 95 3.5 264 0.5190 ~

0.3773 Zr-4 0.0242 0.3225 94 3.5 264 Rod Array Effective Array Dimeiisiohs; Control Rod Guindd.Tube Dimensions (Zr=4), in.

Instrument Tiibe f3$lehsiohs (Zr-4), in; 15xl 5 in.

8;445x8;445

0. 5450DxO. 017 wall. (upper)

0. 4890Dx0. 017 wall (lower)

0. 5450DxO. 017 wal 1 15x15 8.445x8.445 N/A N/A 17x17 17x17 8.432x8.432 8.432x8.432 0.48200x.016 N/A walllupper) 0.4290Dx.016 N/A wall (lower) 0.4820Dx0.016 N/A wall

~Specified enriclNiMt; TABLE 3.1-2 INFINITE MEDIA MULTIPLICATION FACTORS Lattice Cell Parameters Fuel Assembl k

Values CCELL Il I

5 1515 5

1515 5~It I

1717 Nominal Bundle-Averaged 1.424 1.435 1.425 1'..433 1.418 1.430

23 TABLE 3.1-3 ll 17 x 17 Fuel Assembly Reactivity Sensitivity (CCELL) s U Loading,

/cm ~~sU, axial ENRICHMENT, (w/o):

3.5 3.7 3.9, 44.22 47.25 49.80 1.430 1.442 1.453 PELLET DENSITY,

(% PT):

(at 3.5 w/o and 20 C Fuel

+ Moderator Temperature) 90 94, 97 100 42.34 44.22 45.63 47.05 1.431 1-430 1.429 1.428 FUEL and MODERATOR TEMPERATURE, ('C):

(at 3.5 w/o and 94%

p Pellet Density) 20 60 100 44.22 44.22 44.22 1'.430 1.427 1.423 TABLE 3.1-4 Reactivity Calculations 0.

C.

Cook Units 1 and 2

Fuel Type:

W 17 x 17 (3.5 w/o)

Storage Cell:

Stainless Steel Clad BORAL

- 0.218" total thickness (assumed)

TM GT ID:

8.986" (assumed)

GT Center-to-Center Spacing:

10.50" (nominal)

B Loading:

0.020 g/cm per cell plate 2

Case Descri tion k ff +

0 NITAWL-XSDRNPtl-KENO IV 123 rou Nominal 0.908

+.004 2

Worst Case Geometry and Pool Temperature*

0.923

+.004

• See description of assumed temperature conditions in Section 3.1.5.2.

~

0

~ ~

\\ TABLE 3.1-5 BORON SENSITIVITY REACTIVITY CALCULATIONS NITAWL-XSDRNPM Boron Loadin *,

/cm B

0:030

+0:016 0:040 0:0 0:050

-0.011

,*Between fCiel 'assemblies:

~ I g

TABLE 3;l-6 GALCULATED KEFF VALUES FOR CYLINDRICAL ROD-MATER CRiTICAL LATTICES 235 (2.70 Wt.X 0 Stainless-Steel Clad U02 Rods

)

Square Lattice S acin

, iti 0.435 0;470 0;573 0.615 0;665 t4derator-to-Fuel Volume Ratio 1.405 1.853 3.357 44078

. 4~984 Exp'tl.

Critical Gylinder

Radius, cm 26.820 24.294 23.600 24.771 27.172 CCELL-DTF-IV CalcUlated Reactivity (k,<<)

1.016 1.015 1.011 1.009 1.005 CCELL-KENO II (18-group)

Calculated Reactivity (k ~~)

1.008

+.006 1.014

+.005 1.003

+.005 1.010

+.005 1.005 +.005 NITAML-XSDRNPN-KENO IV (123-group)

Calculated Reactivity (k,~)

1.007

+.005 1.013 +.005 TABLE 3.1-7 CALCULATED KEFF VALUES FOR ORNL CRITICAL LATTICES II

\\

~

(4.95 Wt:X U Unclad Uranium Metal Rods

)

Case

~ ~

L'p$$j,ce Number of

.,:,Critical, Water tIumber Rods Hei't Above Lattice',

cm CCELL-KENO II (18-group) k ff +

6 NITAWL-XSDRNPM-KENO IV (123-group) k ff +

6 1

2 22 23 203 195 Rod-Water Lattice Onl 7.1 15.24 0.988 +.006 0.998 +.006 0.999

+.006 104 (Run) 245

~ <

r/

1 ir,

~

Rod-Water. Lattice +

U 0.185 Block 9.5 1.001 +.006 0.993 +.006

~ r~

Rod-Water Lattice +

U 0. 185 Block + BORAL'Sheet 4

'5 105 (Run) 324 359 15.24 11.94 1.038 +.005 1.037 +.006 1.000 +.005

. TABLE 3.1-8 CALCULATED KEFF VALUES FOR BIERMAN CRITICAL LATTICES (2.35 Mt.X U Aluminum Clad U02 Rods

)

235 (8)

Case Experiment Number Number of'uel Clusters in Array, Fuel Rods Critical Separation Between Fuel Clusters, cm CCELL-KENO II (18-group) k ff +

6 NITAML-XSDRNPM-KENO IY (123-group) kff+

a Rod-Mater Lattice Onl 002 014 8.41 1.008 +.005 1.007

+.005 1.004

+.005 0.991

+.005 Rod-Mater Lattice + 304L Steel 028 035.

