ML13331A930
| ML13331A930 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/19/1987 |
| From: | Medford M Southern California Edison Co |
| To: | NRC Office of Administration & Resources Management (ARM) |
| References | |
| TAC-65229 NUDOCS 8705260032 | |
| Download: ML13331A930 (34) | |
Text
REGULATORY @
4ORMATION DISTRIBUTION SYS*1 (RIDS)
ACCESSION NBR:8705260032 DOC.DATE: 87/05/19 NOTARIZED: NO DOCKET #
FACIL:50-206 San Onofre Nuclear Station, Unit 1, Southern Californ 05000206 AUTH.NAME AUTHOR AFFILIATION MEDFORDM.O.
Southern California Edison Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards addl info re modeling error in main feedline break transient analysissubmitted in support of Amend 97 to License DPR-13.Results of original calculation confirmed &
acceptance criteria met.Calculation & analysis also encl.
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Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M.O.MEDFORD TELEPHONE MANAGER OF NUCLEAR ENGINEERING (818) 302-1749 AND LICENSING May 19, 1987 U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Pressure Transmitter PT-459 Failure San Onofre Nuclear Generating Station Unit 1 By letter dated April 30, 1987, the NRC staff was informed of a modeling error in the Main Feedline Break transient analysis which was submitted in support of Amendment No. 97 to the San Onofre Unit 1 License. It was also indicated that because of design features not previously credited in the analysis, the event does in fact meet acceptance criteria when the modeling error is removed from the analysis. It was also indicated that when including the additional design features in the analysis, a new event which had not been previously analyzed must now be analyzed. This new event was the Feedline Break upstream of the in-containment check valves.
This event also meets acceptance criteria.
During subsequent telephone discussions, the NRC staff requested that additional information regarding this subject be submitted for their review. The purpose of this letter is to provide the requested information. provides the conservative calculation performed to determine the containment pressure response following a Main Feedline Break inside containment. The calculation has been revised since the submittal of April 30, 1987. Revisions were made to more appropriately model the thermodynamic behavior of the steam/water/air mixture inside containment. In addition, an appendix has been added which evaluates the sensitivity of the containment pressure response to the blowdown modeling. These additional efforts confirm the results of the original calculation. All trips occur earlier than the 14 second trip used in the previously submitted analysis and are well within the 44 seconds which are available for reactor trip after a feedwater line break downstream of the in-containment check valves as indicated in the submittal of April 30, 1987.
o 8705260032 870519 PDR ADOCK 05000206 P
Document Control Desk May 19, 1987 One important conservatism in this calculation is the assumption that for the base case (Case 1) less than 35,000 pounds of steam/water mixture from the affected steam generator are discharged into containment. In fact, more than 50,000 pounds are discharged into containment from all three steam generators in 44 seconds. In addition, the calculation neglects the contribution to the mass and energy release from the feedwater system through the break. The main feedwater system would continue operating after the break because it is designed to function under similar conditions in safety injection service. The main feed pumps have a dual service at San Onofre Unit 1. They provide main feedwater to the steam generators (normal service) and safety injection to the reactor coolant system (emergency service).
For safety injection (SI) service, the pumps deliver up to 21,000 gpm through the safety injection flow path to the three depressurized reactor coolant system loops.
For feedwater service, the pumps deliver up to 14,000 gpm through the first point feedwater heaters, header isolation valves, and modulating flow control valves to the three pressurized steam generators. As the flow resistance of the condensate/feedwater flow path is larger than that of the SI flow path, the main feed pump flow rate to depressurized steam generators would be expected to be bounded by the safety injection flow rate.
Since four 1/3 capacity condensate pumps and two comparably sized heater drain pumps are normally in service, it is not expected that main feed pump cavitation would occur under these conditions even considering the higher saturation pressure of condensate versus refueling water. Equipment protective trips (e.g., low suction pressure) for these dual-service pumps are not required. provides the analysis of the main Feedline Break (FLB) upstream of the in-containment check valves.
