ML20148C394
| ML20148C394 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 03/15/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148C351 | List: |
| References | |
| NUDOCS 8803220376 | |
| Download: ML20148C394 (4) | |
Text
[ o UNITED STATES
^ ,h NUCLEAR REGULATORY COMMISSION
[
g WASHINGTON, D. C. 20555
- j
.... / SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 59 TO PROVISIONAL LICENSE N0. DPR-45 LA CROSSE BOILING WATER REACTOR (LACBWR)
DOCKET NO. 50-409
1.0 INTRODUCTION
On April 27, 1987 Dairyland Power Cooperative (DPC, the licensee) announced that their La Crosse Boiling Water Reactor (LACBWR) would be permanently shutdown due to economic considerations and on April 30, 1987 the reactor shutdown was completed. By June 12, 1987, all fuel had been removed from the reactor and stored in the Fuel Element Storage Well (FESW). Dairyland Power Cooperative, in a letter dated May 22, 1987 (LAC-12234), requested that Provisional License No. DPR-45, for LACBWR, be amended to a possession-only license. Their request was granted by License Amendment No. 56, issued on August 4, 1987.
By letters dated August 21, 1987, (LAC-12331) and August 28, 1987, (LAC-12339), DPC submitted the third in a series of proposed changes to the LACBWR Technical Specifications to bring them into agreement with the possession-only license. The proposed changes would delete the TS requirements for, and references to, primary containment integrated leakage rate testing (ILRT)/ Type A Testing which is required by 10 CFR 50.54(o) and Part 50 Appendix J for operating power reactors. The licensee stated that, since LACBWR can no longer operate, the peak
. internal pressure of a design basis accident is no longer possible and, I
therefore, the TS requirement for the ILRT, made at the peak containment pressure, is no longer necessary. Type B and Type C Tests as described i
in 10 CFR Part 50, Appendix J, would be retained.
We have also considered an exemption from 10 CFR 50.54(o) and Part 50, Appendix J requirements for ILRT/ Type A testing of the containment building, l 2.0 DISCUSSION l
l As an operating nuclear power plant, LACBWR had an approved Containnent l Building (CB), meeting the requirements of 10 CFR 50.54(o), Appendix A Criterion V, and Appendix J. The objectives of the CB at an operating plant is to provide a final barrier against the release of significant amounts of radioactive fission products in the event of a loss-of-coolant accident (LOCA) when it must contain the pressure and temperature conditions resulting from the worst foreseeable mass and energy releases of a design basis accident. Since LACBWR has been permanently shutdown, the energy source which might pressurize the containment has been removed and ILRT/ Type A testing requirements are no longer Och O 9 h!O P
applicable to LACBWR. The worst foreseeable accident has been reduced to the damage of irradiated fuel cladding and the release of the cladding gap fission products into the FESW and thence to the CB without any buildup of pressure or temperature. Therefore, the potential radiological hazard to the staff and public has been greatly reduced and will continue to be lessened with the passage of time and the decay of fission and activation products.
3.0 EVALUATION In their August 21, 1987, letter the licensee proposed to eliminate the TS requirements for Type A leakage rate testing. The proposal is based on the premise that with the reactor permanently shutdown with all fuel removed, the design bases for the primary containment is no longer applicable as,the energy source to pressurize the containment no longer exists and there are no other postulated accidents which could pressurize the CB. Therefore, the CB is no longer needed as a post-accident pressure boundary.
Since all of the irradiated fuel is stored in the FESW which is located inside the CB, the worst foreseeable accident with radiological consequences becomes a fuel cask drop into the FESW rupturing the spent fuel.
The licensee submitted an evaluation of the radiological consequences of two fuel rupture accidents; the consequences of the rupture of the rods in two fuel elements caused by the drop of one fuel element over the FESW, and the other a fuel cask drop with the rupture of all 338 fuel elements stored in the FESW.
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The evaluation was based upon the criteria of Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences
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of a Fuel Handling Accident, March 23, 1972, and two previous evaluations made by the Staff, one for Amendment No. 18 which permitted increasing the storage capacity of the FESW, dated February 4, 1980 and the other for the Safety Evaluation Program (SEP) Topic XV-20, Radiological Consequences of Fuel Damaging Accidents-LACBWR, dated October 14, 1982.
Since LACBWR was shutdown on April 30, 1987, essentially all iodine activity has decayed and the only gaseous fission product available for release is Kr-85; therefore, only whole body or skin doses are of concern.
Conservative assumptions used in the worst case cask drop accident included: no credit for containment; 30 percent of the 119,200 curies of l
Kr-85 contained in the fuel is released from the ruptured fuel; no Kr-85 absorption in the pool water; and release from the CB at ground level over a two hour period. The atmospheric diffu: ion factor used in their l
evaluation is identical to the most conservative one used by the Staff in the above evaluations, X/Q = 2.2 E-3 sec/ cubic meters' at the exclusion area boundary (EAB) or 338 meters from the release point. The diffusion factor included the building wake effect for ground level releases.
The FESW Kr-85 activity was calculated using the FACT Code develo3ed by the licensee. Results of the FACT calculations were compared wit 1 hand calculations for other long half-life isotopes and with other independent calculations and were found to be in good agreement.
Using the dose factors found in Regulatory Guide 1.109 for exposure to a semi-infinite cloud of noble gases, the calculations indicated that the whole body dose would be 40 mrem and the skin dose would be 3.3 rem at the EAB. Other calculations submitted with the application to amend their Emergency Plan, dated September 29, 1987, indicated the whole body dose at the owner controlled fence, 100 meters from the the release point, would be approximately 188 mrem.
The Staff has reviewed the licensee's methodology in estimating the '
radiological consequences of a cask drop and find that it agrees with the Staff's previous evaluations. Standard Review Plan, NUREG-0800, in Section 15.7.5, Spent Fuel Cask Drop Accidents, states that plant siting and dose mitigating engineered safety features systems are acceptable if the whole body and thyroid doses at the EAB are well within the exposure guideline values of 10 CFR Part 100. The dose at the owner controlled fence is less than 1.0 percent of 10 CFR Part 100 value.
The proposed amendment deletes only the Type A Test requirements. The
- licensee plans to maintain the CB and it's existing penetrations and to continue the required Type B and Type C testing which will assure continued performance of the CB isolation valves, both air locks, and other existing penetrations. The Staff concurs that Type A testing is no longer required as the consequences of the worst foreseeable accident would not cause a significant containment pressure. In addition, if no credit for containment is assumed, the potential offsite exposures would be a small percentage of 10 CFR Part 100 values. The staff also concludes that tha licensee should also be exempted from 10 CFR Part 50.54(o) and Part 50, Appendix J with respect to ILRT/ Type A testing for the reasons discussed above.
l l 4.0 ENVIRONMENTAL CONSIDERATION l
This amendment involves a change to a requirement with respect to the installation of use of facility components located within the restricted i area as def.ined in 10 CFR Part 20. The Staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released l
l t . _ - _ _ - -
offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility fi criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendnent.
5.0 CONCLUSION
The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in a Notice of Consideration of Issuance of Amendment to License in the FEDERAL REGISTER on September 23, 1987 (52 FR 35790). No requests for hearing and no public connents were received.
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
Principal Contributor: X. R. Ridgway Dated: March 15, 1988
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