ML20078M125

Informs NRC of Interim Justification for Continued Operation within Current Licensing Basis.Proposed Action Plan Included

ML20078M125
Person / Time
Site:	Pennsylvania State University
Issue date:	02/06/1995
From:	Dhirley D PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9502130250
Download: ML20078M125 (2)

Text

-

4 PENNSTATE  %"'L Dasid A. Shirley The Penns)hania state University Senior Vice President for Research 207 Old Main and Graduate Education University Park, PA 16802-1503 February 6,1995 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Re: Report of Changes from the Safety Analysisfeport of the Penn State Breazeale Reactor, License No. R-2, Docket No. 50-05

Dear Sir or Madam:

W is a report of changes from the Safety Analysis Report (S AR) of the above referenced facility being reported in accordance with Technical Specification 6.6.2.b.(2). This report is to inform the Nuclear Regulatory Commission ofinterim justification for continued operation within current licensing bases. An action plan is proposed to restore operation to the licensed reactor power of 1 MW.

Backcround The staff at PSDR have been concerned that the peak measured fuel temperature while operating at maximum licensed power (1 MW) with our present instrumented element has increased. In addition it came to the attention of the staff that the containment volume used in the SAR was somewhat higher than actual. In the process of investigating the change in the fuel handling accident consequences as a result of using a more accurate containment volume (1900 m3 vs. 2500 m3) the staff determined that the higher fuel temperatures wem of more importance than the change in volume. The decision was made to review the entire SAR before making any new changes.

The SAR states that by following the specific requirements stated in the Technical Specifications (i.e. Tech. Spec. Docket 50-005, amendment 23 section 3.1.5) that the maximum power density would not exceed 23.2 kW per element. As long as this power density is not exceeded the conclusions of the SAR fuel handling accident analysis are correct. The staff repeated the experiments that were used as the basis for the SAR analysis and found that the maximum power density during 1 MW operation with the present core loading (Loading 47) does exceed 23.2 kW per element. After this discovery on January 9,1995, power was reduced to 570 kW which lowers the operating temperatum and power density below that analyzed in the SAR. At this power level, the maximum power density is approximately 14 kW per element.

Analysis of the Fuelllandling Accident In the SAR some very conservative assumptions wem made in analyzing the fuel handling accident.

1. The PSBR weekly operating history was assumed to be 98 hr shutdown and 70 hr of 1 MW operation.

2. Immediately after the operation the element operating at the highest power density and the highest temperature (since temperature is dependent on the heat flow resistance, i. e. fuel cladding gap size, as well as the power density) is mechanically ruptured in a handling accident, j

• r; n G ;

9502130250 950206 An 14ual opportonsty On er ty PDR ADOCK 05000005 - I.0 P PDR

, Nuclear Regulatory Commission l Page 2 l February 6,1995 l I

3, The rupture' occurs in air.

4. There is no platcout of fission products.

5. No credit is taken for filtration by the emergency exhaust system.

6. The release fraction equation (PSBR SAR chapter 9, equation 41)is assumed to be correct even though the expenmental evidence indicates that it is high by a factor of 4 (The U-Zrlix l Alloy: Its Pronerties and use in TRIG A Fuel , M. T. Simnad, General Atomics Project No. '

4314, February 1980). ,

1 The use of assumption I rather than a more realistic (but still much more intensive than the i operation to date) daily operating history of 16 hr of shutdown followed by 8 hr of full power j operation results in consequences that are more conservative by a factor of 1.4. l The combination of the assumptions 2,3,4, and 5 "... bad to computed doses principally resulting from iodines that may be at least a factor of 100 higher than in more realistic scenarios, for example, those in which the cladding fails while the element is immersed in pool water" (Safety Evaluation Renort related to the renewal of the onerating license for the Research Reactor at Pennsylvania State University, NUREG-1158, USNRC, January 1986).

The net effect of the overly conservative assumptions (i. e. assumptions 1-6) is consequences that are higher by a factor of 500. Even so the consequences are less than the 10 CFR 20 limit for an ,

unrestricted area by a factor of 30. 1 Action Plan Ourintennediate action will be to relax the power restriction to 750 kW which allows a maximum power density of approximately 18.3 kW per element. It was determined that at this power level, the fuel handling accident consequences will still be within those of the SAR. This can be done j while maintaining all of the above overly conservative assumptions except for assumption I which l will be changed to a more realistic daily operating history of 16 hr of shutdown followed by 8 hr of I full power operation. This operational schedule as well as the 750 kW power restriction will be maintained with administrative controls.

Our next action will be to change the S AR to allow variations in core loading to reduce the maximum power density and fuel temperature. With those changes the staff can justify return to full power operation. This action will be delayed until further analysis can be done to support the change.

If you have questions on this matter, please refer them directly to Marcus 11. Voth or l Daniel E. Hughes at (814) 865-6351.

l S accrely, I i

) . 1 Davi A. Shirley Senior Vice President for Research and Graduate Education DAS/ld14009.95 pc: Region 1 Administrator