This event was analyzed because the FLB downstream of the check valves assumes credit for the high containment pressure safety injection reactor trip which may not be available within the time periods discussed above if the break were to occur inside containment and would not be available at all if the break occurs outside containment. The reactor trip occurs on high pressurizer water level at 50% and the transient response for this event is acceptable. The pressurizer level trip equipment and associated cabling connections and terminations are qualified for harsh environment conditions (e.g., LOCA, MSLB) which bound those of the feedwater line break event. Jet impingement and pipe whip effects for the feedwater line break are currently being analyzed as part of the resolution of SEP Topic 111-5, High Energy Line Break (HELB) Analysis.
However, preliminary results of the HELB analysis indicate that jet impingement and pipe whip damage to the above trip circuits will not occur.
By letter dated October 31, 1986, the NRC staff was provided with a schedule for submittal of the design descriptions and single failure analysis for the Auxiliary Feedwater System (AFWS) upgrade which will be installed at the next refueling outage. It was also indicated that the corrective measure to be implemented for the steam/feedwater flow mismatch trip would be submitted on the same schedule. The submittal date was established as May 31, 1987. These modifications are related in that the AFWS response for transient events is dependent on the reactor trip assumed as discussed in our letter dated March 11, 1987 which submitted the single failure analysis for the reactor protection system. The design objective of the AFWS upgrade is to remove the susceptibility of the system to fail to respond acceptably with an
Document Control Desk May 19, 1987 arbitrary single failure assumed concurrent with an event which results in AFWS demand. This susceptibility was recognized by the NRC in the post-TMI SERs dated September 27, 1982; October 22, 1982 and November 18, 1982. In consideration of this condition in the AFWS, SCE committed to provide the above mentioned upgrades prior to startup from the next refueling outage in accordance with the process embodied in the Integrated Living Schedule which was issued on April 20, 1987 as Amendment No. 98 to the San Onofre Unit 1 Operating License.
The newly identified single failure concerns related to the steam/feedwater flow mismatch trip will also be resolved by design changes at the next refueling outage. In the interim, the analysis submitted without credit for that trip demonstrates acceptable results whether or not the mismatch trip is used as the arbitrary single failure if the relaxation of the single failure criterion which was applied previously to the AFWS for the Main Feedline Break cases, continues to be applied until the next refueling outage. In order to incorporate the most recently identified reactor trip aspects into the overall design effort to upgrade the AFWS, the submittal of the design information which was scheduled for May 31, 1987 must be rescheduled to June 19, 1987.
If you have any questions or desire additional information regarding this subject, please contact me.
Very truly yours, Enclosures cc:
R. F. Dudley, NRR Project Manager, San Onofre Unit 1 J. B. Martin, Regional Administrator, NRC Region V F. R. Huey, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3
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O MAIN FEEDLINE BREAK UPSTREAM OF IN-CONTAINMENT CHECK VALVES SAN ONOFRE UNIT 1
Background
The current Main Feedline Break (FLB) event analyzed for SONGS 1 assumed that the Variable Low Pressure (VLP) reactor trip would provide protection in the event that the Steam Flow/Feed Flow Mismatch reactor trip was unavailable. It was recently determined that the incorrect logic for the VLP reactor trip was modeled. The current FLB analysis assumed that a reactor trip would be generated when one loop reached the FLP trip setpoint. However, the correct logic is to generate a reactor trip on VLP when two loops reach the trip setpoint. When the correct logic is modeled, reactor trip would be delayed beyond the trip time of the current analysis.
The current FLB event analyzed for SONGS-1 assumed a break in which all three steam generators would depressurize and would contribute to the release through the break. This is due to the absence of main steamline isolation valves (MSIVs) for SONGS 1. In the process of resolving the modeling mistake, it was determined that credit had not been taken for the newly installed check valves which had been placed in the main feedwater lines (MFW) inside containment. The pre-existing MFW check valves are located outside containment. As such, two scenarios of FLBs can be postulated.
Scenario 1:
This scenario consists of FLB located between the first and second MFW check valve. The steam generators would remain pressurized and no blowdown of the steam generators would occur due to check valve located inside containment.
Auxiliary feedwater (AFW) would be available to two steam generators.
Scenario 2:
This scenario involves a FLB located inside containment between the second MFW check valve and the steam generator.
All steam generators would depressurize and would contribute to the release through the rupture since there are no MSIVs.
The Scenario 1 FLB is analyzed in this report. The Scenario 2 FLB is bounded by the previous FLB analysis since the high containment pressure safety injection reactor trip will respond at an earlier time than previously assumed.
