ML14176A379
| ML14176A379 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/23/1988 |
| From: | Lo R Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14176A380 | List: |
| References | |
| NUDOCS 8803100023 | |
| Download: ML14176A379 (24) | |
Text
6{AA REG&
o0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 Docket No.:
50-261 LICENSEE:
Carolina Power & Light Company (CP&L)
FACILITY:
H. B. Robinson 2
SUBJECT:
SUMMARY
OF FEBRUARY 16, 1988 MEETING ON LOSS OF COOLANT ACCIDENT (LOCA) ANALYSES FOR SAFETY INJECTION WITH SINGLE FAILURE, H. B. ROBINSON 2 Carolina Power & Light Company (CP&L), the licensee for H. B. Robinson 2, determined that under certain failures of the emergency diesel generator (EDG) voltage regulator, the automatic initiation and operation of only one of the three safety injection (SI) pumps can be assured during a LOCA coupled with a loss of offsite power (LOOP) event. On February 16, 1988, CP&L met with the NRR staff to discuss the effect of one SI pump operation on the LOCA analysis.
A copy of the attendance list, the viewgraphs CP&L used in the meeting and an evaluation performed by Westinghouse for CP&L entitled, "Justification for Startup and Operation of H. B. Robinson with altered HHSI flow Performance" are enclosed as Enclosures 1, 2, and 3, respectively.
CP&L presented the results of its evaluation assuming that the SI flow during a LOCA is reduced by 50% from that with two pump operation. Its analysis shows that for a LOCA involving a large break, the critical reactor parameter, the peak cladding temperature, will remain below 22000F, provided that a minor change in the limit of the hot channel factor is made. For a LOCA involving small breaks, the CP&L analyses show that the 10 CFR 50.46, ECCS Acceptance Criteria, can be met provided that manual startup of a second SI pump is assured 30 minutes after the start of the accident. CP&L believes that operator action to manually switch-over a second SI pump to the operable emergency diesel generator and start it within 30 minutes can be accomplished.
8803 100023 680223 PDR ADOCK 05000261 P
CP&L planned to follow up with an application package to restart H. B. Robinson 2 under restricted conditions. CP&L's submittal would include procedures for operator,
action to ensure that a second SI pump would be in operation within 30 minutes of a LOCA.
Ronnie Lo, Project Manager Project Directorate II-1 Division of Reactor Projects I/II
Enclosures:
As stated cc:
The NRC attendees Elinor G. Aoensam Leonard Loffin, CP&L DISTRIBUTION:
See attached sheet PM:PD21:DRPR D:PD21:DRPR RLo EAdensam 02/ /88 02/ /88
Mr. E. E. Utley Carolina Power & Light Company H. B. Robinson 2 cc:
Mr. R. E. Jones, General Counsel Mr. Dayne H. Brown, Chief Carolina Power & Light Company Radiation Protection Branch P. 0. Box 1551 Division of Facility Services Raleigh, North Carolina 27602 Department of Human Resources 701 Barbour Drive Raleigh, North Carolina 27603-2008 Mr. McCuen Morrell, Chairman Darlington County Board of Supervisors Mr. Robert P. Gruber County Courthouse Executive Director Darlington, South Carolina 29535 Public Staff -
