ML13322A323
| ML13322A323 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/01/1989 |
| From: | Trammell C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| IEIN-86-005, IEIN-86-5, NUDOCS 8906070264 | |
| Download: ML13322A323 (42) | |
Text
7-,
June 1, 1989 Docket Nos. 50-361 50-362 FACILITIES:
San Onofre Nuclear Generating Station, Unit Nos. 2 and 3 LICENSEE:
Southern California Edison Company
SUBJECT:
SUMMARY
OF MEETING HELD ON MAY 23, 1989 TO DISCUSS CAPACITY OF STEAM GENERATOR SAFETY VALVES On May 23, 1989, the NRC staff met with representatives of Southern California Company (SCE) to discuss the relief capacity of the steam generator safety valves at SONGS 2 and 3. Persons attending are identified on Enclosure 1.
Viewgraphs presented at the meeting are shown on Enclosure 2. The meeting was held pursuant to notice issued on May 15, 1989. Highlights of the meeting are summarized below.
Following issuance of NRC Information No.tice No. 86-05 on January 31, 1986 on main steam safety valves, a Westinghouse Owners Group Subcommittee was formed representing about sixteen licensees of plants fitted with Crosby safety valves to investigate why the Seabrook and Vogtle valves had low capacities.
The test program involved four valve types, eleven different springs, five different nozzle and guide ring settings and various set pressures involving some 235 tests. The test report prepared by the owner's group subcommittee will be formally transmitted to NRC in about two weeks. An advance copy was delivered at the meeting. Each licensee will need to evaluate the results for its specific plant, e.g., current ring settings, and evaluate the effect of changing ring settings as required.
As applied to SONGS 2 and 3, the currently installed valves have about 75% of the nameplate rating, whereas only 66% capacity is needed to meet the capacity requirements for the worst overpressure transient (loss of load w/no turbine bypass).
SCE plans to restore nameplate capacity by adjustments to both the guide and nozzle rings at each unit's next refueling outage (September 1989 for Unit 2; March 1990 for Unit 3).
SCE will be issuing a Licensee Event Report on this matter within 30 days.
/s/
Charles M. Trammell, Senior Project Manager Project Directorate V Division of Reactors Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. Attendance List
- 2. Viewgraphs C. C, cc w/enclosures See next page
SUMMARY
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 June 1, 1989 Docket Nos. 50-361 50-362 FACILITIES:
San Onofre Nuclear Generating Station, Unit Nos. 2 and 3 LICENSEE:
Southern California Edison Company
SUBJECT:
SUMMARY
OF MEETING HELD ON MAY 23, 1989 TO DISCUSS CAPACITY OF STEAM GENERATOR SAFETY VALVES On May 23, 1989, the NRC staff met with representatives of Southern California Company (SCE) to discuss the relief capacity of the steam generator safety valves at SONGS.2 and 3. Persons attending are identified on Enclosure 1.
Viewgraphs presented at the meeting are shown on Enclosure 2. The meeting was held pursuant to notice issued on May 15, 1989.
Highlights of the meeting are summarized below.
Following issuance of NRC Information Notice No. 86-05 on January 31, 1986 on main steam safety valves, a Westinghouse Owners Group Subcommittee was formed representing about sixteen licensees of plants titted with Crosby safety valves to investigate why the Seabrook and Vogtle valves had low capacities.
The test program involved four valve types, eleven different springs, five different nozzle and guide ring settings and various set pressures involving some 235 tests. The test report prepared by the owner's group subcommittee will be formally transmitted to NRC in about two weeks.
An advance copy was delivered at the meeting. Each licensee will need to evaluate the results for its specific plant, e.g., current ring settings, and evaluate the effect of changing ring settings as required.
As applied to SONGS 2 and 3, the currently installed valves have about 75% of the nameplate rating, whereas only 66% capacity is needed to meet the capacity requirements for the worst overpressure transient (loss of load w/no turbine bypass).
SCE plans to restore nameplate capacity by adjustments to both the guide and nozzle rings at each unit's next refueling outage (September 1989 for Unit 2; March 1990 for Unit 3).
SCE will be issuing a Licensee Event Report on this matter within 30 days.