6.88 11.47 0.994 +.004 0.997

+.005 0.997

+.004 1.000 +.005 Rod-Mater Lattice + BORAL' 020 3

016 6.34*

9.03 0.995 +.005 1.007

+.005 0.999

+.005 0.999

+.004

• 6.33 cm assumed in reactivity calculation.

.,=29-3.2 Fuel Handlin Considerations An analysis of the consequences of a fuel handling accident was performed in the Final Safety Analysis Report for. the D.

C.

Cook Nuclear Plant (Ch. 14.2.1).

The Nuclear Regulatory Commission's Safety Evaluation Report for Cook Plant concluded that the analysis was acceptable.

The modification proposed for the spent fuel racks would not affect the consequences or probability of that accident, nor introduce a different or more severe accident.

3.3 Cask Dro Conse uences The details of the cask drop protection system a'e presented in guestion 14.15 of the Plant's Final Safety Analysis.

The Nuclear Regulatory Commission reviewed this information and ruled to be acceptable.

The proposed spent fuel rack modification does not involve the spent fuel shipping cask area.

Therefore, the proposed modification does not affect the original cask drop evaluation.

Existing Technical Specifications (T.S. 3.9.7) prohibit travel of loads in excess of 2500 lbs over the fuel assemblies while they are stored in the Spent Fuel Pool.

3 4 Material Considerations All permanent structural material exposed to the spent fuel pool environment that is used in the fabrication of the spent fuel storage racks is 300 series stainless steel mostly 304.

This material was chosen for compatibility with the spent fuel pool water.

At the design operating temperature of 120oF, there is no deteri'oration or corrosion of stainless steel in this environment.

There is also no corrosion problem at temperatures up to and including pool boiling.

All other structural components in the. spent fuel pool'ystem, such as'he pool liner, cooling system pipe'onnections, etc.,

are made of stainless steel.

The Donald C; Cook high density spent fuel storage cells utilize Boral material sealed between an inner and outer stainless steel

'-shroud.

The Boral material will be supplied by Brooks and Perkins, Inc.

to Leckenby Company who will fabricate the spent fuel storage module for Exxon Nuclear Company.

The stainless steel shroud (or cladding) is Type 304, meeting the requirements of ASHE SA240.

The Boral consists of an 1100 series aluminum and boron carbide matrix core sandwiched bet>teen two layers of 1100 series aluminum cladding..

Boron carbide particles act as a neutron absorber..

The boron carbide is ASTfl-C750-74 Type II"or equivalent.

Non-destructive testing of the cells will be conducted to insure 100K leak tightness with a 95Ã confidence level..

In addition.to these

programs, Exxon Nuclear Company will conduct an independent neutron transmission testing program on the completed poison cells.

l l

I

~

~ ~

i J

e g

~

I

- In summary, the pool liner, rack lattice structure, and cell exteriors are all stainless

steel, which has demonstrated good corrosion resistance in PWR spent fuel pool environments.

The design, material selection, and the NDE program provide a high degree of

- assurance that the integrity of the fixed poison material will be maintained.

The material used in the new spent fuel storage racks is similar to present components and does not effect or alter previous evaluations.

3.4.1 Poison Verification Pro rams

. Close control and verification of the material properties utilized in the manufacture of'he Boral is assured through the manufacturer's

. guality Assurance Program and is documented on appropriate material

. certification reports.

Prior to inserting the Boral plates into the

. finished cell configuration, each plate is identified in order to

~ allow traceability to the end product.

Records are generated for each

cell identifying the plates installed in that cell by serial

number, tthereby providing positive assurance that the required plates are in

'place.

During rack fabrication, additional care is exercised to prevent

. damage to the stainless steel cladding of the poison cells.

Tracea-

'bility is continued on the cells by providing a cell location map of eaach fuel storage rack module.

,Special handling measures are imposed on the packaging and shipping

~;to minimize the possibility of degrading the quality of the racks

.during transit.

A thorough receipt inspection at the Donald C.

Cook "Plant is performed to assure no damage has occurred.

Documentation is maintained on all testing and surveillance per-iformed on the poison cells as well. as material certification reports oon all materials used in the construction of the cell.

83.4.2 In-Pool Surveillance Pro ram

~Surveillance specimens are provided to allow for surveillance

,:over the lifetime of the fuel storage racks.

The purpose of these

.specimens is to provide assurance that no unexpected corrosion is ooccurring which could compromise the integrity of the Boral.

The ssurveillance specimens are in the form of removable stainless steel cclad Boral sheets, which are prototypic of the fuel storage cells.

TiThese specimens can be routinely'emoved and examined and, then re-innstalled in the spent fuel pool.

3.5 Thermal Considerations 3.5;1 Fuel Assembl Heat Removal The D. C.

Cook spent fuel racks utilize stainless steel encap-sulated Boral shrouds supported in a'stainless steel structural lattice.

Adequate flow paths to the fuel assembly inlet are provided by sufficient space beneath the racks and between the racks and the pool walls.

A six-inch hole at the bottom of

the fuel storage cell serves as the coolant inlet.

Flow paths between fuel storage cells within a rack module are provided to remove gamma heating of the inter-cell coolant.

'~0i C i

'iThe high density spent fuel storage. rack design provides storage

'capacity for slightly more than 10-1/2 cores (2,050 spent fuel-

'assembly storage cells).

The original fuel storage design

.provides for storage of 2-2/3 cores.

