ML19312D366

Amend 61 to License DPR-46,allowing Count Rate in Source Range Monitor Channels to Drop Below Three Counts Per Second When Entire Reactor Core Is Being Removed or Replaced

ML19312D366
Person / Time
Site:	Cooper
Issue date:	02/29/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19312D364	List: ML19312D363 ML19312D366 ML19312D368 ML19312D369
References
NUDOCS 8003240275
Download: ML19312D366 (19)

Text

.- _. -_

v ; p aer g

.,fi f,,

UNITED STATES NUCLEAR REGULATORY COMMISSION g

g n

.a wAsmWGTON, D. C. 20656

.....)

?

NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AENDENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-46 1.

The Nuclear Regulatory Cbmission (the Commission) has found that:

A.

The application for amendment by Nebraska Public Power District (the licensee) date'd January 14, 1980, complies with the standards and requirements of the Attmic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application,

- the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of.the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The-issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

i 2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. OPR-46 is hereby amended to. read.as follows:

-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.~ 61, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(

y-8003240T75 i

i

. t

- -- 3.

This license-amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COPMISSION Thomas A? Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

' Changes.to'the-Technical Specifications.

' Dated:

February 29, 1980-l' 1

Y t

I i

s

~

ATTACMENT TO LICENSE AMENDMENT NO. 61 FACILITY CPERATING LICENSE NO. DPR-46

' DOCKET NO. 50-298 Replacethefollowingpagesof'theAppendix"A"TechnicalSpecificationswith the enclosed pages. The revised pages.are identified by A,nendment number and contain vertical lines indicating the area of change.

Remove Insert

~5 5

5A 5A

51.

61 62a L-86 86 lL 94a' 94a 203 -205 203 - 205a 206

.209a'.

206 - 209b f.;

1 w

--.n. ~.,,

3.

All cetom: tic centoin::nt toolction valves cro cpercbla or da-activeted in the isolated position, 4.

All blind flanges and manways are. closed.

'Q.

Rated rever - Rated power refers to operation at a reactor power of.2381 megawatts thernal.

This is also termed 100" power r-i is the maxinun power Icvel authorized-by'the operating 11ecnse.

Rated sicam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the valtes of these parameters when the reactor is at rated power.

Design power, the power to which the safety analysis applies, is 105% of rated power,~which corresponds to 2486 megawatts thermal.

R.

Reactor Power Operatien - Reactor power operation is any operation with the mode switch in the "Staetup/ Hot Standby" or "Run" position with the reacter critical and above 1% rated power.

S.

Reactor Vessel Pressure - Unicss other ise indicated, reactor vessel pressures listed uin the Technical Specifications are those measured by the reactor vessel steam space detcetors.

T.

Refueling Outace - Refueling outsge is the period of time between the i

shutdown of the unit prior to a refueling and the start,up of the plant after that refueling.

U.

Safetv Linies - The safety limits are limits within which the reasonable maintenanca of the fuel cladding integrity and the reactor coolant systen integrity are assured.

Violation of such a. limit is cause for unit shut-down and review by tne Nuclear Regulatory Commission before resumption of unit operation.. Operation beyond such a limit may not in itself result in serious consequences buc it indicates an operational deficiency subject to regulatory review.

V.

Secondarv Contaiteent Inteeriev - Secondary containcent integrity = cans tha t the reactor bui] ding $s 4.ntact and the following conditions are cet:

1.

At least one door in each access opening is closed.

2.

The standby gas treatment system is operable.

3.

All automatic ventilation system isolation valves are operabic or secured in the isolated position.

W. ' Shutdown - The reactor is in a shutdown condition when the mode switch is in the " Shutdown" or." Refuel" position.

1.

Hot Shutdbwn means conditions as above with reactor coolant temperature greater than 212*F.

2. -Cold Shutdown ceans conditions as-above with reactor coolant temperature equal to or less than 212*F and the reactor vessel vented.

