Justification for Increase in Max Allowable Reactor Protection Sys High Pressure Trip Setpoint, Prepared for Met EdML19210A531 |
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Three Mile Island |
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Issue date: |
01/16/1976 |
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From: |
BABCOCK & WILCOX CO. |
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To: |
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Shared Package |
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ML19210A523 |
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References |
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NUDOCS 7910300570 |
Download: ML19210A531 (9) |
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Category:GENERAL EXTERNAL TECHNICAL REPORTS
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ML20055C9461990-03-31031 March 1990 Final Rept, TMI - Unit 2 Safety Advisory Board, for Mar 1981 - Dec 1989 ML20247R6371989-09-12012 September 1989 Submerged Demineralizer Sys ML20246H9401989-08-25025 August 1989 Corrected Pages 5-10 Through 5-18 & 5-26 to Rev 1 to Defueling Completion Rept,Second Submittal ML20246E1021989-08-18018 August 1989 Portion of Rev 1 to, Defueling Completion Rept ML20246G1161989-08-15015 August 1989 Rev 0 to SER for Removal of Metallurgical Samples from TMI-2 Reactor Vessel, Safety Analysis ML20247J2251989-07-31031 July 1989 Rev 7 to Div Sys Description for Auxiliary Bldg Emergency Liquid Cleanup Sys (Epicor II) ML20246E3481989-07-0505 July 1989 Rev 0 to TMI-2 Defueling Completion Rept ML20245K0151989-06-30030 June 1989 Amend 4 to Post-Defueling Monitored Storage Sar ML20246P8081989-05-31031 May 1989 Rev 12 to Defueling Water Cleanup Sys ML20244E3241989-04-30030 April 1989 Rev 7 to Technical Evaluation Rept 15737-2-G03-107, Waste Handling & Packaging Facility ML20235G5761989-02-0101 February 1989 Rev 0 to Criticality Safety Evaluation for Increasing TMI-2 Safe Fuel Mass Limit ML20155G5581988-10-10010 October 1988 Rev 0 to Technical Evaluation Rept for Processed Water Disposal Sys ML20154F6941988-09-0909 September 1988 Rev 0 to SER for Completion of Upper Core Support Assembly Defueling 1999-08-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K4701999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for TMI-1.With ML20211H5111999-08-31031 August 1999 Non-proprietary Rev 1 to MPR-1820(NP), TMI Nuclear Generating Station OTSG Kinetic Expansion Insp Criteria Analysis ML20211Q3551999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Tmi,Unit 1.With ML20210R4791999-08-13013 August 1999 Update 3 to Post-Defueling Monitored Storage SAR, for TMI-2 ML20210U4791999-07-31031 July 1999 Monthly Operating Rept for July 1999 for TMI-1.With ML20209G0011999-07-0909 July 1999 Staff Evaluation of Individual Plant Exam of External Events Submittal on Plant,Unit 1 ML20210K7651999-07-0909 July 1999 Rev 2 to 86-5002073-02, Summary Rept for Bwog 20% Tp Loca ML20209H8251999-07-0101 July 1999 Provides Commission with Evaluation of & Recommendations for Improvement in Processes Used in Staff Review & Approval of Applications for Transfer of Operating Licenses of TMI-1 & Pilgrim Station ML20209H1421999-06-30030 June 1999 Monthly Operating Rept for June 1999 for TMI-1.With ML20195H0751999-06-0808 June 1999 Drill 9904, 1999 Biennial Exercise for Three Mile Island ML20195H9261999-05-31031 May 1999 Monthly Operating Rept for May 1999 for TMI-1.With ML20209G0351999-05-31031 May 1999 TER on Review of TMI-1 IPEEE Submittal on High Winds,Floods & Other External Events (Hfo) ML20207B6621999-05-27027 May 1999 SER Finding That Licensee Established Acceptable Program to Periodically Verify design-basis Capability of safety-related MOVs at TMI-1 & That Util Adequately Addressed Actions Required in GL 96-05 ML20206R0571999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Tmi,Unit 1.With ML20206D4201999-04-20020 April 1999 Safety Evaluation Granting Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c for Fire Areas/Zones AB-FZ-4,CB-FA-1,FH-FZ-1,FH-FZ-6,FH-FZ-6, IPSH-FZ-1,IPSH-FZ-2,AB-FZ-3,AB-FZ-5,AB-FZ-7 & FH-FZ-2 ML20209G0071999-03-31031 March 1999 Submittal-Only Screening Review of Three Mile Island,Unit 1 Individual Plant Exam for External Events (Seismic Portion) ML20205K6851999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Tmi,Unit 1.With ML20210C0161999-03-0101 March 1999 Forwards Corrected Pp 3 of SECY-98-252.Correction Makes Changes to Footnote 3 as Directed by SRM on SECY-98-246 ML20207M8461999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for TMI-1.With ML20196K3561999-01-22022 January 1999 Safety Evaluation Concluding That Although