Because of the high density

.,storage (compared to the original design),

the design will be

~reviewed to determine if adequate natural convection cooling is available during normal operation to: (a) maintain fuel r'od clad temperatures at acceptable levels; and (b) preclude boiling within the fuel assemblies.

Fuel rod clad temperatures= will al.so be evaluated under hypothetical loss of forced coolant

.circulation conditions where the pool surface is assumed to

,reach a saturation temperature of 212 F.

3.;6

'Mechanical Considerations

.'I!'6,'.1 Sit'.nuctural des'i.gn,cr i teria;for;spent fuel 'storage racks w'i'1'1 gabe

<developed.to:assure,conformance,w'ith recognized codes and applicable regulatory guides, inc'luding the 'OT Position for

-'Review and Acceptance of Spent Fuel Storage 8 Handling Applica-tions, April 14, 1978.

Methods of Anal sis The methods employed by Exxon Nuclear Company in the structural design and analysis will be the same as those reviewed and approved by the NRC in prior submittals.

The following is a

summary of that methodology:

Structural Anal sis The SAP-4 computer program is used for static and dynamic analysis of the fuel storage rack structure.

The analytical model is sufficiently detailed to allow determination of static and dynamic loads on all rack members and includes rack-to-rack interties and wall braces.

Appropriate boundary conditions for the rack interties and the support point nodes at, the base of the structure are *developed..

'L The mass of the water enclosed in the spent fuel storage rack is,lumped together with the masses of 'the fuel assembly

'and the

. rack structure in the lumped-parameter SAP-4 model.

Static analysis output consists of member loads-and nodal deflections.

Dynamic analysis output includes frequencies, mode shapes,-

participation factors and member loads.

Static 'and seismic loads obtained from the SAP-4 model are combined together and with other loads as required by the criteria to calculate stresses in the structural members.

The calculated stresses are then compared with the applicable allowable stresses to confirm structural adequacy.

/

Non-"..L~inear -Effects.:

. Time hist'ory>>analysis 'of;a'ingle 'fuel.:storage. cel 1/fu'e't.>> a'ssembly

".v to account for, the'effe'cts

'of the; clearance gap between.the.',.

storage cell wall and the'fuel. assembly'*will'e performed.

The

0 method of analysis will be identical to that submitted by Arkansas Power and Light Company in its letters dated October 18, 1976 and November ll, 1976, and as approved by the NRC in its Safety Evaluation Report for the Arkansas Nuclear One, Unit 1

Spent Fuel Rack Modification dated December 17, 1976.

Dro ed Fuel Assembl Accident An evaluation of the effects of a postulated dropped fuel

assembly. accident will be performed to confirm that there

would be no effect on the spacing of fuel assemblies stored

~in.the racks.

The method of design and analysis will be identical to that submitted by Omaha Public Power District in its letter dated June 2, 1976, and as approved by the NRC

'in'its Safety Evaluation Report for the Fort Calhoun Station LVnit No.

1 Spent Fuel Rack Modification dated July 2,.1976.

1In addition to the analysis for a vertical drop onto the top

.:-of.the storage

racks, the following cases will be evaluated:

za) fuel assembly dropped inside the storage cell.

'b) fuel assembly dropped from above the racks but with cthe assumption that.the assembly rotates as it drops

~and;,impacts,a row of. storage cells.

4 Cts ~

3.7 References

,(1)

L. M. Petrie and N. F. Cross, "KENO IV:

An Improved Monte Carlo Criticality Program,"

ORNL-4938, Oak Ridge National Laboratory (November 1975).

(2)

N. M. Greene, et al, "AMPX - A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory (March 1976).

(3)

M.

W. Porath, "CCELL Users Guide," BNM/JN-86, Pacific

.:Northwest Laboratories (February 1972).

, (4)

G.

E. Whitesides and N. F. Cross, "KENO - A Multigroup Monte

~ Carlo Criticality Program,"

CTC-5, Union Carbide Corporation Nuclear Division (September 1969).

<=(5)

V.

E.= Grob, et al, "Multi-Region Reactor Lattice Studies

Results of Critical Experiments in Loose Lattices of UO

".Rods in H20," liCAP-1412, Mestinghouse Electric Corporatton (1960).

j'.(6)

K. D. Lathrop, "DTF-IV - A FORTRAN-IV Program for Solving

'rthe Multigroup Transport Equation with Anisotropic Scatter-

'ng," LA-3373, Los Alamos Scientific Laboratory (July'965).

I (j(7) >..Information obtained via personal communication with E: B.

'1:Johnson and G.

E. Mhitesides, Oak Ridge National Laboratory,

. Oak Ridge, Tennessee (September 1976).

tl(8);; S.

R. Bierman, E.

D. Clayton and B.

M. Durst, "Critical S.Separation Between Subcritical Clusters of 2.35 Mt.X U-235 iiEnriched UO Rods in Mater with Fixed Neutron, Poisons,"

PNL-P'2438; Pa'ciftc Northwest Laboratories (October 1977).