X.

Spiral Reload - Pertains to the spiral reloading of the core with fuel, I~

at least 50% of which has previous 1) necumulated a minimum exposure.of 1000 MWD /T.

l;

~

Amendment No.[ 61 - _ _.

y.

Surveillance Frequency -

Surveillance requirements shall be applicable during.the operational conditions associated with individual ILO's unless otherwise stated in an individual Surveillance Requirement.

Each Surveillance Requirement shall be performed within the specified time 4nterval with:

a.

A maximum allowable extension not to exceed 25% of the surveillance interval.

b.

A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3.25 tises the specified' interval.

Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with operability requirements for an'LCO unless otherwise required by the specification.

Z. Surveillence Interval - The surveillance interval is the calendar time between surveillance tests, chacks, calibrations and e::aminations to be perforr.ed upon an instrument or criponent when it is required to be operable. These tests may be waived when the instrunent, cc::ponent or systen is not required to be operable, but the instru.ent, co=ponent or system shall be tested prior to being declared operable or as practicable following its return to service.

I I

_ sA _

AmendmentNo.g61

~

COOPER NbutEAR~ STATION TABLE 3.2.C CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION-Mini =un Number Of Function

' Trip Level Setting' Operable Instrument -

Channels / Trip System (5)-

APRM Upscale-(Flow l Bias).

- ji (0.66W + 42%) 'F10P (2) 2(1)-

APRM Upscale (Startup)

< 12%

, G12D, 12(1)

}

APRM Downscale -(9) 1 2.5%

2 (1) ~

APRM Inoperative (10b) 2(1)

RBM Upscale,(Flow Bias) ji (0.66W + 39%) (2)

,1 RBM Downscale (9)-

1 2.5%

1 R]pt Inoperative (10c) 1 lIRM Upscale (8) ji 108/125 of Full Scale 3(1)

IRM Downscale (3)(8) 1 2.5%

3(1) i$-

IRM Detector Not Full In (8) 3(1)

IRM Inoperative (8)

(10a) 3(1)

SRM Upscale-(8)

L 1 x 105 Counts /Second

~1(1)(6)

SRM Detector Not. Fv"1 In (4)(8)

(1 100 cps) 1(1)(6)

SRM Inoperativa (8)

(10s)-

1(1)(6)

Flow; Bias Comparator ji 10% Difference In'Recire. Flows 1

Flow' Bias Upscale /Inop.

j[ 110% Recire. Flow I

.SRM Downscale (8)(7) 1 3' Counts /Second (11)-

.1(1)(6)

Amendment No. 55, 61

4

- -....... _ a._.

- n-cs n ur i

mu'

, + -

W-EW i;c ; g.;

.I.

..e

. n aa 4

+w: m.v.

x x. : p7.

+

~- <

',?.^

.?:.

17,:

+

A 2;-},_'

'A

~m

. l 2-.

/

a w

l

._1/?,

s,D'E

~^;',,,..,,

4 7._

,c Ts-y s..

r k

-T f v

f -

.r a'

t 2?

h^'

_ m, x,..

,e-..

a

- ~

4

.s 4 *~

4

,c

",i s

~.

p-2-

.f t

E 116 lDuring'._ spiral; unloading /reloa' ding, the SRM count rate will be.below-y'3 M :

a.-

i SeeSpbcificationJ3.10.B.-

'l j

13; cps for some period of time.

G I/.

4 p.

J

~

a a,

1. [

r-t s

o

$.^'

1 s

r,

.m o

et w.

s i E.

A s

J

^,

-~

1 41

g

?).

j e

4 m,,

t.

s

,o z.a. -

t-m'.

4

. k.'

i;.

e 4_

,,4 a

1 d ~,

r

.'..E'

}

q g.

4h 7 s

~

~

1a

~

1P w.

e,a 5

i s

f

~

_e S < /

g g,,

.t

~'

~

v s

1..a.

x

~

.t

.n.-

3

,+..