NCUC P.O. Box 29520 Mr. H. A. Cole Raleigh, North Carolina 27626-0520 Special Deputy Attorney General State of North Carolina P.O. Box 629 Raleigh, North Carolina 27602 Mr. 0. E. Hollar Associate General Counsel Carolina Power and Light Company P.O. Box 1551 Raleigh, North Carolina 27602 U.S. Nuclear Regulatory Commission Resident Inspector's Office H. B. Robinson Steam Electric Plant Route 5, Box 413 Hartsvll'e, South Carolina 29550 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Suite 2900 101 Marietta Street Atlanta, Georgia 30303 Mr. R. Morgan General Manager H. B. Robinson Steam Electric Plant Post Office Box 790 Hartsville, South Carolina 29550 Mr. Avery Upchurch, Chairman Triangle J Council of Governments 100 Park Drive Post Office Box 12276 Research Triangle Park, NC 27709
- Ralegh, orthCaro 760320
ENCLOSURE 1 ATTENDEES AT FEBRUARY 16, 1988 MEETING Ronnie Lo NRR PDII-1 Jan Kozyra CP&L Licensing Ken Eccleston NRR TA -
AD RH Leonard Loflin CP&L Licensing Robert Jones NRR/SRXB Steve Varga NRR/D DRPR I/II Bart Buckley NRR/PDII-1 Wayne Hodges NRR/SRXB Gus Lainas NRR/AD RH Don Tondi NRR/SELB Peter Kang NRR/SELB N.Fields NRR/OEAB
H.B. ROBINSON NUCLEAR STATION LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS ANALYSIS ASSUMPTIONS O
HIGH HEAD SAFETY INJECTION PUMP FLOW CONSERVATIVELY REDUCED 50 PERCENT O
DOUBLE ENDED COLD LEG GUILLOTINE BREAK O
LIMITING DISCHARGE COEFFICIENT (CD = 0.4, PREVIOUSLY DETERMINED)
O 102% OF LICENSED CORE POWER LEVEL (1.02
- 2300 MWT)
O HOT CHANNEL ENTHALPY RISE = 1.65 O
PEAK LINEAR POWER = 102% OF 13.2915 KW/FT O
15 X 15 EXXON FUEL CORE ANALYTICAL MODELS O
WESTINGHOUSE 1981 ECCS EVALUATION MODEL WITH BART SATAN VI WREFLOOD COCO 81 MODEL LOCTA BART INTERIM LOCTA
H.B. ROBINSON NUCLEAR STATION LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS TIME SEQUENCE OF EVENTS O
BREAK INITIATION 0.0 Sec.
O SAFETY INJECTION SIGNAL 0.92 Sec.
O ACCUMULATOR INJECTION 15.1 Sec.
O END-OF-BYPASS 31.22 Sec.
O PUMPED SAFETY INJECTION 25.92 Sec.
O BOTTOM OF CORE RECOVERY 50.07 Sec.
0 ACCUMULATOR EMPTY 56.47 Sec.
H.B. ROBINSON NUCLEAR STATION LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS RESULTS O PEAK CLAD TEMPERATURE (PCT) 2198.5 F O PCT LOCATION 8.00 ft.
O MAX. LOCAL ZIRC/H 20 RXN.
7.14 %
O LOCATION OF MAXIMUM LOCAL ZIRC/H 20 REACTION.
5.75 ft.
O TOTAL ZIRC/H 20 REACTION
< 0.3%
O HOT ROD BURST TIME 49.0 Sec.
0 HOT ROD BURST LOCATION 5.75 ft.
I*
f;
-v UNIT 2
C.
Carlin Power
&_ Lih opayCR PESR H.
8 ROBNSO UPDATED FINAL SAFETY ANALYSIS REPORT
0 c
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CW c ;N E
(t)
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_______N NO.
H. B. ROBINSON FIGUR IE UNIT 2 Carolina Power & Light Company DEAKCLGD*4 15.6.ATU7 UPDATED FINAL DCG(D04 SAFrTY ANALYSIS REPORT
No 0E 0N0.6 M. B. ROBINSON UNIT 2 REFLOOD TRANSIENT FIGU.RE Carolna Pwer Ligt CopanyDECLG (CD a.)
Carlin Poer Liht ompnyDOWNCOMBER AND CORE 15.6.510 UPDATED FINAL WATER LEVELS SAFETV ANALYSIS REPORT
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- 40.
- 80.
120.
160.
200.
240.
280.
320.