(E(9) i-'." J;-':Bathe, E. L.*Wil.son, F: E.':Peterson, SAP-IV,.'"A S5tructural, Analysis'rogram for Sta'ticand Dynamic <Response o.of. 1Linear Systems," 'Earthquake Engineering Research. Center P<Report No..73-11,,Revised, April 1976.

g-~ E"o

REC: 'CASE E G, NRC>>

ORG:

TILLINGHAST J IN 5 NI PMR REGULATORY IN~~"RNATION DISTRIBUTION DISTRIBUTION FOR Ih~'IING NATERIAL SYSTE~'RIDS 50-315 DOCDATE: 02/03/78 i)ATE RCVD: 02/06/78 DOCTYPE:

LETTER NOTARIZED:

YES COPIES RECEIVED

SUBJECT:

LTR 1

ENCL 1

FACILITY OPERATING LICENSE 58 Al"lENDi CHANGE TO TECH SPECS CONCERNING REVISION TO APPENDIX A WITH REGARD TO EXTEND THE HEAT FLUX HOT CHANNEL FACTOR LIMIT TO HIGHER REACTOR EXPOSURES.

PLANT NAME: COOK UNIT '1 REVIEWER INITIAL:'EF DISTRIBUTOR INITIAL:

DISTRIBUTION OF THIS NATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.

<DISTRIBUTION CODE A001) n FOR ACTION:

BR DAVIS44~W/7 'NCL INTERNAL:

EXTERNAL:

EG FILE++M/E L

I HANAUER4+M/ENCL E ISENHUT++W/ENCL BAER++W/ENCL GR INES++M/ENCL J.

NCGOUGH4 4 lJ/ENCL LPDR S ST.

JOSEPH'I ++M/ENCL TICs~+W/ENCL NSIC~~M/ENCL ACRS CAT B44W/ENCL NRC PDR4+W/ENCL OELD~+LTR ONLY CHECK~~M/ENCL SHAO++M/ENCL'UTLER++M/ENCL J.

COLLINS++M/ENCL DISTRIBUTION:

LTR 25 ENCL 24 SIZE:

2P+18P THE END CONTROL NBR:

780-70213

1'

i,>~ i

~o KM<@05.ÃLt(BFKf MPI'NDIANA II MICHIGAN POWER COMPANY P. O. BOX 18 BO WL IN G G RE EN STATION NEW YORK, N. Y. 10004 February 3

1978 Donald C. Cook Nuclear Plant Unit No.

1 Docket No. 50-315 DPR No.

58 Mr. Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U.

S. Nuclear Regulatory Commission Washington, D. C.

20555

.Ji@

~<I~e,~

6~

~~eg

/OR>

Dear Mr.'ase:

Changes are hereby requested to the Donald C. Cook Nuclear Plant Unit No.

1 Appendix A Technical Specifications in Sections 3;2 and 4.2 (Power Distribution Limits) and the corresponding bases.

The purpose of this change is to extend the heat flux hot channel factor limit (Fg) to higher reactor exposures than is currently provided for xn the Technical Specification.

The current Fg. limit is applicable for Cycle 2

exposure up to 10,800 MWD/MTU.

Planned operation for the bal-ance of Cycle 2 may exceed this exposure.

Also the F~ limit must be revised to be applicable to future cycles.

The proposed Technical Specification changes are included in the Attachment to this letter.

These changes are applicable to all future operation of the plant with the cur-rently designed Exxon Nuclear Fuel.

The basis for the proposed F~ limit is provided in the ECCS analysis for the plant as reported in Reference 1 and Supplements 1 and 2 to this document (References 2 and

3).

The NRC staff's evaluation of the Exxon Nuclear ECCS evaluation model applicable to the plant (Reference

4) required changes involving assignment of flow blockage uncertainties during the calculated reflood portion of the LOCA transient.

In Reference 2 further changes to the flow blockage values and uncertainties were proposed and justified for application to the plant.

In Reference 3 additional information was provided in response to questions raised by the NRC staff regarding Reference 2.

l 4

~ ',

ll L

I

o Mr. Edson G.

Case February 3

1978 This Technical Specification change request has been reviewed by both the PNSRC and the required membership of the NSDRC, in accordance with the appropriate provisions of our Technical Specification.

The results of our review indicate that, our request for this Technical Specification change will not jeopardize the health and safety of the plant workers, nor will it in any way compromise the health and safety of the public.

Very truly yours, nghast ice President JT/mab Attachments Sworn agn subscribed to before me this 8

day of February, 1978 in New York County, New York Notary Public cc:

R.

G.

R.

D.

p.

R.

R.

8 NOTARY i'U8t.tC, State, ot New bosk No. 41-4606792 Quelifted iri Queens Count erlificctu tiled in Ncw Y k C

~~r -.e siw.r iaweoa tsldscn 30, tVPp C. Callen Charnoff W. Jurgensen V. Shaller Bridgman W. Steketee Walsh J. Vollen

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO, 58 DOCKET NO. 50-315 Replace the following pages of the Appendix "A" Technical Specifications with t he enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of chanage.

~Pe es 3/4 2-1

. 3/4 2-2 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-15 3/4 2-16 3/4 2-17 3/4 2-18 3/4 2-19 3/4 3-39 3/4 3-50 B 3/4 2-1 B 3/4 2-2 B 3/4 2-5 B 3/4 2-6 Replaced by pages 3/4 2-16, 3/4 2-17)

Replaced by page 3/4 2-18)

Replaced'by page 3/4 2-19 Change page number to 3/4 2-20)

4 ~

V,

~

1Nr

4o, lp

~

~

'9 4.2 POMER DISl'RIBUTION LIMI'IS AXIAL FLUX DIFFERENCE AFD LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a +5% target band (flux difference units) about the target flux difference shown on Figure 3.2-4.