Charles M. Trammell, Senior Project Manager Project Directorate V Division of Reactors Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. Attendance List
- 2. Viewgraphs cc w/enclosures See next page
Mr. Kenneth P. Baskin San Onofre Nuclear Generating Southern California Edison Company Station, Units 2 and 3 cc:
Charles R. Kocher, Esq.
Mr. Mark Medford James A. Beoletto, Esq.
Southern California Edison Company Southern California Edison Company 2244 Walnut Grove Avenue 2244 Walnut Grove Avenue P. 0. Box 800 P. 0. Box 800 Rosemead, California 91770 Rosemead, California 91770 Mr. Robert G. Lacy Orrick, Herrington & Sutcliffe Manager, Nuclear Department ATTN:
David R. Pigott, Esq.
San Diego Gas & Electric Company 600 Montgomery Street P. 0. Box 1831 San Francisco, California 94111 San Diego, California 92112 Alan R. Watts, Esq.
Mr. Paul Szalinski, Chief Rourke & Woodruff Radiological Health Branch 701 S. Parker St. No. 7000 State Department of Health Services Orange, California 92668-4702 714 P Street, Building #8 Sacramento, California 95814 Mr. Sherwin Harris Resource Project Manager Resident Inspector, San Onofre NPS Public Utilities Department c/o U.S. Nuclear Regulatory Commission City Hall of Riverside Post Office Box 4329 City Hall 3900 Main Street Mayor, City of San Clemente Riverside, California 92522 San Clemente, California 92672 Mr. Charles B. Brinkman Combustion Engineering, Inc.
Regional Administrator, Region V 12300 Twinbrook Parkway, Suite 330 U.S. Nuclear Regulatory Commission Rockville, Maryland 20852 1450 Maria Lane/Suite 210 Walnut Creek, California 94596 Mr. Dennis F. Kirsh U.S. Nuclear Regulatory Commission Chairman, Board Supervisors Region V San Diego County 1450 Maria Lane, Suite 210 1600 Pacific Highway, Room 335 Walnut Crekk, California 94596 San Diego, California 92101 Mr. Don Womeldorf Chief Environmental Mangement Branch California Department of Health 714 P Street, Room 616 Sacramento, California 95814 (14)
ENCLOSURE 1 Attendance List May 23, 1989 Meeting NRC SCE G. Knighton F. Nandy C. Trammell M. Kerschthal G. Hammer A. Sistos P. T. Kuo T. McLeod K. Desai J. Bradfute S. Juergens F. Cherny Other T. Hicks, Southern Technical Services
ENCLOSURE 2 Main Steam Safety Valve San Onofre Nuclear Generating Station Units 2 and 3 AGENDA I.
Introduction II.
Background
III.
Westinghouse Owner's Group Subcommittee on Main Steam Safety Valves A.
Objective B.
Test Program C
Safety Valve Computer Model D.
Extended Blowdown Analysis IV.
SONGS 2 and 3 Assessment A.
Main Steam Safety Valve Configuration and Data B.
Design Basis C
SONGS 2 Trip D.
Recent Secondary System Pressure Evaluation V.
Long Term Modifications VI.
Technical Specification Changes VII.
Conclusions
BACKGROUND i2 Seabrook MSSVs were high flow tested 4at Wylie laboratories and found to have low lifts. (Testing was performed in 1984 and 1985) 5 Plant Vogtle MSSVs were high flow tested at Wylie laboratories and found to have low lifts. (Testing was performed in May, 1986)
In both instances, the large positive guide ring settings were changed to a negative setting Information Notice No. 86-05 was issued January 31, 1986.
Supplement 1 to IEN 86-05 was issued October 16, 1986 On August 12, 1986, SONGS 2 experienced a spurious MSIS challenging the secondary overpressure protection system.
All MSSVs actuated.
No design parameters were exceeded.