9,. 2

+5'I_

+gr-ei r

I a

g N'

i u

~.. -

r k-.

w n,e s

~

~

~

}:

..e.,e,,_

? 9:

~3 y-g'*

~

.}

k[

'f r.

O

/

A r

~'

y

' f.

\\

j:

r

'.,.g.;,,#

X' ff',.

(

l...

i r.

+

1

3

_g,

~.

-f.

nw -

a:.

s 3

.:e =. W.i ; > > -

' 1, 7 n Amendment No k, s_61;e

- 62a:--

p,, h 1J

'.w

)

\\

v' "

3 g

.1, 7 ).y j

A-

-E 'g-

'./^-['.

.W'

,g--

u n_,

~

p

+

.4

'4j4-6 4,

f*h 1, h, ?

i ')

M',

y' r

_ 4-1 c.

4 t

.r 7

' ;;y y -

.,g.,

L 95d, itK1..

y

(

-.;,..,.?

a n.

3i s ;

+

Q' e

e' e,$,b K.:( y 'i_]

z.

n 5 ~,: -

o y

q U

3,2 ' BASES - (cont'd) prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

The IRM rod block ' function provides local as well as gross core protection.

The scaling arrangement is such that tri.p setting is less thar. a factor of 10 above the indicated level.

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive.enough.

In either case the instrument will not respond to changes in control rod notion and thus, control rod motion is prevented. The downscale trips are set at 2.5 indicated on scale.

The flow comparator and scram discharge volum high level components have only one logic channel and are not required for safety.

• lhe refueling interlocks also operate one logic channel, and are required for a safety only when the mode switch is in the refueling position.

The effective emergency core cooling for small pipe breaks, the HPCI system, must function since reactor pressure does not decrease rapid enough to allow either core spray of LPCI ta oncrate in time. The automatic pressure relief funct ion is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met.

The specification preserves the effectiveness of the system during periods of maintenance,~ testing, or calibration, and also minimizes the risk of inadver-tent operation; i.e., only one instrument channel out of service.

Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line.

Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip. There is a fifteen minute delay accounte-for by the 30-minute holdup time of the off-gas before it is reached to thc, tack.

Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip. The trip setting of 1.0 ci/sec (prior to 30 min, delay) provides.an improved capability to detect fuel. pin cladding failures to allow prevention of serious degradation of fuci pin cladding j

i integrity which might result from plant operation with a misoriented or misloaded fuel assembly. This limit is more restrictive than 0.39 ci/sec nobic gas release -

rate at the air. ejectors (after 30 min, delay) which was used as the source term for an accident analysis of the augmented off-gas system. Using the.39 ci/sec source term, the maximum off-site total body dose would be less,than the 5 rem limit.

Two radiationf monitors are provided which initiate.the Reactor Building Isolation

~

fur.ction and operation of the standby gas treatment system. The trip is actuated t -

by one hi-hi or two downscale indications.

Amendnent No,[ 61 ' -

a

.+m

LIMITING CON _DITION FOR OPERATION SURVEILLANCE REQUIREMENT I

- 3.3.

(Cont'd) 4.3 (Cont'd)

L;,

.B..

Control Rods B. Control Rods

-- 1. - Each control rod shall be coupled to

1. The coupling integrity shall be 4

its drive.or completely. inserted and verified for each withdrawn control the control rod directional control rod as follows:

valves disarmed electrically.. This.

requirement does not apply in the refuel-

a. When a rod is withdrawn the first condition when the reactor is vented.

time nfter each refueling outage or

-Two or more control rod drives may be after maintenance, observe discern-

~

removed as long'as Specification 3.10.A.5 ible response of the nuclear instru-nor.3.10.A.6 is act.

mentation and rod position. indication.

However, for initial rods when response is not discernible, subsequent exer-cising of these rods after the reactor is above 30% power shall be performed to verify instrumentation response.

b.