APPLICABILITY:

MODE -1 above 50K RATED l'HERMAL POMER*

ACTION:

a.

Mith the indicated AXIAL FLUX DIFFERENCE outside of the

+'5X target band about the target flux difference and with THERMAL POMER:

l.

Above 75K x T(E) of RATED THERMAL POMER, within 15 minutes:'

b)

Either restore the indicated AFD to within the target band limits, or Reduce THERMAL POWER'o less than 75K x T(E)** of RATED l

TflERtOL POWER.

2, Between 50K and 75K x T(E) of RATED THERNL POWER:

a)

POWER OPERATION may continue provided:

')

The indicated AFD has not been outside of the

+5%

target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> s, and 2)

The indicated AFD is within the limits shown on Figure 3.2-1.

Otherwise, reduce THERMAL POWER to, less than 50K of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip. Setpoints to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b)

Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the. limits of Figure 3.2-1.

'A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

• See Special Test Exception 3.10.2

• • T(E) is defined on Figure 3.2-3 and pages 3/4 2-15, 2-16 D. C.

COOK - UNIT 1

3/4 2-1

= Amendment No.

V

~v; l

• I"1 1

I V(

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued b.

c)

Surveillance testing of the APDMS may be performed pursuant to Specification 4.3.3.6.1 provided the indicated AFD is maintained with the limits of Figure 3.2-1.

A total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of operation may be accumulated with the AFD outside of the target band during.this testing without penalty deviation.,

r THERMAL POWER shall not be.increased above 75>> x T(E) of RATED THERMAL POWER unless the indicated AFD is within the + 5X target b'and and ACTION 2.a) 1), above has been satisf7ed.

c.

THERMAL POWER shall not be increased above 50K of RATED THERMAL POWER unless the indicated AFD has not been outside of the

+ 5X target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THEfML POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

l.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor'Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least. once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

D.

C.

COOK - UNIT 1

3/4 2-2 "

.Amendment No.

t If 1

~

i.'z!>

~ l ~

~

~

~H

~ ~

100 fzf f

z>>t:

l I

>> ~

"1

~>>>>J z>>

( f ttf tf!'

~ ~

ftV

.>>ll f

~

hl (fzI It)

~ l fJ>>

~ ~ tl tft

~ >>,f

>>4 zt

~ ~ tr P.ft tfxz f t' J

I (( jiij H'x z! 4 I f O

I-K Og f

2

• i

, (9

~ >>t 4>>

>>4>>

UNACCEPTABLEI OPERATION zz

~.I

~ >>~

'NACCEPTABLE t s OPERATION fz c~f l >>y>> ~ ft~-

~

!">>]>>

zz~zt

~.

~ ~ >>

~

~ 'Iz ACCEPT ABLE OPE RATION 40 Jz>>

~>>ff f ~ J

~ ~

+f-f f+

2,50)t J

+g

~'P J

~ t>>

(22,5 0)

~

4 Wl>>

l

~

~ '

~

~>> ~ J

~>>

JJZ~>>

l ~ >>>>

20

>> ~

tf >>

~t J

~ tl

~ l+t

~ l

~~

~ J ~, ~

l l ~ ~

2>>L

~ 4

~ ~ JJ It ~ ~

t

~

zJ

~ >>g

~ y

~

~

~

~

4

+Jl

~ ~ffry

~ >>

J

~ ~

~

Ifft zfJ i!-:i

~J l

.tel JJ z

~ rfJ t

qz t ~ ~

>>g

~l

~ ~

0

~

~

l >>-t f >>>>

~

I

~ fll

>>4

~

>> ~

,ttz ttf'

~ ~I'I

(('z>>

!t

>> ~

~ >>>>l I>>ft

-50

-40

.30

-20

-10 0

10 20 30 40 50 FLUX DIFFERENCE (Dl) %

FIGURE 3.2.1 AXIALFLUX DIFFERENCE LIMITSAS A FUNCTION OF RATED TIIERMALPONIER D.

C.

COOY - UNIT 1

3/4 2-4 Amendment Ho.

I

't 4

I tt

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F Z

LIMITING CONDITION FOR OPERATION 3.2.2 F (Z,k) shall be limitdd by the following relationships:

LF (E,)j F (1,4)

< ~>

[K(z)3 for P

> 0.5 F~(Z,g)

< 2 iF~(E )] [K(Z)] for P <.0.5 p

THERMAL POWER RATED THER L

POWER F (E<) is the exposure dependent F limit for assembly and is defined on Figure 3.2-3 and pages 3/4 2-15, 2-16.

EL is the maximum pellet exposure in assembly L.

K(Z) is the function obtained from Figure 3.2-2 for a,given core height location.

Fq is defined as the F~(Zg) with the smallest margin, or the greatest excess of the limit.

. APPLICABILITY:

MODE 1

ACTION:

3.>ith F~

a.

b.

exceeding its limit:

Comply with either of the following ACTIONS':

l.

Reduce THERMAL POWER at least 1% for each 1% F~ exceeds thc limit within 15 minutes and similarly redu& the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower aT Trip Setpoints have been reduced at least 1% for each 15 F-exceeds the limit.

The Overpower a7.

Trip Setpoint re3uction shal1 be performed with the reactor subcri tical.

2.

Reduce THERMAL POWER as necessary to meet the limits of Specification 3.2.6 using the APDMS.with the latest incore map and updated R.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F< is demonstrated through incore mapping to be within its limi.t.