SCE requested Crosby to evaluate the adequacy of the SONGS 2 & 3 MSSVs in November 1986 SCE joined the Westinghouse Owner's Group MSSV Subcommittee in January 1987
WESTINGHOUSE OWNERS GROUP MSSV SUBCOMMITTEE Subcommittee formed in September 1986 to address the issues raised by Information Notice 86-05 Subcommittee Members:
Several Licensees Crosby Valve & Gage Company EPRI Continuum Dynamics
SUBCOMMITTEE OBJECTIVES Formed to determine the root cause for the Inadequate capacities of the Seabrook and Vogtle MSSVs Establish "generic" ring settings to provide rated capacity at 3%
accumulation and a maximum 10% blowdown Determine the effects of spring rate and ring settings on blowdown and accumulation Develop an analytical model which can predict valve performance:
Blowdown Lift Given:
Ring Settings Geometry Spring rate Accumulation
MSSV INTERNAL I
TEST PROGRAM A matrix of high flow tests were performed at Crosby Valve Gage Company Test procedure was written by Crosby and approved by the Subcomittee The following MSSV transient data was recorded for each test:
Inlet pressure Outlet pressure Valve lift Test data was copied to a computer Test data was sent to Continuum Dynamics for development and verification of the COUPLE Code Analysis of test data performed by Crosby and Continuum Dynamics
C B CR S
TBY VA LVE 0
GAGE COMPANY Rw R ENTHAM, MASS Test Report Number 4388 Page 10 Supplement 1 FIGURE 2 AMBIENT VALVE LIFT TEMPERATURE (LVDT)
T/C OUTLET PRESSURE 12" (BACK PRESSURE)
TRANSDUCER SPRING TEMPERAT.RE T/C AT MIDPOINT TO DISCHARGE SILENCER BODY TEMPERATURE T/C AT TOP SET E DRAIN ISOLATION HYDRAULIC VLEM VALVE INLET CONTROL VALVES PRESSURE (QUICK OPEN)
TRANDUCE HIGH FLOW STEAM I
HIGH FLOW TEST FACILITY SUPPLY l(300 CU. F.DRUM RECORDER r-n DRAIN L
J VISUAL READOUT
TEST PROGRAM All of the Crosby MSSVs were tested:
6R10 6Q8 6Q8x8 Eleven springs with spring rates bounding design limits were tested Five different ring settings were tested:
Nozzle Ring Guide Ring
-75
-100
-100
-75
-100
-50
-50
-75
-75
-75 Large number of set pressures were tested
C '
Y CROSBY VALVE S GAGE COMPANY WRE NTH AM, MASS Test Report Number 4388 Page 6 Supplement 1 FIGURE 1 Set Test Pressure Ring*
Number Number Valve Spring (psizj Setting Cycles Phase I 1
1 (6R10 #1)
DK 1064 1
3 2
1 (6R10 #1)
DK 1064 2
3 3
1 (6R10 #1)
DK 1064 3
3 4
1 (6R10 #1)
DK 1090 3
3 5
1 (6R10 #1)
DK 1090 2
3 6
1 (6R10 #1)
DK 1090 1
3 7
1 (6R10 #1)
DK 1115 1
3 8
1 (6R10 #1)
DK 1115 2
3 9
1 (6R10 #1)
DK 1115 3
3 10 1 (6R10 #1)
DK 1140 3
