ML20206J919
| ML20206J919 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/23/1986 |
| From: | Keiser H PENNSYLVANIA POWER & LIGHT CO. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20206J924 | List: |
| References | |
| PLA-2654, NUDOCS 8606270314 | |
| Download: ML20206J919 (4) | |
Text
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@ Pennsylvania Power & Light Company Two North Ninth Street
- Allentown, PA 18101 + 215 / 77G5151 Harold W. Keiser Vice President-Nuclear Operations 215/770-7502 JUN 23 M Director of Nuclear Reactor Regulation Attention:
Ms. E. Adensam, Project Director BWR Project Directorate No. 3 Division of BWR Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 38 TO LICENSE NO. NPF-22 PLA-2654 FILE R41-2 Docket No. 50-388
Dear Ms. Adensam:
The purpose of this letter is to request a change in the Main Steam Line Radiation - High Setpoint from three times background to seven times background in the Susquehanna SES Unit 2 Technical Specifications. This change was recently included via Amendment No. 58 in the Susquehanna SES Unit 1 Facility Operating License No. NPF-14. This proposed amendment differs from the Unit 1 amendment in that the supporting calculations consider both 8X8 and 9X9 fuel, as discussed below in the No Significant Harards Consideration.
The proposed changes (attached to this letter in marked-up form) are as follows:
o Table 2.2.1-1, Functional Unit 6 The Main Steam Line Radiation - High Reactor Protection System Trip Setpoint is raised from three to seven times full power background, and the Allowable Value from 3.6 to 8.4 times full power background.
o Table 3.3.2-2, Trip Functions 1.e and 3.b Main Steam Line Radiation - High Trip Setpoints for Primary Containment Isolation and Main Steam Line Isolation are raised from three to seven times full power background, and Allowable Values are raised from 3.6 to 8.4 times full power background.
The purpose of the main steam line radiation monitors (MSLRMs) is to indicate a gross failure of the fuel cladding. Per Technical Specification bases, the MSLRM trip setpoint is intended to be high enough above background radiation
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JtR{ 231986 Page 2 SSES PLA-2654 File R41-2 Ms. E. Adensam levels to prevent spurious trips, yet low enough to promptly detect gross cladding failure. The current setpoint of three times background was shown to be too low on Susquehanna Unit 1 in 1983, when three reactor scrams and full isolations occurred during normal plant evolutions involving the Condensate Demineralizer system.
The proposed Technical Specification change would eliminate the possibility of spurious trips during normal plant evolutions, and maintain similar configuration and operation between Susquehanna Units 1 and 2.
This change will increase plant safety by increasing the reliability of the Reactor Protection System. Reliability is increased by reducing the possibility of spurious RPS trips and full isolations, which in turn reduces the number of challenges to safety related systems.
NO SIGNIFICANT HAZARDS CONSIDERATION I.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, since no credit is taken for operation of the Main Steam Line Radiation - High trip in any FSAR accident analysis.
II.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated, since the Main Steam Line Radiation - High trip is not considered in any FSAR accident analysis. Also, the function of the MSLRMs is not changed by the proposed amendment.
III. The proposed amendment does not involve a significant reduction in a margin of safety. This conclusion is based on the following:
A) No credit is taken for the Main Steam Line Radiation - High trip in any FSAR accident analysis. The main steam line high radiation scram and MSIV closure are, however, used to mitigate the effects of a postulated accident that is not considered in the FSAR. NEDO-10174,
" Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor," presents analyses for both partial and complete blockages of a fuel bundle inlet. The report shows that, for a complete blockages of a bundle inlet, a main steam line high radiation trip will occur if the instrumentation trip in set at seven times background. This report also addresses the offsite doses due to a flow blockage event without any credit for the isolation and scram. These doses are calculated to be small fractions of 10CFR100 guidelines. Specifically for Susquehanna, the results of the Design Basis Control Rod Drop Accident (FSAR Table 15.4-15) are similar to the calculated releases for the flow blockage event (see attached Table 1).
Adjusting for Susquehanna specific meteorology results in further reductions of potential doses due to flow blockage events.
In all cases, the results are bounded by the calculated Design Basis radiological doses for the Loss of Coolant Accident (FSAR Table 15.6-18).
Credit for MSL high radiation isolation and scram with a
a JUN 2 31986 Page 3 SSES PLA-2654 rile R41-2 Ms. E. Adensam Trip Setpoint of seven times background will result in additional reductions in offsite doses for Susquehanna specific postulated flow blockage events. Moreover, since offsite doses would be within a small fraction of 10CFR100 guidelines without credit for the MSL Radiation - Hi6h trip, the difference in radiological consequences between the current and proposed setpoints would not be significant.
B)
The basis for the main steam line radiation detectors is to provide prompt detection of gross failure cf the fuel cladding. PP&L calculation EA-1-NA-002 (Attachment 1, previously transmitted under PLA-2231) demonstrates that, for Susquehanna Unit 1 8X8 fuel, the proposed seven times background trip setpoint will satisfy the basis for the detectors. Attachment 2 expands calculation EA-1-NA-002 to include both Susquehanna Unit 2 8X8 and 9X9 fuel. Attachment 2 demonstrates that, for 8X8 fuel, cladding failure of at least 15 fuel rods will be needed to initiate a three times background radiation RPS trip and full isolation, while failure of at least 44 rods would be needed to reach a seven times background trip setpoint.
For 9X9 fuel, 19 rods must fail to reach three times background, and 56 rods must fail to reach seven times background. For both trip setpoints, the amount of cladding failure necessary for initiation represents a very small fraction of the total number of rods in the core (i.e.,
0.032% to reach three times background, and 0.093% to reach seven times background; the percentages are the same for either 8X8 or 9X9 fuel bundles).
The radiological consequences of the increased number of cladding failures necessary to initiate the proposed seven times background trip are not significant, especially when compared to the radiological consequences of the Control Rod Drop Accident (CRDA).
The CRDA radiological evaluations are based on the assumed failure of 770 fuel rods. The offsite exposures of the Design Basis CRDA, presented in Table 1, are a fraction of 10CFR100 guidelines.
The potential increased offsite doses associated with the proposed seven times background trip can be considered to be a very small fraction of the offsite doses calculated for the CRDA. Thus, based on the ability of the proposed seven times background setpoint to promptly detect gross fuel failure and the small incremental increase in potential offsite doses, the proposed setpoint change does not involve a significant reduction in any margin of safety.
Page 4 SSES PLA-2654 JUN 231993 File R41-2 Hs_. E. Adensam Requests for additional information may be directed to Mr. L. M. Olson (215) 770-7859. Pursuant to 10CFR170.21, PP&L check No. 569-3672 (dated April 4, 1985) should be credited for the application fee.
Very truly yours,
() /
H. W.
ei er Vice President-Nuclear Operatians Attachments cc:
M. J. Campagnone USNRC L. R. Plisco USNRC T. M. Cerusky, Director Bureau of Radiation Protection PA Dept. of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120 J
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