ML20138J765

Application for Amend to License DPR-45,changing Tech Specs to Incorporate Use of Design AA Replacement Control Rods & Deleting Tech Spec 4.2.4.10.Fee Paid

ML20138J765
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	12/12/1985
From:	Taylor J DAIRYLAND POWER COOPERATIVE
To:	Zwolinski J Office of Nuclear Reactor Regulation
Shared Package
ML20138J769	List: ML20138J765 ML20138J775
References
LAC-11317, NUDOCS 8512180028
Download: ML20138J765 (7)

Text

.. ; ; .s D A LAND

1. [M[COOPERAT/VE

• P.O. *BOX 2615 EAST 817 AV SOUTH +

LA CROSSE. WISCONSIN 54601 (608) 788-4000 December 12, 1985 In reply, please refer to LAC-11317 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation Attn: Mr. John Zwolinski, Chief Operating Reactors Branch #5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

DAIRYLAND POWER C00EPRATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)

PR071dIONAL OPERATING LICENSE NO. DPR-45 APPLICATION FOR AMENDMENT TO LICENSE

References:

(1) 10 CFR 50, Section 50.90 (2) 10 CFR 170.12 (3) DPC Letter, Linder to Keppler, LAC-9236, dated July 29, 1983 (Copy to US-NRC-I6E, Washington, DC)

(4) DPC Letter, Madget to Reid, LAC-4523, dated February 25, 1977.

(5) DPC Letter, Linder to Ziemann, LAC-6846, dated April 1, 1980.

(6) DPC Letter, Linder to Ziemann, LAC-6905, dated May 8, 1980.

(7) NRC Letter, Crutchfield to Linder, dated December 5, 1990 (Sic. 1980).

Gentlemen:

Dairyland Power Cooperative (DPC) hereby requests the following changes to the LACBWR Technical Specifications. In accordance with the provisions of Reference 1, this application to amend Provisional Operating License No.

DPR-45 for the La Crosse Boiling Water Reactor is being filed with three (3) signed original applications, together with thirty-seven (37) copies.

Technical Specifications 2.7.3.2 and 2.7.3.3 currently state:

2.7.3.2 The control rode shaLL be of cruciform shape and shall consist of 80 Inconel 600 tubes containing B4 C pellete and packed in a perforated sheath of AISI Type 304 stainless steel. The control rod blades shall nominally be \

0.313 inch thick and 9.750 inch wide and 85.876 inch long. b0 9 EB (LIAW) pey PSB (L. HULMAN) .00 uD'4 . A . EICSB ($RINIVASAN)

F5121EDO2B 851212 R58 ( ACTIP"3) g PDR- ADOCK 05000409 P

FOB (VASSALLO) g PDR AD - c. LAINAs (Ltr only)_ _ ac.* _

r 5 .

.Mr. John Zwolinski, Chief December 12, 1985

_; Operating Reactors Branch #5 LAC-ll317 2.7.3.3 ' The BgC pellete shall have nominal diameter of 0.157 inch; and the

. pellet column shall'have a nominal Length of 83 inch.

DPC requests that these Technical Specifications la combined and changed to read as'follows:

2 7.3.2 The control rods shall be of cruciform shape and shall conform to

'either the Allis-Chalmers (A-C) ^ tube-sheath design or the ASEA-ATOM (AA) plate design with horisontally ch illed absorber holes. The neutron absorber section of each control rod shall have a nominal length of 83 inch.

This change is needed because the. replacement control rods for the LACBWR are of the AA design instead of the tube sheath design referenced in our current Technical Specifications. The AA control rod design is described in

' detail in the Topical Report TR-UR-85-225, "BWR Control Blades for U.S. BWR's" submitted to the NRC by ASEA-ATOM in October 1985. In the attachment to this

. letter'the~ specific AA control rod for the LACBWR is described along with a comparison of the physical and neutronic parameters of the LACBWR A-C and AA

-control rods.

-The AA control rods for the LACBWR have been designed to closely match Ethe reactivity _ worth of the original (A-C) control rods and to be rechanically compatible with all reactor-components and control rod handling equipment.

E The AA control rod is essentially identical in exterior envelope to the A-C rod. .The AA control rod is slightly lighter (approximately 7 lbs.) than the sc . -C rod and therefore, scram times.for the AA rod should be approximately the A _

same as or slightly-faster than for the A-C rod. Scram times for the AA control rods will be measured after installation in the reactor as required by current procedures.

t The reactivity worth of the.AA control rods for the LACBWR, relative to L the original A-C control rods, was' calculated by ASEA-ATOM using the two-dimensional lattice depletion code PHOENIX. The B 4 C region of the AA

. ' control rods was found to have a consistentl3 aigher reactivity worth, for the all rods inserted condition,L of about 2.6-3.0% (relative) for the various reactor. conditions. Corresponding values for the'AA control rod tip (5.9 inch) with Hf are 5.0-5.9%'(relative) less worth than that of the tip of the A-C rod.. The over all reactivity worth of the AA control' rod, in the all rods inserted condition,'is about 1.3-1.8% (relative) greater than the A-C rods for the various reactor conditions. In the hot, full power, 20% moderator void case with,75%'or more of the? control rods withdrawn from the core (average operating condition) the relative reactivity worth of the AA control rod is

. 0.0-0.3% less than the A-C rod.

in For. negative reactivity insertion events (scram events) the slightly

, greater reactivity worth of the.AA control rods is expected'to be beneficial.

