ML20237F762

Application for Amend to License DPR-45 & Request for Exemption from 10CFR50,App J Requirement to Perform Type a Containment Bldg Integrated Leak Rate Tests.Amend to Delete Requirements Re Type a Leak Rate Tests.Fee Paid

ML20237F762
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	08/21/1987
From:	Taylor J DAIRYLAND POWER COOPERATIVE
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20237F764	List: ML20237F762 ML20237F780
References
LAC-12333, NUDOCS 8709010413
Download: ML20237F762 (8)

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y >G D DA/RYLAND

[8/h[a/ COOPERATIVE

• P O BOX 817 2615 EAST AVE SO LA CROSSE. WISCONSIN 54602-0817 (608) 788-4000 JAMES W. TAYLOR General Manager 2h 1987 l In reply, please refer to LAC-12333 1

DOCKET NO. 50-409 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 '!

Gentlemen: I

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SUBJECT:

Dairyland Power Cooperative La Crosse Boiling Water Reactor (LACBWR) l Provisional Operating License DPR-45 Exemption Request l Application for Amendment to License

REFERENCES:

(1) 10~CFR 50, Appendix J  !

(2) 10 CFR 50, Section 50.12 l (3) 10 CFR 50, Section 50.90  :

(4) 10 CFR 170.12 i (5) DPC Letter, Taylor to Docket Control Desk, l LAC-12234, dated May 22, 1987 (G) NRC Letter, Berkow tra Taylor, dated August 4, 1987 l (7) 10 CFR 100 l (8) NRC Letter, Ziemann to Linder, dated February 4, 1980 l (9) NRC Letter, Heid to Madgett, dated October 22, 1975 l I I 1

Dniryland Power Cooperative requests an exemption from the 10 CFR 50  :

. Appendix J (Reference 1) requirement to perform Type A Containment Building l Integrated Leak Rate Tests. DPC also requests revisions to LACBWR Technical  ;

Specifications to delete the requirements pertaining to Type A leak rate j tests. Therefore, in accordance with the provisions of Deferences 2 and 3, l an exemption request and an application to amend Provisional Operating l License No. DPR-45 for the La Crosse Boiling Water Reactor is hereby filed. l l

Reference 5 contained a request to amend the plant's license to a  !

possession-only license. LAChWR was pernmnently shut down on April 30, 1987. j Reactor defueling was completed June 11, 1987. The Nuclear Regulatory '

Commission issued the possession-only license on August 4,1987 (Reference 6).

8709010413 870821 PDR ADOCK 05000409 \/\ i P PDR O j s I

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Document Control Desk LAC-12333 page 2 August 21, 1987 Reference 2 requires that specific circumstances must be present for the Commission to grant an exemption. Two of the specific circumstances listed apply to this exemption request. They are:

50.12.(a)(2)(ii) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule; and 50.12.(a)(2)(iii) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated; The primary purpose of the Type A test is to assure that the containment boundary is maintained in a condition such that its leakage will not exceed an allowable amount when subjected to the peak internal pressure of the design basis accident. The maximum allowable leakage rate is associated with the maximum allowable accident dose to a person at the Exclusion Area Boundary (EAB) per 10 CFR 100. The maximum allowable dose to an individual at the EAB is 25 rem whole body or 300 rem thyroid within a 2-hour period immediately following the onset of the release.

Since the reactor at LACBWR has been permanently shut down, the design basis accident, which is a recirculation line break while operating et full power, can no longer occur. There is no postulated accident remaining which can pressurize the Containment Building. Therefore, the Containment Building no longer needs to act as a iost-accident pressure boundary.

The Fuel Element Storage Well (FESW) is located in the Containment Building. All irradiated fuel is now being stored in the FESW. The previously analyzed fuel damage accidents involving fuel in the FESW have been re-examined using the existing fuel loading. These 2 accidents are a fuel handling accident in which 2 assemblies are damaged and a cask drop event in which the cladding of all the spent fuel is ruptured. The cask drop accident is the more severe event.

The assumptions used in the re-evaluation were similar to those used in the re-racking analyses (References 8 and 9). The only significant gaseous fission product available for release is Kr-85. It was conservatively assumed that 30t of the Kr-85 inventory would be released to atmosphere. No credit was t oken for containment integrity. The fuel invent ory projected for October 1987 was used. Attachment 1 documents the dose calculations. If a ground level release was assumed, the projected whole body dose to a person at the exclusion area boundary was approximately 37 mrem within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of l the cask drop accident. If an elevated release was assumed, the maximum '

offsite whole body dose was approximately 4.0 mrem. Both these estimated doses are considerably below the 10 CPR 100 dose of 25 Rem (25,000 mrem) whole body within a 2-hour period.