Original Licensee Thermal Model Was Unacceptable for Ampacity Derating Assessments Revised Model Identified in 970624 Submittal Acceptable for Installed Electrical Raceway Ampacity Limits ML20207A9291998-12-31031 December 1998 1998 Annual Rept for TMI-1 & TMI-2 ML20196G4661998-12-31031 December 1998 British Energy Annual Rept & Accounts 1997/98. Prospectus of British Energy Share Offer Encl ML20196F6861998-12-0202 December 1998 Safety Evaluation Accepting Licensee Second 10-yr Interval ISI Program Plan Request for Alternative to ASME B&PV Code Section XI Requirements Re Actions to Be Taken Upon Detecting Leakage at Bolted Connection ML20198B8641998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for TMI-1.With ML20195C6921998-11-12012 November 1998 Safety Evaluation Supporting Amend 52 to License DPR-73 ML20195J8591998-11-12012 November 1998 Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan ML20196B7191998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for TMI-1.With ML20203G1211998-10-30030 October 1998 Informs Commission About Staff Preliminary Views Concerning Whether Proposed Purchase of TMI-1,by Amergen,Inc,Would Cause Commission to Know or Have Reason to Believe That License for TMI-1 Would Be Controlled by Foreign Govt ML20155E7511998-10-15015 October 1998 Rev 1 to Form NIS-1 Owners Data Rept for Isi,Rept on 1997 Outage 12R EC Exams of TMI-1 OTSG Tubing ML20154L5541998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for TMI Unit 1.With ML20153A9941998-09-16016 September 1998 Safety Evaluation Denying Request to Remove Missile Shields from Plant Design ML20151U8821998-09-0808 September 1998 SER on Revised Emergency Action Levels for Gpu Nuclear,Inc, Three Mile Island Nuclear Plant Units 1 & 2 ML20151V2811998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Tmi,Unit 1.With ML20237A8331998-08-12012 August 1998 Safety Evaluation Accepting USI A-46 Program Implementation at Plant,Unit 1 ML20237C6411998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Tmi,Unit 1 ML20236R2201998-06-30030 June 1998 Monthly Operating Rept for June 1998 for TMI-1 ML20236W9961998-06-0909 June 1998 1998 Quadrennial Simulator Certification Rept ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20249A1061998-05-31031 May 1998 Monthly Operating Rept for May 1998 for TMI-1 ML20247G0761998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Three Mile Island Nuclear Station,Unit 1 ML20212A2191998-04-22022 April 1998 Rev 3 to Gpu Nuclear Post-Defueling Monitored Storage QAP for Three Mile Island Unit 2 ML20248H6991998-04-0808 April 1998 Requests,By Negative Consent,Commission Approval of Intent to Inform Doe,Idaho Operations Ofc of Finding That Adequate Safety Basis Support Granting Exemption to 10CFR72 Seismic Design Requirement for ISFSI to Store TMI-2 Fuel Debris ML20216K1061998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Three Mile Island Nuclear Station,Unit 1 ML20217E0811998-03-24024 March 1998 Rev 0 to TR-121, TMI-1 Control Room Habitability for Max Hypothetical Accident ML20212E2291998-03-0404 March 1998 Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected ML20216F0981998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Three Mile Island Nuclear Station,Unit 1 ML20202F8121998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for TMI Nuclear Station, Unit 1 ML20199G8371998-01-22022 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Three Mile Island Nuclear Station,Unit 1 ML20198N2901998-01-12012 January 1998 Form NIS-1 Owners' Data Rept for Isi ML20199J3251997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Three Mile Island Nuclear Station,Unit 1 1999-09-30
[Table view] |
Text
JUSTIFICATION FOR THE INCREASE IN THE MAXIMUM ALLOWABLE REACTOR PROTECTION SYSTEM HIGH PRESSURE TRIP SETPOINT INTRODUCTION This report has'been prepared to justify the increase in the Reactor Protection System (RPS) high pressure trip setpoint at Three Mile Island Unit 1 from the current technical specification value of 2355 psig to that of 2h05 psig.
The safety analyses summarized in this report supplement the TMI-l FSAR Chapter 14 analyses. These analyses demonstrate that a Peactor Protection System trip setpoint of 2405 psig is acceptable and prevents the safety limit of 2750 psig from being exceeded. The differences in the analyses reported here and those in the FSAR are made possible because the key nuclear parameters affecting those accidents that result in high reactor coolant system pressure are less limiting after the beginning of Cycle 1 and are sufficiently less restrictive toward the end of Cycle 1 and into Cycle 2 that the increases in the RPS high pressure trip setpoint may be justified.