D.

C.

COOK - UNIT 1

3/4 2-5 Amendment No.

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fx shall be evaluated to determine if F~(Zg) is within its limit by:

a.

Using the movable incore detectors to obtain a power distribu-tion map at any THERMAL POWER greater than 5/ of RATED THERMAL POWER; b.

c ~

Increasing the measured F

component of the power distribution" map by 3X to account for Qnufacturing tolerances and further increasing the value by 5X to account for measurement uncertainties.

Comparing the Fx computed (Fx

) obtained in b, above to:

C

.,1.

The Fx..limits for RATED THERMAL POWER (F

) for the xy appropriate, measured core planes given in e and f below, and 2.

The relationship:

Fx

= F[l 0.2(l-P)]

where Fx is the limit for fractional THERMAL POWER L -.

operation expressed as a function of, F and P is xy the fraction of RATED THERMAL POWER at which F

was xy measured.

t d.

Remeasuring Fx according to the following schedule:

1.

When Fx is greater than the Fx limit for the appropriate C

~

RTP measured core plane but less than the F

relationship,.

xy addit'ional power distribution maps shall be taken and Fx compared to Fx and F

C RTP L

.'y xy

')

Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20/ of RATED THERMAL POWER or greater, the THERMAL POWER at which Fx was last determined, or C

D.

C.

COOK - UNIT 1

3/4 2-6

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS Continued b)

At least once per 31 EFPD, whichever occurs first.

2.

When the F

is less than or equal to the F

limit for xy xy the appropriate measured core plane, additional power distribution maps shall be taken and F

compared to.~

C F.

and F

at least once per 31 EFPD.

xy xy e.

The Fx limits for RATED planes shall be:

1.

F 1.71 for all "D" control rods or THERMAL POWER within specific core core planes containing either bank any part length rods, and 2.

Fx

< 1.55 for all unrodded core plane's.

'TP f.,

The Fx limits of e, above, are not applicable in the fol-xy lowing core plane regions as measured in percent of cor'e height from the bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100K inclusive.

3.

Grid plane regions at 18.4 + 2X, 36.6 + 2X, 54.7 + 2/ and 72.9

+ 2/., inclusive.

4.

Core plane regions within + 2X of core height

(+ 2.88 inches) about the bank demand position of the bank "D" or part length control rods.

g.

With F exceeding F

xy xy'.

1.

The F~(Z,Z') limit shall be.reudced at least lX for=each lX F exceeds Fx

, and C

L xy 2.

The effects of F

on F<(Z,4) shall be evaluated to determine xy if F~(2,4) is within its limit.

4.2.2.3 When F~(Z,L) is measured pursuant to specification 4.10.2.2; an overall measured F~(Zg,) shall be obtained from a power di'stribution'ap and increased by 3Ã to account for manufacturing tolerances and further increased by 5/ to account for measurement uncertainty.

D.

C.

COOK - UNIT 1

3/4 2-7

k'

POWER DISTRIBUTION LIMITS AXIAL POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION 3.2.6 The axial power distribution shall be limited by the following relationship:

F. (Z)j 1.95 K Z (R

) (PL) (1 03) (1 + <

) (1 07) FP Where:

a ~

F;(Z) is the normalized axial power distribution from thimble j aX core elevation Z.

b.

PL is the fraction of RATED THERMAL POWER.

C.

d.

K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

R. - 1s the R m for thimble j. and llmltlng fuel batch m

R.m f3%

each batch) m is determined from at least n=6 in-core flux~

maps covering the full configuration of permissible rod patterns above 84Ã x,T(E) of RATED THERMAL POWER in accordance with:

Rjm n

n R...

1 jm'..

1 j[11 FSelect

~im FSel ect

~im j

MAX Q.

k jMAX by batch m

ik FMeas g.

is the relative linear power density for the k

fuel segment in the i> full core flux atap.

T(Ek)

= Fq(Ek)/Fq(O)

D.

C.

COOK - UNIT 1

3/4 2-15 Amendment No.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued where F (E<)

=

The exposure dependent F limit on the maxi~am linear heat generation rate for~fuel segment k.

F~(E) is defined in Figure 3.2-3.

F~(0)

=

The F limit at zero exposure.

L The limiting fuel rod in the core is the rod which produces the maximum value of Fselect.

The limiting batch m

is the fuel batch lm Containing limiting fuel segment k

Define T(E) as T(E)

=

Fq(EI,-)/Fq(0)

[F..(Z)j is the maximum value of the normalized axial distri-buHon al 5levation Z from thimble j in map i which had a measured peaking factor without uncertainties or densification allowance of FMeas e.

a. i is the standard deviation associated with thimble j~ and limiting fll batch m

, expressed as a fraction or percentage of R..

a.

for each batch m is derived from n flux maps froto the relatiohPhip

bhTow, or 0.02, (2X) whichever is greater.

n

=

L~

(R.

- R" ) j

/R.

1

~

2 1/2 jm n

1

~

1 gm ijm jm f.

The factor 1.07 is comprised of 1.02 and 1.05 to account for the axial power distribution instrumentation accuracy and the measure-ment uncertainty associated with F~ using the movable detector system respectively.

The factor 1.03 is the engineering uncertainty factor.

g.

Fp is an uncertainty factor to account for the reduction in the Fq(E) curve due to an accumulation of exposure prior to the next flux map.

