ML20024G755

Application for Amend to License DPR-22,consisting of Suppl 1 to Tech Spec Change Request 3

ML20024G755
Person / Time
Site:	Monticello
Issue date:	09/21/1972
From:	Larkin W NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G754	List: ML20024G753 ML20024G755 ML20024G756
References
NUDOCS 9104290332
Download: ML20024G755 (5)

Text

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r UNITED STATES A'IOMIC ENERGY C0!NISSION Re:'ived w'Lir estod -

NORTHERN STATES POWER COMPANY Monticello Nuclear Generating Plant Ibcket No. 50 263 REQUEST FOR AUTHORIZATION OF A CHANGE IN TECHNICAL SPECIFICATIONS OF APPENDIX A

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PROVISIONAL OPEPATING,.ENSE NO DPR.22 (Supplement No.1 to Change Request No. 3)

Northern States Power Con:pany, a Minnesota corporation, requesto authorization for cha.nges to the Technical Specifications as shown on the attachnents labeled Exhibit A and Exhibit B.

Exhibit A describes the proposed changes along with reasons for change.

Dchibit B is a copy of the Technical Specifications marked up to indicate the proposed changes.

This request contains no restricted or other defense infornation.

NORTHERN STATIE POWER COMPANY g,f'

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$;//ch' Ki>' 'X/n Dy Wade Larxin Group Vice President. Power Supply i

h/EmN'd On this S/

day of

,1972, before me a notary public 4

in and for said County, personally appeared Wade larkin, Group Vice President Pover Supply, and being first duly sworn acknowledged that he is authorized to execute this docu=ent in behalf of Northern Statec Power Cos:pany, that he has read it and knows the contents thereof, that to the best of his knowledge, info:r.ation and belief, the statements cade in it are true and that it is not interposed for delay.

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I EXHIBIT A 4

MONTICELID NUCLEAR GENERATING PIANT g

DOCKET NO. 50-263 k

M" SUPPLEMENT NO.1 IO CHANGE REQUEST NO. 3 7

PROPOSED CHANGES TO ';HE TECHNICAL SPECIFICATIONS APPENDIX A 0F PROVISIONAL OPEPATING f'

LICENSE No. DPR-22 (i

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Control Rod Worth and Rod Worth Minimiter l;c pages 77 & 78, Items 3 3.B.3 and 4 3.B.3, Change to read:

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3 3.B.3 (a) Control rod withdrawal sequences shall be established so that the maximum calculated reactivity that could be added 3

by dropout of any increment of any one contml blade vill 4-c not make the core more than 1 5% A k supercritical.

2' 3 3 3 3 (b) Whenever the reactor is in the Startup or aan mode below lo) rated themal power, no contml rods shall be moved unless the w

rod vorth einimizer is operable or a second independent opera-tor or engineer verifies that the operator at the reactor

,'M console is following the control rod program.

k.3.B.3 (a) To consider the Bod Worth Minimizer operable, the following steps must be perfomed:

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i) 'Ibe control rod withdraval sequence for the Rod J v' Worth Minimi:er computer shall be verified as ff correct.

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( 11) The Rod Worth. Mini =1:er co=puter on-line diagnos-

!e tic test shall te successfully completed.

(iii) Proper annunciation of the selection ermr of at least one out-of-cequence control rod in each fully c

inserted group shall be verified.

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( iv) The rod block function of the Rod Worth Minimi:er y

shall be verified by atterpting to withdraw an out-of-sequence control rod beyond the block point.

k.3.B.3 (b) If the Rod Worth Minimi:er is inoperable while the reactor is in the Startup or Run mode belov 1012 rated ther.::al power, the

,F second independent operator or engineer shall verify that all

-Q rod positions are correct prior to comneneing the withdrawal of each md group.

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2-Page 8k, Bases, Change Section 3 to read:

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The Bod Worth Minimiser restricts withdravals and insertionc of control rods to those listed in prespecified control rod cequences which are 1

i established such that the maximum calculated worth of any control 2cd I

increment prior to withdrawal vill not make the core more than 15% 25 k g

supercritical.

These sequences are developed to limit the reactivity vorths of control rods in the core and together with the integral rod velocity limiters limit potential reactivity insertion cuch that the a

results of a control rod drop accident will not exceed a max' mum fuel energy content of 280 calories / gram.

The peak fuel energy content of 260 cal /gm is below the energy content at which rapid fuel dicpercal 7

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and primary system damage are assumed to occur.

The philocophy of de-veloping a control rod withdrawal sequence, the associated rod worthe, (i

and the consequences of a centrol rod drop accident for cuch a rod V

pattern are discussed in General Electric Topical Report NED0-10527, T

" Sod Dmp Accident Analysis for large Boiling Water Beactors," thrch, p

1972 gw.