When the rod is fully withdrawn the first time after each refueling outage or after maintenance, observe that the drive does not go to the over-travel position.

1 4

1 w

TLAmendment No. 32, 61'

--94a-m

_.a

' LIMITING CONDITIONS ~FOR OPERATION SURVEILLANCE REQUIREMENTS

3.10 CORE ALTERATIONS 4.10 CORE ALTERATIONS

-Adplicability Applicability

. Applies to the fuel handling and.

Applies to the. periodic testing of ccore reactivity limitations.

those interlocks and instrumentation used during refueling and core alterations.

Objective ~

Objective

.To ensure th'at core reactivity is To verify the operability of

~

within the capability of the control instrumentation and interlocks used rods and to prevent criticality in refueling and core alter.itions.

during refueling.-

Specification Specification A.:

Refueling Interlocks A.

Refueling Interlocks

~1.

The reactor mode switch shall be 1.

Prior to any fuel handling with the locked.in the " Refuel" position head off the reactor vessel, the during cote alterations and the

. refueling interlocks shall be refueling interlocks shall be functionally tested. They shall be operable except as specified in tested at weekly intervals thereafter 3.10.A.5 and 3.10.A.6 below.

until no longer required.

They shall also be tested following any repair work associated with the interlocks.

2.

Tuel'shall not be loaded into the' 2.

Prior to performing controt rod or reactor core unless all control control rod drive maintenance on a

' rods are fully inserted or unless control cell without removing fuel the spiral unlohd/ reload' technique assemblies, a shutdown margin is used.,

demonstation shall have been performed for the existing core configuration as per Specification 4.3.A.1;' additionally, the. refuel-

.ing interlocks shall be operabib for all other control rods.

t

-Amendment No. 61

-203-

9 1.IMITING CONDITIONS FOR ~0PERATION SURVEILLANCE REQUIREMENTS 3.10.A (Cont'd) 4.10.A (Cont'd) 3.

'The fuel grapple hoist 1s24 switch 3.

Whenever the reactor.is in the refuel

~

~

shall be set at < 650 lbs.

mode and rod block interlocks are being bypassed for core unloading, one licensed operator and one member of the reactor engineering staff will verify that all fuel has been removed before the corresponding control rod is withdrawn.

4..

If the f rame-mounted auxiliary hoist,

4. 'Following the withdrawal and bypassing the monotait-mounted auxiliary hoist, of a control rod, two licensed operators or che service platform hoist is to will verify that the interlock bypassed

~

be used for handling fuel with the is on the correct control rod.

head _off the reactor vessel, the load limit switch on the hoist to be used shall be set at < 400 lbs.

5.

A maximum of two nonadjacent control 5.

Prior to loading fuel in a control cell

-rods may be withdrawn from the core (using the spiral reload technique),

for the purpose of performing control the control room operator and a licensed rod and/or control rod drive maintenance, operator and a member of'the reactor provided the following conditions are engineering staff on the refueling floor satisfied:

shall verify that the control rod is inserted in the cell to be loaded.

c.

The reactor mode switch shall be locked in the " refuel" position. The refueling interlock which prevents more than one control rod from being withdrawn may-be bypassed for one of the control rods on which maintenance is being

performed. A11'other refueling interlocks _shall be operable.

b.-

A sufficient number of control rods shall be operable so that the core can be made suberitical vith the strongest operable control rod fully _

-withdrawn and all;other operable control rods _ t ully-Inserted, or all directional control valves for remaining control rods.shallibe disarmed electrically

md 'suf1icient margin to criticality shall be demonstrated..

~ lf maint enance is L o be perfomed on t

c.c t wo cont rol rod drives, they must be separated by more than two control jella in any direction.

d.

' An appropriat e number. of - SRM's are t

l

.available as defined in specification'

~

'l.10. B.

AmendAent No.. 61

-204-x

-o

~,f[ ' i.