This correction is only required when T(E) for the limiting fuel segment is less than 1.0.

F

= 1.0 for T(E)

= 1.0

~

P F

= 1.01 for T(E) s 1.0 E

is the peak pellet exposure in the limiting fuel rod.

D.

C.

COOK - UNIT 1

3/4 2-16 Amendment No.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued APPLICABILITY:

MODE 1 above 84% x T(E) of RATED THERMAL POWER

.¹ ACTION:

a.

With a Fi(Z) factor exceeding t F.(Z)]

by < 4 percent, reduce TWERIIRL POWER one percent for every percent by which the F>(Z) factor exceeds its limit within 15 minutes and within the next two hours either reduce the F.(Z) factor to within its limit or reduce THERNL POWER to 844 x T(E) or less of RATED THERMAL POWER.

b.

With a F;(Z) factor exceeding 1 F>(Z)]q by > 4 percent, reduce THERNL POWER TO 84% x T(E) or less oV RATED THERMAL POWER within 15 minutes.

¹ The APDMS may be out of service:

1) when incore.maps are being taken as part of the Augmented Startup Test Program or 2) when surveillance for determining power distribution maps is bei ng performed.

D.

C.

COOK - UNIT 1

3/4 2-17 Amendment No.

'OWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS

'.2.6.1 F, (Z) shall be determined to be within its limit by:

a.

Either using the APDMS to monitor the thimbles required per Specification 3.3.3.6 at the following f) equencies.

l.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2.

Imediately and at intervals of 10, 30, 60, 90, 120, 240 and 480 minutes 'following:

a)

Increasing the THERMAL POWER above 84K, x T(E) of RAl'ED THERMAL POWER, or b)

Movement of control bank "0" more than an accumulated total of 5 steps in any one direction.

b.

Or using the movable incore detectors at the following fre-quencies when the APDMS is inoperable:

l.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2.

At intervals of 30, 60, 90,

120, 240 and 480 minutes following:

a)

Increasing the THER!IAL POWER above 84K x t',(E), of RATED THERMAL POWER, or b)

Movement of control bank "D" mo)';. than an accumulated total of 5 steps in any one di~ection.

4.2.6.2 h'hen the movable incore detectors are used to monitor F (2), at least 2 thimbles shall be monitored and an F.(Z) accuracy equiva ent to that obtained from the APDMS shall be maintained.

0.

C.

COOK - UNIT 1

3/4 2-18 A(Pensdl))en t No.

't ~

c ~

2.0

.I.;;.,

I"(20~1 9S) 90) 27 1.9 1.8 (40;. ;79 07-14E L:

Q:

1.7 I

20, I 00.).

1.0 27, 974 T(P)

"'I-.

(40, 92eIt.

0.9 I

18E 2..(E;:.

87 T E

"~

0.0

~

~ l 5.7 D.

C.

COOK - UNIT 1

3/4 2-19 AMENDMENT NO.

4~

8 12 16 20 24 28 '2 36 40 44 48 PEAK PELLET EXPOSURE IN MWD/KG FIGURE 3.2-3 EXPOSURE DEPENDENT FQ LIMIT, FQ(E),

AND NORMALIZED LIMIT T(E)

AS A FUNCTION OF PEAK PELLET BURNUP

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION'.3.3.2 The movable incore detection system shall be OPERABLE with:

a.

At least 75/ of the detector thimbles, b.

A minimum of 2 detector thimbles per core.quadrant, and c.

Sufficient movable detectors,

drive, and readout equipment to map these thimbles.

APPLICABILITY: When the movable incore detection system is used for:

a.

Recalibration of the axial flux difference detection, system, b.

Monitoring the QUADRANT POWER TILT RATIO, or c.

Measurement of F>H and FQ(ZQ )

N ACTION:

With the movable incore detection system inoperable',

do not use 'the system for. the above applicable monitoring or calibration functions.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3:2 The movable incore detection system shal,l be demonstrated OPERABLE by normalizing each detector output to be used during its use when required for:

a.

Recalibration of the excore axial:flux difference detection

system, or b.

Monitoring the QUADRANT POWER TILT RATIO, or c.

Measurement of F H and FQ(Z,4)

N D.

C.

COOK-UNIT 1

3/4 3-39

E A

INSTRUMENTATION 0

SURVEILLANCE RE UIREMENTS Continued) a.

If.h b

1 1 f~ i g

h 2.

gh 1 jm m

jm jm map shall be completed to verify the new R... If the second map shows the first to be in error, the first Bp shall be dis-regarded.

If the second map confirms the new R.m-, four more 'maps (including rodded configurations allowed by the~)nsertion limits) will be completed so that a new R-. and a.. can be defined"from the

. six new maps.

Jm jm

.3.3.6.2 The APDMS shall be demonstrated OPERABLE:

a.

By performance of a CHANNEL FUNCTIONAL TEST within 7 days prior to its use and at least once per 31 days thereafter when used for monitoring F.(Z).

b.

At least once per 18 months, during shutdown or below 5/ of

'ATED THERMAL PO'l<ER, by performance of a CHANNEL CALIBRATION.

D.'.

COOK-UNIT 1

3/4 3-50

3/4;2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integ-rity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core 1.30 during normal operation and in short term transients, and (b) lrmiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are meet and the ECCS acceptance criteria limit of 2200"F is not exceeded.