TIME (SEC) m1.0 xa om1L
IV Cn CV(
AME.~q ND. 6 H. B. ROSINSON FGR UNIT 2ACCUMULATOR FLOW Carolina Power & Light Company (BLOWDOWN DECLG (CD 0.4) 15.6.512 UPDATED FINAL SAFETY ANALYSIS REPORT
H.B. ROBINSON NUCLEAR STATION SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSIS ANALYSIS ASSUMPTIONS O
HIGH HEAD SAFETY INJECTION FLOW FROM ONE PUMP FOR 30 MINUTES O
HIGH HEAD SAFETY INJECTION FLOW FROM TWO PUMPS AFTER 30 MINUTES (OPERATOR ACTION ASSUMED)
O SPECTRUM OF SMALL BREAKS IN THE COLD LEG
- 2-Inch, & 3-Inch Equivalent Diameter Breaks O
102% OF LICENSED CORE POWER LEVEL (1.02
- 2300 MWT) 0 HOT CHANNEL ENTHALPY RISE = 1.65 O
PEAK LINEAR POWER = 102% OF 12.938 KW/FT (Based On Limiting Top Skewed Power Shape)
O 15 X 15 EXXON FUEL CORE ANALYTICAL MODELS O
WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL INCORPORATING THE NOTRUMP CODE NOTRUMP SMALL BREAK LOCTA-IV
H.B. ROBINSON NUCLEAR STATION SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSIS TIME SEQUENCE OF EVENTS EVENT 2-INCH 3-INCH (Time in Seconds)
O BREAK INITIATION 0.0 0.0 O REACTOR TRIP SIGNAL 12.99 5.79 O S-SIGNAL GENERATED 21.70 9.85 O LOOP SEAL STEAM VENTING 1009.1 450.3 O BOILOFF CORE UNCOVERY 1683.5 798.2 O BOILOFF CORE RECOVERY 3709.8 2231.2 O MAXIMUM CORE UNCOVERY 2114.3 1182.1 0 ACCUMULATOR INJECTION N/A 1099.6 0 PEAK CLAD TEMPERATURE 2491.9 1229.9
H.B. ROBINSON NUCLEAR STATION SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSIS RESULTS EVENT 2-INCH 3-INCH O PEAK CLAD TEMPERATURE (F) 1409.1 1771.6 O PCT LOCATION (ft) 12.0 12.0 O MAXIMUM LOCAL ZIRC/H 20 0.44 2.31 REACTION O LOCATION OF MAXIMUM LOCAL 12.0 12.0 ZIRC/H 20 REACTION 0 TOTAL ZIRC WATER REACTION
< 0.3%
< 0.3%
2400.
2291 2988.
1800.
IL 4 188.
IN9e.
5ee.
1
- e.
15s.
29-9.
29-9.
T1ME ISECI B. ROBINSON UNIT 2 UPPER PLENUM PRESSURE 3-INCH COLD LEG BREAK
- 22.
To-P-0 F
- 28.
.See.
los.
15 2
- 8.
T IME ISECI H.
B.
ROBINSON UNIT 2
CORE MIXTURE LEVEL 3-INCH COLD LEG BREAK
I see.
6".
12e0t6ta 140 4.
164 mI.
Me.
ftt(
iSCOi H. B. ROBINSON UNIT 2 HOT SPOT CLAD TEMPERATURE 3-INCH COLD LEG BREAK FIGURE 15.6.2-5
62S.
6Se.-
468.
- 56.
SSW.
5-3 t-IS I.J.
C.
476.
4 Se.
T IE ISEC)
H. B. ROBINSON UNIT 2 ACCUMULATOR PRESSURE 3-INCH COLD LEG BREAK
2400.
2200.
2000.
1800 1600.
1400 S1200.
000 600 li.
600. 1000. 1600. 2000. 2500. 5000. 5500. 4000. 4600. 6000.
TIME (SEC)
H. B.
ROBINSON UNIT 2 UPPER PLENUM PRESSURE 2-INCH COLD LEG BREAK
54.
252. ------
20..
- 10.
500.
1000. 1500. 2000. 25OO 300 03. 4.000. 4500. 5000.