SUMMARY
AND CONCLUSION The purpose of the high pressure trip is to provide first protection against RCS overpressure transients, and to function as a backup trip for other system transients. The RCS pressure safety limit that must not be ex-ceeded in any transient is 2750 psig.
In the event of a loss of load accident the RCS pressure excursion following turbine trip is such that a high RCS pressure trip could occur if the setpoint is kept at the present level of 2355 psig. With the high pressure trip setpoint set at 2L05 psig, the probability of a reactor trip on a loss of load transient would be greatly reduced and potentially the plant availability would be increased.
The accidents reported in the FSAR which result in the most limiting RCS pressure excursions were selected for analysis. The analyses of these accidents were performed using Cycle 2 data and a 2355 psig trip setpoint, to show that the resulting RCS peak pressure is lower than for the corresponding Cycle 1 transient, due to the more favorable Doppler and moderator coefficients characterizing Cycle 2. The analyses for Cycle 2 and a 2405 psig setpoint shew that even in these conditions the peak pressure is well below 2750 psig and, in all accidents but one, below first cycle peak pressure.
_1_ 7910300[7O 1487 328
Sensitivity studies were performed for the most severe pressure transient previounly analyzed on Doppler and moderator coefficients, which are the parameters cost responsible for the differences in peak pressure between Cycle 1 and Cycle 2.
These sensitivity studies cover both Cycle 1 and Cycle 2 parameters. Furthermore, they bound the end of first cycle, making the results of the study applicable to end of Cycle 1 conditions. l The results of this analysis support the conclusion that the high pressure trip setpoint can be safely increased up to 2405 psig at the end of first cycle.
ANALYSIS The purpose of tha high pressure trip is to provide protection against RCS I
~
pressure transients, in order to prevent tha RCS pressure from reaching the safety limit of 2750 psig. Chapter 14 of the FSAR shows that for all accidents analyzed, i
a high pressure trip setpoint of 2355 psig will provide protection against exceeding ;
the safecy limit. i f
The results of these analyses show that even increasing the setpoint to 2405 :
psig at the end of Cycle 1 and during Cycle 2 the RCS pressure is still prevented froc reaching the safety limit of 2750 psig for all possible transients. Of all the accidents analyzed in the FSAR, the startup accident and the rod ejection accident from 10 -3 F.P. result in the highest peak RCS pressure. In addition to these two accidents, the feedwater line break accident is analyzed. The analysis of these accidents has been performed at the beginning of the second cycle, although the changes in the RpS trip setpoint may be made at the end of the first cycle, because:
- a. For operation during any riven cycle, the three accidents are most severe at the beginning of that cycle, when the Doppler and moderator coefficients are the most positive for that cycle.
- b. The moderator temperature coefficient is much less negative at the beginning of Cycle 2 than at the end of Cycle 1, as shown in Tabic 1.
The less negative coefficient will result in calculating a higher RCS ,
transient peak pressure.
- c. Although the Doppler coefficient is slightly more negative at the beginning of Cycle 2 than at the end of Cycle 1, as shown in Table 1, and the less negative coefficient would result in calculating a higher RCS transient peak pressure, sensitivity studies have shown that the differences in pre;sure are small co; pared to tho e resulting from the changcc in moderator coefficient.
1487 129 i
Therefore, the analyses performed at the beginning of the second cycle, with appropriate sensitivity studies on Doppler and moderator coefficients on the most severe FSAR troasient, will bound also the end of the first cycle.
The analyses were performed with CADD I) , which is the digital point kinetics co_puter code presently used to analyze RCS transients. This is an improved version I
of KAPPB, which is the code used for the TMI-l FSAR analyses. For all the three i accidents, the highest peak precsura was calculated for Cycle 2, BOL conditions, using a high pressure setpaint of 2355 psig. The same CADD cases were rerun with Cycle 1, BOL Doppler and moderator coefficients. The Cycle 2 cases were rerun !
I with setpoint of 2405 psig. For all cases, peak pressure was found to be well ;
below 2750 psig.
Startup Accident Figure 1 shows the results of the analysis. Peak RCS pressure occurs for the highest withdrawal rate resulting in a high pressure trip. Figure 1 shows that in going from Cycle 1 to Cycle 2 moderator and Doppler coefficients, RCS peak pressure is reduced from 2733 psia to 2562 psia. Raising the setpoint to 2405 psig, RCS peak pressure increases only about 15 psi for all withdrawal rates resulting in a high pressure *. rip. The reason for such a small increase in peak pressure is that in all the transients terminated by the high pressure trip, the trip occurs about 20 to 30 seconds after the beginning of the accident. However, at the time of trip, RCS pressure is raising so rapidly that the increase of the setpoint to 2405 psig only delays the trip by about 0.5 seconds. The energy added in that half second is very small as compared _to the energy already added to the system in the previous 20 to 30 seconds.