^~

LIMITING'C'ONDITIONS FO'R OPERATION SURVEILLANCE: REQUIREMENTS 13.10.Aji(Cont'd)'

4 10 (Cont'd)

Y6k.!Anynumberof.controirodsmay'be withdrawn or remov'ed'from the.

creactor core providing the following conditions are satisfied:

9c,--. cThe. reactor mode'.. switch is locked in

~

.the;" refuel" position..The. refueling-

"intericak which prevents more than one control ~ rod-from being withdrawn may.

be bypassed on a withdrawn control rod after the-fuel assemblies in the cell-containingL(controlled-by) 'that control rod have been removed from the reactor

-core..All other refueling. interlocks

'shall be-operable.

-B.

Core Monitoring B..

Core Monitoring During' core alterations two SRH's Prior to making any alterations to shall be' operable,'one in the core the core, the SRM's shall be quadrant where fuel or control rods functionally tested and_ checked for are being moved and one in an'ad-neutron response. Thereafter, while jacent quadrant. For an SRM to be required to be operable, the SRM's considered operable, the following will be checked daily for response Econditions-shall be satisfied:

(or every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until 3 cps is attained if the spiral reload 1.

"The-SRM shall be inserted to.the normal technique is being used).

. operating level.

(Use of spe'cial move-

~ ble; dunking type detectors during a

initial fuel loading and major core

~J alterations in place of normal detectors

.is permissible as-long as the detector

'is conne'cted to.the normal'SRM circuit.).

f J 2.'.

Operable SRM's shall have'a minimum.of 3 cps.except as'apec.ified in 3.and-4 below.

.l.

Prtot to sptral unloading, the'SRM's

shall have.3n'initfal

count rate of

~

~

l eps.

Durlug spiral unloading, the count-rate"on the SRM's may-drop

below;.3 eps.

J'.

Amendment:No.; l61-

~LIMITINC: CONDITIONS FOR' OPERATION-SURVEILLANCE REQUIREMENTS 3.10.B -(Cont'd)i 4.10 (Cont'd) ~

l 4..-

During spiral reload,~3RM operability will be verified by using a portable-1 external 1 source' every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the1 required amount of fuel is. loaded to maintain 3 cps. "As an alternative

-to thejabove, two fuel assemblies will-be-

ded in different cells containing cor al-blades aro'ind each.SRM to.obtain

'the required 3' cps. Until these two assemblies have'been loaded, the 3 cps

.requireraent' is not necessary.

n

-C.

' Spent Fuel Pool Water Level C.

Spent Fuel Pool Water Level Whenever irradiated fuel is stored When irradiated fuel is stored in the in the-spent fuel pool, the pool spent fuel pool, the water level shall water level.shall be maintained at be recorded daily.

sor above 8 ' above the top of the

~

~

fuel.

D.-

Time' Limitation Irradiated fuel shall not be handled in or above the reactor. prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown.

?

I'-

J Amendment No.

61 -

-205a--

N g

1

)

t LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

[3.10 (Cont'd)L 4.10~ (Cont'd)'

E.~,

Spent' Fuel Cisk Handling E.

Spent Fuel Cask ~ Handling 1

1 Fuel: cask handling'above the 931' 1.

Prior:to fuel cask handling operations

~

level of:the Reactor Building will be

.the redundant. crane including the rope, done in the RESTRICTED MODE only'except.

hooks, slings, shackles-and other as specified in 3.10.E.2 operating mechanisms will be inspected.

The rope will be-replaced if'any of the following conditions exist:

a.

Twelve (12) randomly. distributed broken wires in one lay'or four'(4) broken wires in'one strand of one rope lay.

b.

Wear of one-third the original diameter of outside individual wire.

c.

Kinking, crushing, or any other damage resulting'in distortion of_the rope.

d.

Evidence of any type of heat. damage, e.