3 11 1 (6R10 #1)
DK 1140 2
3 12 1 (6R10 #1)
DK 1140 1
3 13 1 (6R10 #1)
DK + 10%
1140 1
3 14 1 (6R10 #1)
DK + 10%
1140 2
3 15 1 (6R10 J/1)
DK + 10%
1140 3
3 16 1 (6R10 #1)
DK + 10%
1115 3
3 17 1 (6R10 #1)
DK + 10%
1115 2
.3 18 1 (6R10 #1)
DK + 10%
1115 1
3 19 1 (6R10 #1)
DK + 10%
1090 1
3 20 1 (6R10 #1)
DK + 10%
1090 2
3 21 1 (6R10 #1)
DK + 10%
1090 3
3 22 1 (6R10 #1)
DK + 10%
1064 3
3 23 1 (6R10 #1)
DK + 10%
1064 2
3 24 1 (6R10 #1)
DK + 10%
1064 1
3 25 2 (6R1O #2)
DK 1140 3
3 26 3 (6R10 #3)
DK + 10%
1064 3
3 27 4 (6R10 #14)
DK 1140 3
3 28 2 (6R10 #2)
DK + 10%
1064 3
3 29 3 (6R10 #3)
DK 1140 3
3 30 4 (6R10 #4)
DK + 10%
1064 3
3 Phase II 31 1 (6R10 #1)
EK 1170 3
3 32 1 (6R10 #1)
EK 1260 3
3 33 1 (6R10I #1)
EK + 10%
1260 3
3 34 1 (6R10 #1)
EK + 10%
1170 3
3 35 2 (6R10 #2)
EK(-0%,+10%) 1170 3
3 36 2 (6R10 #2)
EK(-0%,+10%) 1260 3
3
S r CROSBY VALVE
& GAGE COMPANY 0
WRENTH AM, MASS Test Report Number 4388 Page 7 Supplement 1 FIGURE 1 Set Test Pressure Ring*
Number Number Valve Spring psig)
Setting Cycles Phase II (cont) 37 3 (6R10 #3)
EK(-0%,+10%) 1260 3
3 38 3 (6RI
- 3)
EK(-0%,+10%) 1170 3
3 Phase III 39 5 (6Q8 pl)
BK 1050 4
3 40 5 (6Q8 #1)
BK 1050 5
3 41 5 (6Q8 #1)
BK 1105 5
3 42 5 (6Q8 #1)
BK 1105 4
3 43 5 (6Q8 #1)
BK + 10%
1105 4
3 44 5 (6Q8 #1)
BK + 10%
1105 5
3 45 5 (6Q8 #/1)
BK + 10%
1050 5
3 46 6 (6Q8 #1)
BK + 10%
1050 4
3 47 6 (6Q8 #2)
BK(-0%,+10%) 1050 4
3 48 6 (6Q8 02)
BK(-0%,+10%) 1105
- 4.
3 Phase IV 49 6 (6Q8 #2)
AK 1035 4
3 50 6 (6Q8 #2)
AK 985 4
3 51 6 (6Q8 02)
AK + 10%
985 4
3 52 6 (6Q8 # 2)
AK + 10%
1035 4
3 53 5 (6Q8 #1)
AK(-0%,+10%) 1035 4
3 54 5 (6Q8 i1
)
AK(-0%,+10%) 985 4
3 Phase V 55 5 (6Q8 #1)
CK 1175 4
3 56 5 (6Q8 #1)
CK 1190 4
3 57 5 (6Q8 #1)
CK + 10%
1190 4
3 58 5 (6Q8 #1)
CK + 10%
1175 4
3 59 6 (6Q8 #2)
CK(-0%,+10%) 1175 4
3 60 6 (6Q8 #2)
CK(-0%,+10%) 1190 4
3 Phase VI 61 7 (6R8x8 #/l)
EK 1235 3
3 62 7 (6R8x8 i1)
EK 1185 3
3 63 8 (6R8x8.?2)
EK + 10%
1185 3
3 64 8 (6R8x8 #/2)
EK + 10%
1235 3
3 Phase VII 65 9 (6Q8x8 #1)
CK 1200 4
3 66 9 (6Q8x8 #1)
CK 1130 4
3 67 10 (6Q8x8 #2)
CK + 10%
1130 4
3 68 10 (6Q8x8 #2)
CK + 10%
1200 4
3 Phase VIII 69 2 (6R10 #2)
FK(-0%,+10%)
985 3
3 70 2 (6R10 J/2)
FK(-0%,+10%) 1025 3
3 71 1 (6R10 1l)
FK(-O%,+10%) 1025 3
3 72 1 (6R10 /1)
FK 985 3
3
TEST PROGRAM Test results plotted lift and blowdown:
Nondimensional lift (lift/bore, L/D) plotted vs. nondimensional pressure (opening pressure*bore/spring rate, D*Pop/K-rate)
Blowdown plotted vs. nondimensional pressure
WOO Test Results Phase 1. II. VI & Vill. -100/-50 0.340 0.320 -.....