This accident results in the most severe pressure transient of all the FSAR i i
accidents. Furthermore, in this accident, the system takes a long time to trip as compared, for example, with the feedwater line break accident, so that this is the l accident most sensitive to moderator and Doppler coefficient variation. Therefore,th(
sensitivity studies on Doppler and moderator were performed for this accident.
The results are shown in Figures 2 and 3. Both sensitivity studies cover both BOL and EOL conditions for Cycle 1 and Cycle 2 (see Table 1). In particular, the results show that the BOL, Cycle 2 results bound the E0L, Cycle 1 results. Going from BOL, Cycle 2 coefficients to EOL Cycle 1 coefficients would cause an increase in peak pressure of about 14 psi due to the Doppler feedback, a decrease of about 32 psi due to the moderator temperature feedback, and thus an overall reduction of about 18 psi.
Chanses in other parameters between EOL first Cycle and BOL second cycle will not substantially affect these results.
1487 330
Rod Ejection Accident Figure 4 shows the results of his analysis. The peak pressure response in this analysis is similar to the one for the startup accident. Increasing the setpoint to 2405 psig results in less than 15 psi increase in peak pressure for all ejected rod worths resulting in a high pressure trip. All results are wall below the 2750 psig safety limit.
Feedwater Line Break Accident This accident was simulated through a conservative forcing function for the steam generator heat demand following the break. The results of the analysis are shown in Table 2. Trip occurs so rapidly that peak pressures for Cycle 1 and Cycle 2 are almost identical. Increasing the trip setpoint to 2405 psig, the peak pressure increases to 2655 psia, still well below 2750 psig.
(1) "CADD - Computer Application to Direct Digital Simulation of Transients in Water Reactors," BAW-10080, Rev. 1, Babcock & Wilcox, October 1974.
148.7 331 TABLE 1 Doppler and Moderator Temperature Coefficients (ok/k/F)
FSAR Cycle 2
-5 Doppler BOL -1.17x10~ -1.49x10 EOL -1.33x10~ -1.53x10~
~4 Moderator BOL +0.5x10~ -1.06x10
~4 EOL -3.0x10~ -2.63x10 1487 332
TABLE 2 Feedwater Line Break Accident Results High Pressure Trip Case Setpoint, psig Pressure, psia Hip Time, s
- 1. Cycle 1 2355 2637 5.15
- 2. Cycle 2 2355 2633 5.175
- 3. Cycle 2 2405 2655 5.65 1487 333
800 !
Trip on Trip on !
_ Pressure : Flux _ l 2700 Cycle 1, 2355 psig Setpo nt Trip on Trip on g Pressure Flux
- 2600 l ' ~
Cycle 2, 2405 psig Setpoint 3
- / -
2 2500 Cycle 2, 2355 psig Setpoint b
u 2400 !
2300 2200 i e i i e i i i e i , , e i l 10-3 10 4 10-5 Rod Witt1drawal Rate (ak/k/Sec)
PE AK RCS PRESSURE VERSUS #1THDR tVL RATE FOR A STARTUP ACCIDENT FROM SUBCRITICAL 1487 334
l i
i l l I I I 2640 2630 !
h 32620 \ BOL, Cycle I b - \ -
. 2610 \
b
\ EOL, Cycle I
~ BOL, Cycle 2 b
2500 \
c" - N E0L, Cycle 2-2580 w
2510 2500 1 I I i i I
0.9 -1.0 1.1 1.2 1.3 1.4 1.5 -1.6 5
Doppler Coefficient (Ak/k/F x 10 )
l i
, PEA" RCS PRESSURE VERSUS DOPPLER COEFFICIENT FOR A STARTUP ACCIDENT.
USING FOR EACH C3 EFFICIENT THE WITHORA1: L RATE THAT RESULTS IN THE HIGHEST RCS PEAK PRESSURE Figure 2 1487 335
i i i i I i i i i i "
l I I I I i 2700 Z
C n -
" BOL, Cycle 1 2650 l 2 5 r :
5 -
2 2600 y EOL, Cycle 1 E0L, Cycle 2 E
2550 ,_ BOL, Cycle 2 -
2 d
2500 2 Z
1 I I i I I I I I I I I I I I I I -
3.0 2.0 1.0 0.0 +1.0 4 1 Moderator Coefficient (Ak/k/F x 10 )
g PEAK RCS PRESSURE VERSUS RODERATOR COEFFICIENT FOR A STARTUP ACC10ENT ECTED US ING FOR E!.CH C0E F F ICIENT TliE WITHORtWAL RATE THAT RESULTS IN THE HIGHEST RCS PEAK PRESSURE gure 4 Figure 3 1487 336