2he Bod Worth Minimizer provides automatic supervision to assure that ff cut-of-sequence control rods will not be withdrawn or inserted, i.e.,

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it limits operator deviations from planned withdrawal sequences.

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Section 7 9 PEAR.

It serves as an independent backup of the normal h$

withdrawal procedure followed by the operator.

In the event that the j(yve.

RdM is out of service, when required, a second independent operator or f

engineer can manually fulfill the operator-follover control rod pattern bf conformance function of the R4M.

In this case, an extra measure of 5

procedural control is exercised in that all control rod positionc are "1

verified after the withdruval of each group, prior to proceeding to the next group.

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Penson for chance:

"i Tbc wording of the present Specification 3 3.B.3 (a) is more restrictive A

than intended. A strict, verbal interpretation of the present wording

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1) prohibite withdrawal of the first rod during startups, and 2) inter-feres vitn testing required elsewhere in the Specifications.

The pro-i posed vording allows for these situations while being within the ori-i ginal intent of the Specifications.

The maximum allevable worth of a control rod is presented differently baced on the improved evaluation

' st" of the Rod Drop Accident.

Ibe specific number associated with the i

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Specification was derived by referring to Figure 3-9 in the referenced t;

topical report.

This shows dotted curves of constant accident conse-i "3

quences as a function of rod worths vs. moderator density and power level.

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Limitirs the reactivity by which the co2e can be made supercritical by M

dropout of any increment of any one rod to 15% ts k provides sufficient i

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margin to the design limit of 280 calr/gm.

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Specifications 3 3.B.3 (b) and 4 3.B.3 increase the requirements for the jn, surveillance of the rod vorth minimi:er and for the substituted inde-pendent procedural controle during an outage of the RdM.

The second

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Person verifying the rod pattern vill perform a manual verification of 1

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I 1!a all control rod positions after cc:npleting the witharaval of each group.

The order of withdraval of rods within a group is r.at re-m stricted and therefore a more frequent verification is not required.

m As many as 50 rod groups may be withdrawn before reaching ICr% of rated p

pover therefore, the md pattern may be verified as many as 50 times i

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,..V during startup.

This metbod of surveillance is to be contrasted to O

the RWM method which performs a single control rod scan when initiated and thereafter runs in an operator fo11over mode.

The reduced poesi-r h" '-

bility of human error in light of the increased frequency of full core L

scans must be weighed against the pmbability that an extremely inter-gg mittent mechanical problem might cause the INM to ignore an erroneous T' I'-

rod movement and not be corrected by a suosequent scan. The reliabil-L ity of the Ibnticello RWM has shown that there is no need to excec-l*

sively turn to the second, independent person to verify rod movements.

However, to mquire RWM availability for reactor startups places an 6

unreasonable demand on the reliability of a piece of non-redundant equipment.

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Reactivity Marcin - Core Leading Page 82, Item 3 3. A.1 (basis), Change the second paragraph to read:

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The value of R is the difference between the calculated core reactivity at the beginning of the operating cycle and the calculated value of core l

'P reactivity any time later in the cycle where it vould be greater than at l

94 the beginning.

For the first fuel cycle, R was calculated to be.012 o k.

.: q A new value of R must be detemined for each fuel cycle.

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Reason for chance.

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[k Becent calculations, as reported in our February 3,1972 letter to Dr.

i Peter A lbrris, indicate that the value of R for the first cycle is.012 g

rather than.003 ok as is currently stated in the Tech &al Specifica-V tions.

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.; L8uppression Chamber Inspection

!Page 139, It.m 4.7. A.1, Change to mad:

' he suppression chamber water level and temperature chall be checked once s

per day. An inspection of the interior of the pressure suppression cham-t M.

ber shall be conducted during each refueling outage.

Th e inspection shall consist of a visual examination of mechanical and stractural integ-m rity of hangers, piping and structural members, and a visual examination 1

( of painted surfaces.

+ Reason for chance'

?De proposed wording defines what the suppression chamber inspection in to

include, i

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,[Iirborne Effluent Measurement a

page 169, Item 4.8. A.1, Chrmge the last centence to read:

Gaseous releese of tritium shall be calculated on a monthly basis from tr measured stack sar::ples obtained quarterly.

Beason for chance:

22e change replaces the vord " data" with the words " stack ste:ples obtained p~,

qua ne.-ly."

This more explicity defines how the surveillance vill be conducted.

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Page 171, Item 3 6.C.2. Change to read:

The concentration of gross beta and gnmn activity........

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I Benson for change; 1

The Specification is precently worded to include only beta activity.

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should correctly include both beta and gamma activity.

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