0.300 0.280 0.260 0.240 -
O co aD 0D 0
to 0.220 0
00 0.200 0.180 0.160 0.140 0.120 0
0.100 -
0 0.080 0.185 0.195 0.205 0.215 0.225 Pop
- Bore / K-Rate
WOG Test Results Phase 1. II. VI & VIII. -100/-50 19.0%
18.0%
17.0%
16.0%
O 15.0%
14.0%
13.0%
12.0%
c 11.0%
o 10.0%
O 9.0%
000 8.0% -
0DO 0
0 0
0 7.0%
0 0
0 7.0%
0 0
5.0%
0 00 5.0% -
DO O
0 D
4.0%
3.0%
2.0% -
O O
1.0%
0.0%
O
-1.0%
0.185 0.195 0.205 0.215 0.225 Pop
- Bore / K-Rate
TEST PROGRAM Additional testing performed in late 1988 to determine the cause of low lift/high blowdown test results Valve springs were also tested by Continuum Dynamics at Princeton University to determine if large eccentricities caused anamolous test results Large eccentricities can result in excessive stem to bearing friction affecting valve performance Additional testing resulted in expected MSSV lift and blowdowns One spring was found to have an exceptionally large eccentricity
e0 CROSBY VALVE S GAGE COMPANY CROSBY WRE NTHAM, MASS Test Report Number 4388 Page 8 Supplement(1, Revision 1 FIGURE 1 Set Test Pressure Ring*
Number Number Valve Spring Psig)
Setting Cycles Additional 73 7 (6R8x8 #1)
EK + 10%
1185 3
3 Tests 74 7 (6R8x8 #1)
EK + 10%
1235 3
3 75 8 (6R8x8 #2)
EK 1185 3
3 76 8 (6R8x8 J#2)
EK 1235 3
3 77 7 (6R8x8 #1)
EK(-0%,+10%) 1170 3
3 78 8 (6R8x8 #2)
EK(-0%,+10%) 1170 3
3
TEST PROGRAM Meeting conducted with the NRC on March 8, 1988 to discuss the program Crosby Test Report 4388, rev. 1 was issued Nov. 30, 1988 Crosby Test Report 4388, Supplement 1, was issued Feb. 2, 1989 Supplement report includes test runs 73 through 78 Continuum Dynamics Test Report and Model Report will be issued in June, 1989
TEST PROGRAM Each licensee is to utilize the results of the Owner's Group:
Evaluate the effect of the current MSSV ring settings on continued plant operation Evaluate the effect of changing ring settings (if required)
MSSV Computer Model COUPLE code developed by Continuum Dynamics and EPRI as part of the primary Safety Valve test program in early 1980's Predicts safety valve lift and blowdown given:
Ring setting Inlet pressure Spring rate Valve geometry COUPLE has been refined and verified using high flow MSSV tests performed at Crosby and 1986 Seabrook MSSV test results Code showed that even though rated capacity was not achieved at 3%
accumulation, rated capacity would be achieved at a higher accumulation
s1.5 SEABROOK U-25 TABLE 1 RUN 1 EK1 n
= -25 1.25n r-
=
+150 g
1.00 C.
C CL 0.75 LUi 0-50 LU COUPLE prediction
<2.
Data 0.25 0
0 2
4 T
1S0 12 cu-mTIME CSECJ 1
1.s0 SEABROOK U-25 TABLE 1 RUN 2 EK+10 125-nT
=
-25 n=
0 0
1.00:
F CO) 0 CL 0.75 Wi COUPLE prediction Co 0.50-Data 0-25 0
0 2
4 6
8 10 14 TIME CSEC)
MODIFICATIONS REQUIRED TO PREDICT LIMITED LIFT SEABROOK DATA When disk is below guide ring guide ring doesn't control deflection of exiting jet guide ring disk + holder
EXTENDED BLOWDOWN ANALYSIS Westinghouse prepared and issued."Analytical Report For The Effect Of Increased MSSV Blowdown" In Nov., 1988 Larger blowdowns than those used in Safety Analyses may occur as a result of implementing ring setting changes to provide stronger lifts Conservative analysis of the effects of larger blowdowns on the Loss of Load and Steam Generator Tube Rutpure Events was performed Analysis shows that increasing the blowdown to 15% and 20% does not result in exceeding any plant safety limit or 10CFR100 off site dose limits
IV.
SONGS 2 & 3 Assessment Plant Configuration & Main Steam Safety Valve Data 18 valves total, 9 per steam generator Valves are Crosby, Spring loaded, enclosed bonnet, safety valves with an "R" orfice (16 in2 orfice area)
The valves were designed, manufactured and certified in accordance with Section III of the ASME Boiler & Pressure Vessel Code, 1974 Edition.