~

Slightly more negative reactivity will be inserted faster than with the A-C control = rods. -The~ shutdown margin will also be'slightly greater with AA

_ control. rods than with the current'A-C control rods.

. For positive reactivity insertion events, i.e. inadvertant control rod m

. - WP3.6.

,

Mr. .Jchn Zw311n ki, Chief Dacssbar 12, 1985 Operating Reactors Branch #5 LAC-11317

. withdrawal and control rod drop accident, the slightly greater worth of the AA control rod is also expected to have-a minimal effect. The limiting anticipated transient for the LACBWR is the inadvertant control rod withdrawal aat operating power (References 4 and 5). A recalculation of the limiting

, control rod-withdrawal transient for the beginning of Fuel Cycle 9 using a conservative 3% (relative) greater control rod worth resulted in a MCPR of 1.529 compared to a MCPR of 1.539 calculated for the A-C rods. The effect produced by the actual LACBWR AA control rod with a relative worth equal to or slightly.less than the A-C rods in the operating reactor would be completely negligible. In the LACBWR, the consequences of a control rod drop accident are greatest when the reactor is at power. The results of the probability study of the LACBWR control rod drop accident (References 6 and 7) are not very sensitive to the specific worth of the control rods and the small differences between the AA and the A-C control rods would have an insignificant effect on the results.

-- Technical Specification 4.2.4.10 currently states:

~~4.2.4.10. During every scheduled shutdown for refueling or replacement of shroud cans, the most exposed control rod or other control rod with canparatively high exposure shall be inspected by use of a 0.477 inch go-gage.

If the control rod vill not pass the go-gage, it shall be removed from service and the next most exposed rod examined. However, any rod uhich has controlled an average of 1.00% detta k/k for 12,800 MWD /MT 'of reactor operation (12,800%

' detta k/k MWD /MT) or whose product of reactivity controlled and time in NWD/MT

-exceeds'12,800 % delta k/k NWD/MT, must be inspected within intervals of 2500%

detta k/k MWD /MT.

-DPC requests that this specification be deleted from the LACBWR Technical Specifications.

This request is based on the fact that the Techncial Specifications for other U.S. boiling water reactors and the NRC Standard Technical Specifications for BWR's do not contain a similar requirement anj alao on the

.following discussion.

.The requirement to gage the control rods was originally based on the belief that the life of the rod would be limited by the pressure buildup in the B4 C tubes due to helium release from the irradiated B4 C and that the gage would detect tube swelling before tube failure. : Experience has shown that absorber tubes can fail by intergranular stress corrosion cracking (ICSCC) and B 4 C be lost (Reference 3) before any swelling is' detected by gaging. We have never detected swelling of a control rod blade even with 0.420 inch go gages.

The industry has concluded, after extensive research, that absorber tube failure by IGSCC and B4 C loss is more a function of B-10 depletion, and resultant B 4C swelling and. change in physical characteristics, than a result of pressure buildup from helium release. The requirement for more frequent inspections after reaching an exposure of "12,800% delta k/k MWD /MT" (and the resultant requirement to remove the rod from the' rod group used for control while.at power) was also based on the theory of pressure buildup due to helium release and not on any limitation due to reduction of reactivity worth or B-10 depletion.

WP3.6

- Mr. Jsha Zwolinrki, Chief

~

Dscamber 12, 1985 Operating Reactors Branch #5 LAC-11317

,~ In'any case,,this technical specification would not be appropriate for Jthe AA control" rod because of its different design and much longer expected u'seful life.- Some of the improvements in the design of the AA control rod'

• which promote a longer life are:

"* The'use'of hafnium metal for neutron absorption in the top 5.9 inches of

.the absorber region. Hf does not deplete as fast as B-10 in B C4 and do'es

t_ not swell'with irradiation and, therefore,' effectively eliminates the very significant problem of high local B-10 depletion.in the tip of an all B4 C

, rod like the A-C rod.

Approximately 50% greater B4 C content per unit length in the B4 C region

-than the A-C rod. Therefore, a correspondingly longer exposure time is .

' required for a given decrease in reactivity worth or relative B-10 depletion.

High purity (reduced levels of Si, P, _ and S impurities) Type 304L stainless l steel which'has a lower susceptibility to ICSCC than ordinary 304L

' stainless steel.

. Minimum stainless steel surface in contact with the reactor coolant.

Combined _with the absence of crevices, this further minimizes corrosion risk. .

' Equalization of gas pressure'throughout'the complete control rod wing.

'In lieu of the requirements in the current technical specification 4.2.4.10,~DPC will continue to follow the exposure history and B-10 depletion of each control. rod in the LACBWR 'and will. base a prudent control rod examination and control rod shuffle'and/or discharge program on this ' data.