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Document Control Desk LAC-12333 page 3 August 21, 1987 j Since these doses were projected assuming no containment integrity, it can be concluded that containment is not necessary to protect the public from an accident at the permanently shutdown and defueled La Crosse Boiling Water Reactor. Therefore, a Type A Integrated Leak Rate Test is not necessary.

The lack of a postulated pressure-creating accident also confirms that a Type A test is not necessary to protect the public from a pressure generating accident in the Containment Building. For these reasons, the criteria discussed in 10 CFR 50.12. (a) . (2) . (ii) apply to the performance of Type A tests at LACBWR.

performance of a Type A test is an extensive, complex, expensive evolution. Performance of the test at the permanently shutdown LACBWR would result in costs and radiation dose significantly in excess of those incurred by other permanently shutdown plants who are not performing Type A testing.

Therefore, the exemption criterion contained in 10 CFR 50.12.(a).(2).(iii) also applies.

Since the specific circumstances at LACBWR meet the criteria for an exemption, DpC believes the NRC should grant this request and exempt the plant from the Type A test requirement.

Sections 5.2.1.1, 5.2 1.2, 5.2.1.4 and 5.2.1.5 of LACBWR Technical Specifications contain references to Type A tests. DPC proposes to delete Sections 5.2.1.1 and 5.2.1.4 which apply solely to Type A tests and modify Sections 5.2.1.2 and 5.2.1.5.

The paragraph in Section 5.2.1.2 which refers to Type B testing of components which develop leaks during Type A tests will be deleted.. The acceptance criteria for Type B and C tests will be changed to specify the combined leak rate shall be less than 60% of the design leak rate of 0.1 l percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> rather than 60% of the maximum al.lowable Type A test leakage rate (which is 0.1 percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). Section 5.2.1.5 which covers reporting of test results will be modified to delete mention of Type A )

tests and of the supplemental verification test, which ie conducted as part of Integrated Leak Rate Tests t.o check Type A test data.

Attached are the proposed revised Technical Specification pages.

Finding of No_Significant Razards We have reviewed the hazards considerations referenced in 10 CFR 50 J Sections 91 and 92 and have determined that, with these criteria, no significant hazards considerations result from this proposed amendment.

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Document Control Desk LAC-12333 page 4 August 21, 1987

1. Operation of LACBWR in accordance with the proposed amendment will not igvolve a significant increase in the probability or consequences of an accident previously evaluated.

Elimination of Type A testing cannot cause an increase in the probability of an accident, since the containment only serves as a barrier. Since LACBWR has been permanently shut down and defueled, the Containment Building is not needed to protect the public against the design basis accident of a pipe break. Neither the fuel handling -

accident, nor the cask drop accident, as previously analyzed, assumed containment integrity. Since these are the postulated fuel damage accidents remaining for LACBWR, elimination of Type A testing will not involve a significant increase in the consequences of an accident previously evaluated.

2. Operation of LACBWR in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Elimination of Type A testing only serves to remove the assurance that the containment pressure boundary will limit leakage during a postulated accident. Deletion of the testing requirement, by itself, cannot create the possibility of any accident.

3. Operation of LACBWR in accordance with the proposed amendment will not

_ involve a significant reduction in the mar _ gin of safety.

Since the plant has been permanently shut down and defueled, the only remaining postulated fuel damage accidents are those involving fuel handling or a cask drop. Since the analyses for both these events did not assume any containment, elimination of Type A testing will not involve a significant reduction in the margin of safet y.

As determined by the analysis above, this amendment has to significant hazards consideration.

The license runendment has been reviewed by the appropriate review connai t t ees.

The application fee required by Reference 2 for processing a license l amendenent or exemption request ir, enclosed. I i

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Document Control Desk LAC-12333 page 5 August 21, } N7 If there are any questions, please contact us.

Sincerely, DAIRYLAND POWER COOPERATIVE ,

James W. Taylor, General Manager i.

I JWT: LSG: dh Enclosure.s cc: Mr. A. Bert Davis U.S. Nuclear Regulatory Cormaission Region III Mr. Peter B. Erickson, LACBWR Project Manager Division of Nuclear Reactor Regulations U.S. Nuclear Regulatory Commission NRC Resident Inspector Mr. L. L. Smith, Director Electric and Water Bureau Wisconsin Public Service Commission P. O. Bor 7854 i Madison, WI 53707 STATE OF WISCONSIN )

)

COUNTY OF LA CROSSE )

l Personally mne before n:e tbia dh day of _

, 1987, the above named, James Taylor, to me known to be the pe._ n who executed the foregoing instrument and acknowledged the same, c

n ~~

/

Notary Public, L rosse County Wisconsin My commission expires O' N .

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• q h'4 , 8 k 'N , -et' ATTACINENT 1

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ILOS,E CA1CULATIONS x i 5

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These calculations ace, based on the fuel inventory projected for October 1987. +

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A. Fuel Handling Accidspj l'

1his accident involveh the rupture of the cladding of the two spent fuel assemblies with the highest inve .' ory of Kr-85 during a fuel handling accident. fhe maximum dose to an individual at the Exclusion Area o Boundary (CAB) fr om the release of Kr-85 is determined conservatively.