The valves lift sequentially in pairs, the first two valves lift at 1100 psia (Main Steam Design Pressure) and the last two at 1155 psia (105% design pressure)
Figure 1 illustrates the main steam pressure relief schematic
c 0
O I
-)
mm 0
00 TO ATMDSPHERE TO ATMOSPWER 0
TO ATMOSPHERE TO ATMOSPHERE CL z
- "i Pesv F
swe.os 8***
"a 0
a, s
V esy Fsy es v
o
$418 841. 6 84 50y S3" I.D.
3 I.D.
C-)
2V" I.D.
2"I0 C) o TO "NISIIE 01.0.TO TISIME 5:~
oorun.0.
TO TURAStNE 48S S.D.
STyAM STEAM EN.N.1 *MGEN.
no.2
-AM "U
CC m
STEM LINE SAFETY VALVES PER LOOP VALVE M1ER LIFT SiTTING (O 1)'
oRIFICE SIZE Line No. 1 Line No. 2 ma.
2PSV-8401 2PSV-8410 1100 psia 16 in
- b.
2PSV-8402 2PSV-8411 1107 psia 16 In' C.
2PSV-8403 2PSV-8412 1114 psta 16 fn.
- d.
2PSV-8404 2PSV-8413 1121 psla 16 in
- e.
2PSV-8405 2PSV-8414 1128 psta 16 in'
- f.
2PSV-8406 2PSV-8415 1135 psia 16 in'
- g.
2PSV-8407 2PSV-8416 1142 psia 16 ins
- h.
2PSV-8408 2PSV-8417 1149 psia 16 In' 2PSV-8409 2Sv-as16 115 pasia 16 In'
The sizing calculation for the Main Steam Safety Valves was issued by Bechtel on 11/17/75.
The design basis for sizing the Safety Valves is as follows:
In accordance with subarticle NC-7300 of Section III of the ASME Code, the total rated releiving capacity of the Safety Valves shall be sufficient to prevent a rise in pressure of more than 10% above the design pressure under the most severe anticipated operational transient. The design pressure of the Main Steam Safety Valve is 1100 psia.
In the absence of specific FSAR transient analysis, the Safety Valves were conservatively design to handle the "valves wide open" steam flow rate of 15.1 x 106 lb/hr (100% of rated reactor power, 3410 MWt)
The valves rated capacity at their lift settings range from 818,685 to 859,646 lb/hr at 3% overpressure.
On 7/82, Combustion Engineering issued Rev. 2 of the overpressure protection report for the Nucleas Steam Supply Systems.
The Following assumptions were used:
Reactor power at 3480 (rated power plus 2% uncertainty)
Reactor does not trip on loss-of-load but will trip on high pressurizer pressure No credit for letdown, charging, pressurizer spray, secondary bypass, nor feedwater flow Safety valves lift at their maximum popping pressure The most severe anticipated transient was concluded to be a loss of turbine generator load with a delayed reactor trip.
Under this transient, the maximum steam generator pressure achieved was 1150 psia with only 16 valves lifting The results of this analysis are summarized in appendix 5.2A of the FSAR.
For more detailed evaluation of the loss-of-load transients, Appendix 5.2A refers to Section 15.2 of the FSAR.
FSAR ACCIDENT ANAYSIS Section 15.2 of the FSAR addresses loss of external load incidents.
The most severe of this incidents was determined to be a loss of condenser vacuum with a concurrent single failure The postulated failure of a pressurizer level measurement channel is considered to produce the most adverse effects following a loss of condenser vacuum FSAR Table 15.2-2 lists the assumptions used for the loss of condenser vacuum FSAR Table 15.2-5 summarizes the sequence of events and results obtained in this analysis The analysis concludes that for the most severe operational transient, the main steam peak pressure remains below 110% of design pressure
SONGS 2 TRIP On Aug. 12, 1986, a spurious Main Steam Isolation Signal (MSIS) resulted in the closure of both Main Steam Isolation Valves (MSIVs) and a Loss of Load transient Event parallels Loss of Condenser Vacuum (LOCV) event described in FSAR section 15.2 Plant response from this transient was reviewed and no design parameters were exceeded Transient was modelled using RETRAN in an effort to determine installed MSSV characteristic (i.e., flow vs. accumulation)
Fig.
2 RETRAN Model of Main Steam Safety Valve Lifting' Characteristics (MSSV Model B).
100 -
75 -I*
accumulation MSSV Flow Capacity 0
Pn Pt Pa Setpoint Pressure, psia Pn = nominal safety valve setpoint pressure, see Table 3.