TIME (SEC)
H.
B.
ROBINSON UNIT 2 CORE MIXTURE LEVEL 2-INCH COLD LEG BREAK
Ie ONe 1ae 66d.
160e.
2WM.
22M.
2404.
260.
209e.
SS 9.
Sna.
Sa88.
TIME iSEC, H. B. ROBINSON UNIT 2 HOT SPOT CLAD TEMPERATURE 2-INCH COLD LEG BREAK FIGURE 15.6.2-2a
CPL-88-510 Westinghouse Power Systems NuCer' Tec"0100, Electric Corporation systerr0 s:o' February 15, 1988 NS-OPLS-OPL-II-88-092 Mr. S. F. Zimmerman, Manager Nuclear Fuel Carolina Power & Light Company P. 0. Box 1551 Raleigh, NC 27602 ATTENTION: T. Clements CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON JUSTIFICATION FOR STARTUP AND OPERATION OF H. B. ROBINSON WITH ALTERED HHSI FLOW PERFORMANCE
Dear Mr. Zimmerman:
At the request of Carolina Power & Light (CP&L), Westinghouse performed a Safety Evaluation to examine the effects of a reduction in the high head safety injection (HHSI) flow equivalent to an extended startup of one of the HHSI pumps on the loss-of-coolant accident (LOCA) analyses for the H. B. Robinson nuclear power plant.
This evaluation reports the results of the Westinghouse large break and small break LOCA emergency core cooling system (ECCS) evaluation model analyses for a plant very similar in design to H. B. Robinson Unit 2 including the representation of Advanced Nuclear Fuels Corporation 15x15 fuel parameters.
A limiting large break LOCA analysis was performed for a double ended cold leg guillotine break with a 0.4 discharge coefficient using the NRC approved Westinghouse 1981 ECCS Evaluation Model incorporating the BART analysis methodology.
The large break OCA analysis indicates that a calculated peak cladding temperature of 2198.5 F was obtained at the 8.0-foot elevation assuming a core power level corresponding to 102% of 2300 MWth for a peak linear heat generation rate of 102% of 13.2915 kW/ft with a hot channel enthalpy rise factor of 1.65, when only one high head safety injection pump is available.
Two small break LOCA analyses were also performed using the NRC approved Westinghouse small break LOCA ECCS Evaluation Model incorporating the NOTRUMP analysis methodology.
The analyses assumed a core power level corresponding to 102% of 2300 MWth at a total core peaking factor (FQT) of 2.32 with a hot channel enthalpy rise factor of 1.65.
The analysis of a 3-inch equivalent diameter break in the co;d leg resulted in the highest small break LOCA peak cladding temperature of 1771.6 F at the 12.0-foot elevation.
An analysis of a 2-inch equivalent diameter break in the cold leg was performed to assure that the reduction in the safety injection flow would not shift the size of the break resulting in the highest small break LOCA peak cladding temperature calculation.
The analysis of a 2-inch equivalent diameter break in the cold leg resulted in a small break LOCA peak cladding temperature of 1409.1 0F at the 12.0-foot elevation, when credit is taken for manual operator action to align and start the swing high head safety injection pump.
The results of the analyses and evaluations show that the H. B. Robinson Unit 2 Nuclear Power Plant may be started and operated in compliance with the requirements of 10CFR50.46 when flow from only one high head safety injection pump is initially available and manual operator action is taken to start and align flow from a second high head safety injection pump up to 30 minutes after the start of the accident.
If you have any questions, please contact the undersigned.
Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION VG..
al, ager Carolina Area &
D. L. Cecchett/dmnr Attachment cc:
L. H. Martin, (CP&L) 1L, 1A T. M. Dresser (CP&L) 1L, 1A B. G. Rieck (CP&L -
HBR) 1L, 1A B. M. Slone (CP&L -
HBR) 1L, 1A R. J. Muth (CP&L - HBR) 1L, 1A R. S. Pollock (W - Raleigh) 1L, A