Reductions from nominal diameter of more than 1/16 inch for a rope diameter from 7/8" to ik" inclusive.

~ 2. -.' Fuel cask handling in other than the 2.

Prior to operations in the RESTRICTED RESTRICTED MODE will be permitted in MODE:

' emergency or. equipment failure ~

' situations only to the extent necessary a.

the controlled area limit switches to get the cask to the closest acceptable will be tested; stable location.-

b.

the "two-block" limit switches will be tested; c.

the " inching hoist" controls will be tested.

- 3.

. Operation.with a failed' controlled area 3.

The empty' spent fuel cask will be limit: switch is permissible for~48 hours lifted-free of all support by a

~-providing an operatorc is on -the refueling

. maximum of'l foot and left hanging for Lfloor to assure the crane is operated 5 minutes prior to any series'of fuel within the restricted zone painted on-cask' handling' operations.

.the floor.

~41.-

Spent 1 uel casks weighing in excess -of -

f L

140,000.lbs. shall not be handled.

J Amendment-No. 35,}. ' -206 -

n 3.10 BASES-A..

Refueling Interlocks

'The refueling _inte' locks'are designed to back up procedural core r

~ reactivity controls'during refueling operations.. The interlocks prevent an inadvertent cricicality during refueling operations (when

~

the reactivity potential of the. core is being altered) by restricting

~

~

the movement of' control rods and:the~ operation of refueling equipment.

The interlocks include circuitry which senses _the condition of the refueling equipment and the control rods. Depending on the sensed condition, interlocks are actuated which prevent the movement of the refueling equipment or_ withdrawal of-control rods (rod block). Cir-cuitry is provided which senses the following conditions:

1.

All rods-inserted.

(2.

Refueling platform positioned near or over the core.

3.

. Refueling platform hoists are fuel-loaded (fuel grapple, frame-mounted hoist, monorail-mounted hoist).

4.

, Fuel grapple.not full up.

5.

' Service platform hoist fuel-loaded.

6.

One rod withdrawn.

When the mode switch is in the " Refuel" position, interlocks prevent

~ he refueling platform from being' moved over the core if a control rod t

is withdrawn-and-fuel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. When the mode' switch is in the refuel

. position, only_one control rod can be withdrawn. The refueling inter-locks,Lin combination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality. The nuclear-

. characteristics of the core assure that the reactor is suberitical even

.when-the highest worth control rod is fully withdrawn. The combination of2 refueling interlocks for control rods and the' refueling platform provide redundant methods of preventing inadvertent criticality even after procedural violations. The interlocks int hoists provide yet

~

another method of_nvoiding' inadvertent criticality.

Fuel handling is normally conducted with the fuel' grapple hoist. The total 1oad on this hoist-when the-interlock is required consists of the welghtlof the fuel grapple and the fuel assembly. This total is--

'approximately:980_lbs., in comparison to the load-trip setting of 650 lbs. - Provisions'have also been made to' allow fuel handling with either of-the three. auxiliary hoists and still maintain the refueling inter-locks.~ The 400-lb. load-trip setting on these hoists is adequate to trip the ' interlock when one of the more than 600 ~1b. fuel bundles is ibeing handled.

? mendment. No. c 61 j

A

-207-

N 3.10' HASES (Con't'd)

'During certain.pariods, it is dest'rable to perform maintenance on two control' rods and/or control rod. drives at thcLsame time. -The maintenance p-Lis performed withLthe mode switch in the " refuel" position to provide the refueling interlocks normally_available during refueling operations.

.In order 4 o withdraw a second control rod after withdrawal ofithe first t

rod, itHis necessary to bypass the refueling interlock on the first control' rod which prevents more than one control rod from'being with-drawn at the same time. -The-requirement that an adequate shutdown 1 margin be demonstrated or that.all remaining control rods have their directional control valves electrically disarmed ensures that Inadvertent criticality cannot occur during this maintenance.

