ML20212K797
| ML20212K797 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/05/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20212K772 | List: |
| References | |
| 3289F, NUDOCS 8703090419 | |
| Download: ML20212K797 (7) | |
Text
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f-METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTP.IC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.154 i
This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1.
As a part of this request, proposed replacement pages for Appendix A are also included.
GPU NUCLEAR CORPORATION hhk h MAddd BY:
Vig'e Presipt g Director, TMI-l Sworn and Subscribed to before ire this M dayofy/Jd24MLp, 1987.
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Notary PubfiTc SMARGid P. D20WH, WOTARf PU8t!C MIDD' T07:1 BC:lo, DAUPHIN CJUf'TY J
EY C0sE!3!10P. EXPIRE 3 JUNE 12. IlJe l'entet Fennsylvania Assoelsuas of We,fstin 8703090419 870305
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PDR ADOCK 05000289 P
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'l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 GPU NUCLEAR CORPORATION This is to certify that a copy of Technical Specification Change Request No.154 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:
Mr. Jay H. Kopp, Chairman Mr. Frederick S. Rice, Chairman Board of Supervisors of Board of County Comissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 Mr. Thomas Gerusky, Director PA. Dept. of Environmental Resources Bureau of Radiation Protection P.O. Box 2063 Harrisburg,.PA 17120 GPU NUCLEAR CORPORATION
- h. k.
//8.M BY:
Vicefre81 dent pigector, TMI-1 DATE:
March 5.1987
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I.
Technical Specification Change Request No. 154 GPUN requests that the' attached revised pages and figure replace the following pages'and figure of the current TMI-1 Technical Specifications.
Replace. 2-4, 2-7,-2-8, 2-9, Figure 2.3-1, 3-28, 3-30, 4-7 9
- II.
Reasons for Change
~
This TSCR proposes to raise the reactor protection system (RPS) high reactor coolant system (RCS) pressure trip setpoint from 2300 psig to 2355 psig. Also, this TSCR proposes to raise the arming threshold for the anticipatory reactor trip on turbine trip from the current'20%
reactor power level to a level of 45% reactor power. These proposed changes will improve operational safety. This TSCR'also proposes to modify the language of the basis of the Tech. Spec. Safety Limit section concerning High RCS Pressure Trip in order to accurately
-reflect the history and meaning of the current limit.
~ III.-
Safety Evaluation and Justification of Changes In 1985, the Babcock and Wilcox Owners Group-(B&WOG), as part of the Transient Assessment Program, performed significant studies involving computer analyses and detailed evaluation of actual plant data in order to improve operational safety and performance through a reduction in the frequency of. reactor trips. The significant effort on the part of the'B&WOG resulted in two (2) reports. The two (2) reports were submitted for review to the NRC Division of Licensing's Standard and Special Project Board Chief, Mr. C. O. Thomas, on November 4, 1985.
These reports were:
(1) " Justification for Raising Setpoint for Reactor Trip on High Pressure", BAW-1890 September 1985.
(2) " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip", BAW-1893, October 1985.
The NRC has accepted the two (2) above listed reports for use via evaluations issued on April 22, 1986 and April 25, 1986, respectively.
The objectives of these studies were to demonstrate that the RPS high RCS pressure trip setpoint could be increased to 2355 psig and that the arming threshold for the anticipatory reactor trip on turbine trip could be raised abcve the 20% reactor power level without adverse effect on safety. The reports noted above demonstrate these objectives for Babcock and Wilcox (B&W) 177 fuel assembly (FA) plants.
These reports and their conclusions are therefore applicable to TMI-1 because it is of the B&W 177 FA plant type and because TMI-1 specific information was used, along with other B&W plant information, in the generation of the reports. Discussion of the two (2) studies and their findings appears below.
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,The report entitled,(" Justification for Raising Setpoint for' Reactor Trip on High Pressure" (BAW-1890), documents the results of an engineering evaluation performed to justify increasing the setpoint for
-reactor. trip ~on high RCS pressure _from its current value of 2300 psig to a.value of 2355 psig. The potential safety benefit of the change is an expected reduction in the frequency of reactor' trips.
Both' pre-1979 and post-1979 high RCS pressure trips of B&W plants were Lused for the. data base in this evaluation. The pre-1979 data base includes TMI-1 while the post-1979 data base does not include TMI-1 because the unit was not operating in-this period per NRC order.-
However,-the evaluation of and conclusion derived from both data bases are applicable to TMI-1, because the plant belongs to the same group of generic 177.FA B&W plants and has the same Technical Specifications for reactor trip on high pressure setpoint as the other B&W plants.
Of the post-1979 reactor trips evaluated in the data base, 12 of the high RCS pressure trips are categorized as slow moving transients and, if increased time margin were available, are identified as having good potential for avoidance through operator action.
In addition, if the
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power level for. arming the anticipatory reactor trip (ART) on turbine -
trip were raised to a value above 40%, turbine trips' occurring at power levels less than 40% would no longer result in reactor trips. This contributes a reduction of_an additional 12 trips in the data base.
The combination of 24 trips being avoided in the four-year period post-1979 of all B&W operating plants constitutes a 13% reduction in the average trip frequency. Assuming a future mix of transients similar to that of the four-year period data base, it is projected that a 10% reduction in average trip frequency would be realized. This conclusion is applicable to TMI-1.