The definitions of hot channel factors as used in these specifi-cati,ons are as follows:

FH aH Nuclear Enthalpy Rise Hot Channel

Factor, is defined as the ratio of the integral of linear power along the rod with the hi hest inte rated ower to the aver'age rod power.'

g p

3/4.2.1 AXIAL FLUX DIFFERENCE AFD

'<'(2~k)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation 2

divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods.

Target flux difference is determined at equilibrium xenon conditions with the part length control rods withdrawn from the c'ore.'he full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value o.f the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at 'RATED THERtlAL POWER for the associated core burnup'conditions.

Target flux differences for other THERt1AI POWER levels are obtained by multiplying the RATED THERt1AL POWER value by the appropriate fractional'HER11AL POWER level.

The periodic updating of the target'lux dil fereiice value is necessary to

> oflect core bur>>imp cu>>iiilr ~ iLions.

D.

C.

COOK-UNIT 1

B 3/4 2-1

~

~

)

I

POWER DISTRIBUTION LIMITS BASES Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the

+ 5X target band about the target flux difference, during rapid plant THERMAL POWER reductions; control rod motion will cause the AFD to deviate outside of the target band at re-duced THERMAL POWER levels.

This deviatinn will nnt affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERNL POWER (with the AFO within the target band) provided the time duration of the devi-ation is limited.

Accordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit.cumu-lative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50K and 75K x T(E) of RATED THERNL POWER.

For THERNL POMER levels between 15K, and 50K of RATED THERMAL POWER, devia-tions the AFO outside of the target band are less significant.

- The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message imnediately if the AFO for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERNI POWER is greater than 75K x T(E) of RATED THERNL POWER. During operation at THERMAL POWER levels between

'50% and 75%

x -7(F) and between 15~ and 50K RAl'EO THERMAL POWER, the computer outputs a<<alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

The upper bound 'mit (84'A x.T(E.)'f RATED THERMAL POWER) on AXIAL FLUX DIFFERENCE assures that the,Fqp,4} envelope of 2.32 times;;i=,:!

K(Z) x T(E) is not to exceed during either normal ojeration or. in.

the event of xenon redistribution following power changes.

The lower bound limit (50K of RATED THERMAl POWER) is based on the fact that at THERMAL POWER levels below 50K of RATED THERMAL POWER, the average linear heat generation rate is half of its nominal operating value and below.hat value, perturbations in localized flux distributions cannot affect the results of ECCS or ONBR ana'lyses in a manner which would adversely affect the health and safety of the public.

Figure 8 3/4 2-1 shows a typical monthly target band near the beginning of core life.

O.

C.

COOK-UNIT 1

B 3/4 2-2 Amendment No.

POMF') DISTRIBUi'ION Litili' OASES a..abnormal perturuations in ]he radial power shape, such as from rod misalignment, effect.F H more directly than F, b.

although rod movement has a direct influence upon limiting Fg to wHithin its 'limit, such control is not readily available to lsmit F H, and errors in pr diction tor control power shape detected during startup physics tests can be compensated for in F by rystri-ting axial flux distributions.

This compensation for F" is less readily available.

aH A burnup dependent Fn is specified as a result of the ECCS evaluation in accordance Pith 10 CFR Part 50 Appendix K and to meet the acceptance criteria of 10 CFR 50.46.

The basis for this dependence is given in document XN-76-51, Supplement 1

and 2.

3/4.2.4 QUADRANT POMER TILT RATIO The quadrant po~e~ tilt ratio limit assures that the radial power distribution ~atisfies

~he design values'used in the power capability ana'iqsis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

tiie limii of 1.02 ~t >,hich corrective action is required provides DNB rui! linear iieat. generation rate protection with x-y plane power tilts.

A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in l is depleted.

lie limit of 1.02 was selected to.provide an allowance for %he uncertainty a:.'ciated with the indicated power tilt.

(t.

'.,wo hour t'iii~ aliowance for operation with a tilt condition greater

'.l:i 1.02 but less tnar, 1.09 is provided to allow identification and cor-r.i.inn of a dt.opped oi misaligned rod.

In the event such action does not

o. << t i:iIt: t.ilt, th. -iai sjin fci uncertainty on F

is reinstated by r>> l..ing tii. power by 3 i,erc~nt for each percent Igf tilt in excess of 1.0.

D. I.. U)UK-i>'<ll I

B 3/4 2-5 Amendment No.

a

POQN DISTRIBUTION LIMITS

ASES 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that'each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and act,ident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytical/y demonstrated adequate to maintain a minimum DNBR of 1.30 throughout:

each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru in-strument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month periodic measurement of th.

RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a

12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

3/4.2.6 AXIAL POWER DISTRIBUTION The limit on axial power distribution ensures that F

> ill be controlled and monitored on a more exact basis through us3 or the APONs when operating above 84/ x T(E) of RATED 1HEPf1AL POWER.

This additional limitation on F is necessary in order to provide assurance that peak clad.temperatures will remain below the ECCS.acceptance criteria limit of 2200'F in the event of a LOCA.

The unit may operate with fuel assemblies supplied by the Exxon Nuclear Company and by Westinghouse Electric Corporation.

The specified limit 'for F~ represents the Exxon Nuclear supplied fuel which has the more re5trictive power peaking limit.

D. 'C.

COOK-UNIT 1

B 3/4 2-6 Rmc.'ndment No,