The adequacy of'the shutdown. margin is= verified by demonstrating that the core is shut down by a margin of 0.38 percent ok with'the strongest operable control rod fully withdrawn, or that at least 0.38 percent.6k shutdown margin is

-available-if the remaining control rods have had their directional control valves disarmed. Disarming the directional control valves does not inhibit control rod scram capability.

Specification 3.10.A.6 allows unloading of a significant portion of the

-reactor core. This operation is performed with the mode switch in ti.e

" refuel" position to provide the refueling interlocks normally available

-during refueling operations.

In order to withdraw more than one control rod -it-is necessary to bypass the refueling -interlock on each withdrawn control rod which prevents'more than one control rod " rom being withdrawn at a time. The requirement that the fuel assemblics he cell controlled by the control rod be removed from the reactor core be. se the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality.

Prior to removal of the last two diagonal fuel assemblies, a double blade guide shall be inserted to

. properly support the control rod and fuel assemblics. After removal of the:last two fuel assemblies and withdrawal of the control rod, the double blade ~ guide _may be removed.

Each control rod provides primary reactivity control for the fuel assemblies in the call associated with that control rod. Thus, removal of an entire _cc11 (fuel assemblies _plus control rod) results in a lower

-reactivity potential of the core. The requirements for SRM operability

-during these_ core alterations assure sufficient core monitoring.

To mintmtre the possibility of loading fuel into a cell containing no control-rod, when' refueling interlock input signals are bypassed, it.is

~

l required that_the control room operator and-a licensed operator and a member of.the reactor engineering staff on the refueling floor verify

~ that the' control ~ rod is inserted in the cell to be loaded.

Prior to insertion of the control rod, it shall be verified that a double blade the fgulde;was-placed in the cell to be loaded to properly support cont rol rod and fuel assemblies.

-iE'-

f L Amendment No. 61

-208-

F..

-)-

~ ', ~

4, 3.10fBASES '(Cont'd)[

B.

Core' Monitoring-

'The SRM's are provided'to monitor the core during periods'of station

,. shut _down~and'to guide the operator during refueling operations and 3 -

' station startup.

Requiring two operable SRM's in or' adjacent to any

~

core quadrant.where fuel or control rods are being' moved assures

adequate monitoring of that quadrant during such alterations.

The

~

requirement of >3 counts per second provides assurance that neutron

' flux is being monitored and insures that startup-is' conducted only if the: source range flux level is above the minimum assumed in the control rod drop accident.

A spiral unloading pattern is_one by which the fuel in the outermost cells (four fuel bundles surrounding a control blade) is removed first.

. Unloading continues by removing the remaining outermost fuel cell by

-cell.

The center cell will be the last removed.

Spiral reloading is the. reverse of unloading.

Spiral unloading and reloading will preclude the creation of' flux traps (moderator filled cavities surrounded on all sides.by fuel).

During. spiral unloading, the SRM's shall'have an initial' count rate of

>3 cps with all' rods _ fully inserted. The count rate will diminish during fuel removal.' After all the fuel is removed from a cell, the control rod may be withdrawn in that cell. After the control rod is withdrawn, the refueling interlock will be bypassed on that control rod.

Following the withdrawal and bypassing of che control rod, two

. licensed' operators will verify that-the interlock bypassed is on the correct control rod. -Once the_ control Iod is withdrawn, it will be valved'out of' service..The refueling interlocks will prevent the O

withdrawal of another control rod unless the control rod just withdrawn

~from the unloaded cell is' bypassed.

Under this'special condition of complete spiral core unloading, it is

. expected that the count rate of the SRM's will drop below 3 cps before-all'of the fuel is unloaded. Since there will be no reactivity _ additions,

-a lower number of' counts will not present a hazard. When all of the fuel'.has been: removed to-the spent fuel storage pool, the SRM's will-no longer be required.

Requiring the SRM's to'be operational prior to fuel removal assures that'the SRM's are operable and can be relied on

'even when ~ the count i ate may ' go below 3 cps.

