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)

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• Ragulat ory Flie Cy~ 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 Wade Larxin Group Vice President . Power Supply i

On this S/ day of h/EmN'd ,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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- EXHIBIT A 4 MONTICELID NUCLEAR GENERATING PIANT g DOCKET NO. 50-263 k

M" SUPPLEMENT NO.1 IO CHANGE REQUEST NO. 3 PROPOSED CHANGES TO ';HE TECHNICAL SPECIFICATIONS 7 APPENDIX A 0F PROVISIONAL OPEPATING f' LICENSE No. DPR-22 (i

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$ 1. 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: 1 f" l

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

by dropout of any increment of any one contml blade vill 4-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

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,. console is following the control rod program.

k.3.B.3 (a) To consider the Bod Worth Minimizer operable, the following

s. e steps must be perfomed:

J v' f ( i) 'Ibe control rod withdraval sequence for the Rod 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 c least one out-of-cequence control rod in each fully

," inserted group shall be verified.

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

increment prior to withdrawal vill not make the core more than 15% 25 k 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 7 260 cal /gm is below the energy content at which rapid fuel dicpercal 4'-

and primary system damage are assumed to occur. The philocophy of de-veloping a control rod withdrawal sequence, the associated rod worthe, (i

V and the consequences of a centrol rod drop accident for cuch a rod T pattern are discussed in General Electric Topical Report NED0-10527, p " Sod Dmp Accident Analysis for large Boiling Water Beactors," thrch, 1972 gw .

2he Bod Worth Minimizer provides automatic supervision to assure that ff b" cut-of-sequence control rods will not be withdrawn or inserted, i.e.,

4,* it limits operator deviations from planned withdrawal sequences. net.

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

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-ginal intent of the Specifications. The maximum allevable worth of a i control rod is presented differently baced on the improved evaluation i

' st" of the Rod Drop Accident. Ibe specific number associated with the l

(( Specification was derived by referring to Figure 3-9 in the referenced i t; topical report.

"3 This shows dotted curves of constant accident conse-i quences as a function of rod worths vs. moderator density and power level.

M Limitirs the reactivity by which the co2e can be made supercritical by i dropout of any increment of any one rod to 15% ts k provides sufficient

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,;.g 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

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

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!a all control rod positions after cc:npleting the witharaval of each m 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.

p As many as 50 rod groups may be withdrawn before reaching ICr% of rated 4

V pover therefore, the md pattern may be verified as many as 50 times 3 i

during startup. This metbod of surveillance is to be contrasted to Or the RWM method which performs a single control rod scan when initiated and thereafter runs in an operator fo11over mode. The reduced poesi-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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? 2. 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

'P at the beginning of the operating cycle and the calculated value of core l reactivity any time later in the cycle where it vould be greater than at 94 the beginning. For the first fuel cycle, R was calculated to be .012 o k. l

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

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

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

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:

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

per day. An inspection of the interior of the pressure suppression cham- t ber shall be conducted during each refueling outage. Th e inspection  !

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

( of painted surfaces.

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?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:

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22e change replaces the vord " data" with the words " stack ste:ples obtained qua ne.-ly." This more explicity defines how the surveillance vill be conducted.

,, iguid Effluents l l_ Page 171, Item 3 6.C.2. Change to read:  ;

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

I Benson for change;  !

1 The Specification is precently worded to include only beta activity. It i ,- -

should correctly include both beta and gamma activity.

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