The evaluation explored the impact on the opening frequency of the PORV and has concluded _that raising the high pressure trip setpoint by 55 psi will have a negligible impact-on the PORV opening frequency. The total PORV opening frequency is dominated by the frequency of opening due to operator actions under Abnormal Transient Operating Guidelines (ATOG). The additional PORV opening frequency for overpressure transients is 1.86 x 10 openings per reactor year with a high RCS pressure trip setpoint of 2355 psig and a PORV opening setpoint of 2450 psig. This is a negligible additional contribution to the opening frequency in light of the contribution of other causes (e.g. operator action in conducting AT0G).
Since TMI-l's PORV belongs to the same type of valves evaluated in this study and since the TMI plant specific Abnormal Transient Procedures (ATPs) are also based on ATOG, the same (negligible) contribution to PORV opening frequency by raising the high RCS pressure trip setpoint can be concluded.
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The evaluation used a statistical evaluation of instrumentation errors with the trip setpoint. Data input from seven (7) B&W plants are listed in the report as follows:
TABLE 4-1 of BAW-1890 Plant RPS Settings for High Pressure Reactor Trip Oconee Rancho Seco 1,2 & 3 ANO-1 CR-3 DB-1 Trip setpoint (psig) 2290 2290 2298 2296 2285 Max expected string error at bistable during weekly checks 2.4 1.55 2.0 2.96 4.0
(+/- psi)
(RMS)
(RMS)
(RMS)
(Algebraic Sum)
TMI-1 has an RPS high RCS pressure trip setpoint set at 2294 psig. The maximum expected string error is t 3.52 psig. These settings are comparable to that used in the evaluation, therefore, the conclusion of the evaluation can be applied to TMI-1.
The report entitled, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip" (BAW-1983), documents the results of an engineering evaluation performed to justify raising the arming threshold for anticipatory reactor trip on turbine trip above the current 20% reactor power level. The potential safety benefit of this change is an expected reduction of at least 5% in reactor trip frequency.
Factors affecting the runback capability are evaluated in this study, and it is concluded that the initial power level, total bypass steam flow and moderator coefficient primarily determine the runback capability, while the ICS runback rate and pressurizer spray capacity have a minor impact on the overall results.
This study concluded that for the most conservative (i.e., least moderater coefficient feedback) assumption of beginning-of-cycle (B0C) condition, the maximum initial power level for successful runback is constrained mainly by the total bypass steam flow capacity.
This capacity includes the flow through turbine bypass valves, atmospheric dump valves and the first bank of main steam safety valves (MSSVs) (for TMI-1, this includes those valves set at 1040 and 1050 psig). The total steam bypass capacities prior to the RCS reaching the 2355 psig reactor trip setpoint for TMI-1 are listed below:
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Available Total Bypass Capacities Prior to Reaching'2355 psig Trip Turbine Bypass Atmospheric Dump First Bank MSSVs
' (% Full Power Steam Flow)
(% Full Power Steam Flow)
(% Full Power-Steam Flow) 22.5 6.4 48 In-the study, no credit was taken for the atmospheric dump valves.
Credit was taken for a 5% power reduction prior to reaching the high RCS pressure trip setpoint.
The maximum initial reactor power level for successful runback was then calculated by adding the 5% power reduction to the turbine bypass and MSSV steam flow percentages. The maximum initial reactor power level for TMI-1 was calculated to be 75%. However,-the maximum initial reactor power level which was applicable to all B&WOG plants was 45%. The difference in the values arises from different turbine bypass and MSSV steam flow values for the plants. To be consistent with the other B&WOG plants and to be conservative, TMI-1 will use the 45% value for the maximum initial reactor power level.
Since TMI-1 has a calculated runback capacity for turbine trip from 75%
power prior to reaching the 2355 psig high RCS pressure trip setpoint, it is concluded that the arming threshold can be raised from the current 20% to a value equal to 45% for TMI-1.
The proposed wording changes to the bases for the High.RCS Pressure Trip section of the TMI-1 Tech. Spec Safety Limits are intended to more clearly identify the history and meaning of the current safety limits.
The FSAR uses a high RCS pressure trip setpoint of 2390 psig for analysis purposes. This section of the Tech. Specs. does not clearly' identify this fact. The modification of this passage as proposed herein will help to clarify this point. These word changes have no adverse impact on safety.
IV.
No Significant Hazards Considerations These proposed changes are most similar to Examples (i) and (vi) of
" Amendments Not Likely'to Involve Significant Hazards Considerations",
Federal Register Volume 48, No. 67 at 14870, dated April 6, 1983.
A no significant hazards considerations determination is proposed for this TSCR because:
- 1) The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because the accident analyses contained in the TMI-1 FSAR were determined for a high RCS pressure trip setpoint of 2390 psig.
l This is well above the 2355 psig trip setpoint proposed by this i
TSCR.
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- 2) The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because, as noted above, the TMI-1 FSAR accident analyses Nound the changes proposed via this TSCR.
- 3) The proposed changes do not involve a significant reduction in a margin of safety because the proposed changes in this TSCR, raising the high pressure reactor trip setpoint and raising the arming threshold for the anticipatory reactor trip on turbine trip, will result in a negligible increase in the PORV opening frequency (1.86 x 10-5 openings per reactor year). These changes will also result in an improvement in operational safety.
This improvement is due to a reduction in the number of challenges to safety systems, fewer equipment transient cycles and a reduction in the number of reactor trips.
V.
Implementation It is requested that the Amendment authorizing these changes become effective upon issuance and be implemented within 30 days.
VI.
Amendment Fee (10CFR170.21)
Pursuant to the provisions of 10CFR170.21, attached is a check for
$150.00.
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