~

~During' spiral reload..SRM' operability will be verified by usinh a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amounttof'

. fuel is loaded,to maint.:;n'3' cps..As an alternative to the above, two fuel assemblies.wi11 be laaded in different cells containing control blades'around each SRM to obtain'.the required 3 cps. Until these two i

.assemblien have-been' loaded,^the 3 cps requirement is not necessary.

C.

iSpent Fuel pool Water-Level' To assure"thatJ here is adequate' water to shield and cool-the irradiated

~

t

' fuel assemblies. stored in the pool, a minimum pool water level-is

.catablished.-- The minimum water level of 8 ' above the top of:the fuel lis established because-it-provides adequate shielding-and is well above-l

theflevel.to-assure: adequate cooling.

u IAmendmentiNo. 61

-209-'

~

p 37-,

4:

'*\\

e.

?3.10 : BASES ((Cont'd)L D.

Time Limitation The radiological consequences'of a fuel' handling accident are based-

upon the' accidentoccurringfat least 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-after reactor shutdown.

E.. Spent; Fuel Cask Handling C

'The operation of the redundant crano in.the Restricted Mode during' fuel cask handling operations assures that the cask remains within.the

-controlled area'once it has been removed from its transport vehicle (i.e., once it.is above the 931' elevation).. Handling of the cask on the Refueling Floor"in the Unrestricted Mode is allowed only in the-case lof equipment; failures or emergency conditions when the cask is already suspended. 1The Unrestricted Mode of operation is allored'only to the. extent necessary to get the cask to'a suitable stationary position so the required' repairs can be made.

Operation with a failed controlled area' microswitch will be allowed' for a 48-hour period providing an Operator is.on'the floor in addition to the crane operator to assure

that the cask handling.is limited to the controlled' area as marked on the-floor. _This will allow adequate time to make repairs but still

.will not' restrict cask handling operations unduly.

4.10. BASES

'A.',

Refueling Interlocks Complete' functional-testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.

By loading each hoist with.a weight equal to the fuel assembly, positioning the; refueling' platform and withdrawing control rods, the interlocks can be subjected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant-logic element can-independently perform its functions.

.B.

Core Monitoring I

LRequiring-the SRM's to be functionally tested prior to any core alteration

. assures that the SRM's will be operable at the st.irt of that' alteration.

=The daily response check (or 12-hour check for spiral reload) of the SRM's ensures their continued operability.

~

E.

Spent' Fuel Cask' Handling.

The' Surveillance-Requirementsespecified assure that the redundant crane ts-adequately inspected iniaccordance with the accepted ANSI Standard (B.30.2.0) and manufacturer's recommendations to determine that the equipment'is in satisfactory condition. The testing of.the controlled-

areasilmit switches assures that the crane operation will be limited to

'the designated area in the. Restricted Mode of operation. The test of the:"two-block": limit switch l assures the power to the' hoisting motor

~

will be? interrupted before an actual "two-blocking" incident can occur.

7

'The test';of-the' inching hoist assures that this mode of load control-is

~

'available when(required.

- 7Amendinent No. J 38,5 61"

-209a--

e

av

<c.

{

r

<j3 -

[

E.-

4f.10.BA5ES (Cont'd)-

i.

E as:[ Requiring.the. lifting-andholding{ofthecaskfor-5minutesduring,the-

-initial liftiof each series of~ cask. handling operations puts a load-

~

-'w ltestion the. entire crane lifting mechanism as well as'the braking

~

.. sys tem.

-Performing this': test.when the cask is'bbing: lifted initially from the

'; cask" car assuresTthat the.' system is-operable prior to lifEing the load

~

.to an excessive height.

i

/

s s

v 8

h

'.%1 l

< ~

E

.. _h -

t, a.

DAmen M ntfNo.: 35..; 61.

s

.-209b-

.