There is about 1.192 E+5 Curies of Kr-85 remaining in the 333 spent fuel assemblies located in the FESW.*

Fuel assemb$ies #4-3 and #4-18 from fuel cycle 11 should have the e highust Kr-85 inventory. .'These fuel assemblies are computed to have

.J438 Curies of Kr-85 total'. ] !

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l. ,Gfound level release, assuming go containment building integrity E2r,.J191d-up, w .

l 1his is a conservative calculation and should represent a worst case < situation for this type of accident. Fumigation conditions at

/ the EAB are assumed with a release duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> commensurate

$ with 10 CFR 100 and Regulatory Guides 1.24 and 1.25.

A FESW water de-ontacination factor of 1.0 (no decontamination) for l'

dr-85 is assumed. The plenum or gap Kr-85 activity represents i about 15% of the total, or 279 Curies. However, for conservatism and commensurate with Table 1 of the Safety Evaluation for the

, July 13, 1979 License Amendment, 30% of the total Kr-85 activity, s, or 431.4 Curies, is assumed to be released in this accident Scenario.

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2. Average ir-85 Release Rate 1 431.4 Curies = 6.00 E-2 Ci/sec 2 hrs. x 3600 sec/hr X

3. Worst case Q for two hours,at 338m NE or 338m SSE, using Regulatory Guide 1.25 = 2.2 E-3 see f m3- 1 1

• The fuci radionuclides inrentory was calculated using an internal computer program, (ACT, which looks at number of fuel assemblies, their

{ power level and exposure, fission product production rate, nuclide activity, nuclide decay constant, and decay time.

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4. Kr-85 average concentration at 338m.

6.00 E-2 Ci/sec x 2.2 E-3 sec = 1,32 E-4 91 m3 m3

5. Immersion Dose Conversion at 338m.

1 Kr-85 Gamma Whole Body Dose Factor (Regulatory Guide 1.109):

1.61 E+1 mrem /yr x 105 uCi x 1.142 E-4 yE = 1,830 mrem /hr  !

uCi/m3 Ci hr Ci/m3 I

Whole Body. Dose at 338m:

1,839 mrem /hr x 1.32 E-4 gi x 2 hr = 0.49 mrem Ci/m3 .m3 s

l Kr-85 Beta / Gamma Skin Dose Factor (Regulatory Guide 1.109): '

1.34 E3 mrem /hr x 106 ugi x 1.142 E-4 IE = 1.53 E5 mrem /hr  !

uCi/m3 Ci hr C1/m3 Skin Dose at 338m:

1.53 E5 mrem /hr x 1.32 E-4 91 x 2 hr = 40.39 mrem Ci/m3 m3 B. Cask or Heavy Load Drop Accident This accident involves the rupture of the cladding of all 333 spent fuel assemblies located in the FESW, due to the droppage of a spent fuel shipping cask on top of the two-tiered racks containing the fuel. There is about 1.192 E45 Curies of Kr-85 present inside the 333 fuel assemblies. This analysis is a conservative calculation assuming no containment' integrity and a ground level release. Again, 30% of the Kr-85 inventory is assumed to be released to the atmosphere which is at least a factor of two times the fuel plenum (gap) Kr-85 inventory.

Therefore, 35,760 Curies of Kr-85 is released within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> post- cask drop accident.

1. Average Kr-85 Release Rate 35.760 Ci =

4.967 Ci/sec 2 hrs x 3600 sec/hr PC4-20a 2

2. Worst Case X/Q, ground level / immersion at 338m = 2.2 E-3 sec/m3 1

3. Kr-85 average concentration at 338m. i 4.967 Ci/sec x 2.2 E-3 sec/m3 = 1.093 E-2 gi m3  ;

4. Whole Body Dose at 338m.

1.093 E-2 C1 x 1839 mrem /hr x 2 hr. = 36.78 mrem m3 Ci/m3

5. Skin Dose at 338m:

1.53 E5 mrem /hr x 1.093 E-2 gi x 2 hr = 3.345 mrem Ci/m3 m3 C. Elevated Releases Elevated releases, which would be more likely, would result in an estimated highest X/0 at 500m E of 2.3 E-4 sec/m3 for 0-2 hours.

Therefore, for the two accident scenarios, the offsite whole body and skin doses would be:

1. Fuel llandling Accident

a. 2.3 E-4 x 0.49 mrem = 0.05 mrem whole body penetrating 2.2 E-3

b. 1.0456 E-1 x 40.39 mrem = 4.22 mrem skin

2. Cask Drop Accident  !

a. 1.0456 E-1 x 40.19 = 4.02 mrem whole body penetrating

b. 1.0456 E-1 x 3345 = 350 mrem skin 1

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