= yhe present plans are 'to discharge 'all- the high exposure A-C control rods presently in the core at the next_ refueling outage scheduled for early spring

'1986. During subsequent operation AA control rods will be used for controlling rods during power operation. cThe exposure level (B-10 depletion) of the A-C control rods remaining in the LACBWR core will be a small fraction L , of that experienced by the control rod examined in the hot lab (Reference 3)

-and these control rods will be in positions where~further B-10 depletion

. averaged over the top quarter of their length will-be negligible.

, ASEA-ATOM has accumulated many years of operating experience with some L

1200 control rods of similar design (except for the Hf tip) to the LACBWR AA l- control rods. These control rods have performed very well in the ASEA-ATOM 1 boiling water reactors. Only after 8 to 12 fuel cycles were~a few instances of intergranular' stress corrosion cracking observed along B C 4absorber holes near the top of control rods.where the B-10 depletion was greater than 60%.

-Hot lab destructive ~ examination and neutron radiography of several control rod segments Sith IGSCC showed that B4 C washout occurs in the hole with a crack but only to a limited extent in neighboring holes due to the gas pressure equalization passages. These examinations also showed that a simple _ visual in pool inspection can detect almost all cracks, thus a visual inspection

'WP3.6: .

J - .

--- l 'i 1

m.

Mr. John Zwolinski, Chief December 12, 1985 Operating Reactors Branch #5 LAC-11317 program can be used to check the integrity of AA control rods. The improved AA control rods with Hf tips and high purity 304L SS should have a considerably longer failure free life than the rods in the above examination program.' Based on this background, ASEA-ATOM currently recommends the following guidelines for AA Hf tip control rods:

Control rods should be visually inspected each refueling when the average B-10 depletion in the B 4 C segment of the top quarter of the rod exceeds 50%.

Control rods which show no visible cracks can be used without any

restrictions for another fuel cycle.

Control rods which show visible cracks should be replaced.

The nuclear life expectancy (based on reduction in reactivity worth of the top quarter segment) corresponds to at least a 55% depletion of the average

.B-10 concentration in the B 4C of the top quarter of the rod.

Finding of no significant hazards We have reviewed the harards considerations referenced in 10 CFR 50 Sections 91 and 92 and have determined that with these criteria no significant hazards result from these proposed amendments. 10 CFR 50.92 (c)(1-3) provide the questions for review of significant hazards considerations. They are repeated here for reference.

10 CFR S0.92 (c)

(1)^ Involve a significant increase in the probability or consequences pf an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin pf safety.

As discussed above, the use of the AA' control rods as replacement rods for the A-C control rods has no adverse safety consequences. Since the AA contral rods are mechanically compatible in all respects with the LACBWR system and have essentially the same reactivity worth, their use will not involve a significant increase in the probability or consequences of an accident previously evaluated. The limitations of other control rod related Technical Specifications such-as 4.2.5.2 and 4.2.5.3 will still be conservatively met during operation with the AA control rods. Their improved design will significantly reduce the probability of IGSCC failure and resultant loss of control material (B 4 C) and therefore, will increase the margin of safety. Therefore, in accordance with 10 CFR 50.92 (c), a no significant hazards (ttermination is justified with regard to the changes requested ~in LACBWR technical specifications 2.7.3.2 and 2.7.3.3.

The deletion of LACBWR technical specification 4.2.4.10 also satisfies WP3.6 +

L

Mr. John Zwolinski, Chief December 12, 1985 Operating Reactors Branch #5 LAC-11317 the criteria for no significant hazards determination under 10 CFR 50.92 (c) because it only removes a surveillance requirement which experience has shown to be completely ineffective (and also is not required by Standard BWR Technical Specifications). Control rod lifetime will be appropriately and prudently based on exposure histories, B-10 depletion and visual examinations as is-currently the case for the rest of the U.S. BWR reactors.

The information submitted in this application for license amendment has been reviewed by the LACBWR Committees as prescribed in LACBWR Technical Specifications.

DPC requests approval of this request as soon as possible. The AA control rods are to be installed in-the LACBWR during the next refueling outage which is tentatively scheduled for early March 1986.

A copy of the revised pages of LACBWR Technical Specifications is attached. A check for $150 accompanies this letter to cover the required application fee per Reference 2.

If you have any questions concerning this submittal, please contact us.

Sincerely, DAIRYLAND POWER COOPERATIVE James W. Taylor, General Manager JWT:SJR:sks Attachment

-cc: James G. Keppler NRC Resident laspector John Stang, Project Manager Mr. Clarence Riederer, Chief Engineer Wisconsin Public Service Commission P. O. Box 7854 Madison, W1 53707 y

WP3.6

' Mr. John Zwolinski, Chief December 12, 1985 Operating Reactors Branch #5 LAC-11317 STATE OF WISCONSIN )

)

COUNTY OF LA CROSSE )

Personally came before me this !d day of , 1985, the above named, James W. Taylor, to mm known to be the person who executed the foregoing instrument and acknowledged the same.

/ L Notary P :.ic, La Crosse County, Wisconsin My commission expires February 21, 1988 a

,o h

'WP3.6 -

7-