Analysis The bounding scenario for the FLB located between the two MFW check valves assumes the unavailability of the Steam Flow/Feed Flow Mismatch reactor trip.
Protection is expected to be provided by the high pressurizer pressure, the high pressurizer water level, or the variable low pressure (assuming the 2/3
-2 loop logic) reactor trip. As in the previous FLB analysis for SONGS 1 documented in References 1 and 2, the LOFTRAN code is used to simulate the transient. The assumptions modeled in this analysis are noted below.
- 1. The plant is initially operating at 103% of rated power.
- 2. Initial reactor coolant average temperature is 40F above the nominal value (551.5 0F), and the initial pressurizer pressure is 30 psi above its nominal value (2100 psia).
- 3.
Initial pressurizer water level is 37.5% narrow range span (NRS).
- 4. Initial steam generator water level is at the nominal value.
- 5. Main feedwater to all steam generators is assumed to stop at the time the break occurs.
- 6. Pressurizer power-operated relief are available, but no credit is taken for pressurizer sprays.
- 7. High Pressurizer water level reactor trip setpoint of 50% NRS plus 4% NRS for uncertainties is assumed with a delay time of 2 seconds.
- 8. High pressurizer pressure reactor trip setpoint of 2260 psia (including uncertainties) is assumed with a delay time of 2 seconds.
- 9. Loss of reactor coolant pumps with SONGS 1 specific RCP coastdown is modeled, see description provided in Reference 1. An operating pump heat addition to the RCS of 3 MWt/pump is assumed.
- 10.
1979 ANS 5.1 Decay Heat is modeled.
- 11.
Auxiliary feedwater is assumed to be manually actuated and the system manually aligned to deliver flow of 250 gpm to two steam generators 15 minutes after the initiation of the event (time of feedline rupture).
- 12.
An AFW temperature of 100*F was assumed.
- 13. A feedwater system purge volume of 73 ft3/loop was assumed. This piping volume must be purged of the relatively hot main feedwater before the colder AFW enters the steam generators.
Results The results of the FLB located between the two MFW check valves transient are shown in Figure 1 through 9. The time sequence of events is presented in Table 1. The results show that reactor trip occurred on high pressurizer
-3 water level.
Calculations of this case show that the core remained coolable during this FLB scenario. The detailed calculations showed that the mass relieved through the pressurizer PORVs and safety valves between the time of initial relief through the PORVs and the time the PORVs reseat was not sufficient to uncover the core. The valves reseat because the heat removal capability of the AFW exceeds the core decay heat generation. As such, the ultimate acceptance criteria for a FLB event that the core remains coolable during the transient was shown to be met.
Conclusions For the Scenario 1 FLB event with the assumptions provided in the analysis section, an AFW flow of 250 gpm delivered to two steam generators initiated 15 minutes after the event (feedline rupture) is adequate to remove decay heat and to meet the ultimate acceptance criteria of a coolable geometry in the core.
References
- 1. Letter M. 0. Medford (SCE) to G. E. Lear (NRC), dated May 1, 1986
- 2. Letter Kenneth P. Baskin (SCE) to H. R. Denton (NRC), dated November 12, 1986 RO:8460F
TABLE 1 TIME SEQUENCE OF EVENTS FOR SCENARIO 1 FLB Event Time, sec.
Feedline Break between the 2 MFW check valves
- 10.
Pressurizer PORVs open (2200 psia)
- 24.
Reactor trip on high pressurizer water level
- 84.
Rods begin to drop
- 86.
Pressurizer fills 333.
Boiling begins in hot leg 477.
Pressurizer safety valves open (2500) 731.
AFW manually started of 250 gpm to 2 steam generators 910.
Peak pressurizer pressure (2558 psia) 911.
Pressurizer safety valves close 987.
Cold AFW reaches 2 steam generators 1173.
Heat removal of AFW is capable of removing core decay heat 1189.
Pressurizer PORVs close 1223.
SCE FEEDLINE BREAK PRESSURIZED 250 gpm at 15 MINUTES
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SCE FEEDLINX BREAK PRESSURIZED 250 gpwn at 15 MINUTES 7000 L4 67S.
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SCE FEEDLINE BREAK PRESSURIZED 250 gpm at 15 MINIJrhS rI A.SE-5
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