Pt = safety valve lifting pressure with positive tolerance, see Table 3.
Pa = accumulation pressure
= 105% of Pt 23
Fig. 8 Stearn Generator Pressure Response to the SONGS 2 8/12/1986 MSIV Closure Event 1300-1200 1100 S1050 CL 1000-=
950-0 5
10 15 20 25 30 Time, seconds a
Plant Data (P10 13A)
Plant Data (P1023A)
MSSV Model A (RETRAN
- MSSV Model B (RETRAN
SAN ONOFRE NUCLEAR GENERATING STATION MAIN STEAM SAFETY VALVE LIFT 1 90.0%
0 1- 00.0%
RATE 9-H T-+
-+
+---
-+
-+
+
-+
I-AC UM c)0.o%
GO.0%
50.0%
40.0%
71 0
2 4
6 8
% ACCUMILATION D
UFT
+
DESIGN LIFT
SAFETY ANALYSIS EVALUATION Combustion Engineering (CE) has modelled the LOCV event using CESEC code Initial conditions have been modified to maximise peak secondary pressure Two MSSV models were used:
Linear flowrate vs. accumulation (0 % flow capacity at 0%
accumulation and 75% flow capacity, max., at 3% accumulation)
"Best estimate" model using COUPLE code and RETRAN analysis of SONGS 2 trip Additional analysis will be performed and the MSSV models will be refined
SOUTHERN CALIFORNIA EDISON COMPANY San Onofre Nuclear Generating Station Initial Conditions For The Loss Of Condenser Vacuum Analysis FSAR SONGS UNIT 2 CE PARAMETER ASSUMPTIONS TRIP ANALYSIS Initial core power level, MWt.
3,478 3,410 3,478/3,478 Core inlet coolant temperature, degrees F.
542 553 560/560 Core mass flowrate, 106 lbm/hr.
164.9 Reactor coolant system pressure, lb/in. 2a.
2,050 2,250 2,050/2,050 Steam generator pressure, lb/in.2a.
810 930 954.5/954.5 Mod rator temperature coefficient,
+0.13
-2.09
-0.7/0.0 10P % / F Steam Bypass control system Inoperative Not available Inop/Inop Reactor trip on turbine trip Inoperative Ocurred after Inop/Inop High Pzr Trip Pressurizer level control system Inoperative Operable Inop/Inop Pressurizer pressure control system Inoperative Operable IRop/Inop
SOUTHERN CALIFORNIA EDISON COMPANY San Onofre Nuclear Generating Station Sequence Of Events For The Loss Of Condenser Vacuum FSAR SONGS Unit 2 Trip CE Analysis TIME TIME TIME EVENT SECONDS VALUES SECONDS VALUES SECONDS VALUES Closure of turbine stop valves on turbine 0.0 0.0 0.0/0.0 trip due to loss of condenser vacuum.
(Unit 2 event was a spurious trip of the MSIVs)
High-piessurizer trip signal condition, 8.4 2,422 3.5 2,378 5.9/5.9 2,422/2,422 lb/in, a.
High-pressurizer trip signal condition 9.5 3.7 7.0/7.0 generated.
Pressurizer safety valves begin to open, 10.0 2,525 DID NOT REACH 8.1/8.1 2,525/2,525 1b/in. a.
PRESSURE TO LIFT Steam generator safety valves begin opening, 10.1 1,100 4.5 1,100 3.95/3.95 1,111/1,111 lb/in. a.
CEAs begin to drop into core.
10.3 6.0 8.91/8.91 Maximum core power.
10.3 103.2% OF 0.0 100%
FULL POWER Maximum RCS pressure, lb/in.2a.
12.4 2,146 1.0 2,480 8.65/8.65 2,631/2,636 Maximum pressurizer liquid volume, ft.3.
15.0 935 12.0 65% LEVEL Pressurizer safety valves closed, lb/in.2a.
15.5 2,463 NEVER OPENED 11.20/12.55 2,400/2,400 Maximum steam generator pressure, lb/in.2a.
16.9 1,154 10.0 1,175 14.8/14.2 26.5/1199.5 Steam generator safety valves close, 650.0 1,056 lb/in.a.
Operator opens atmospheric steam dump valves 1,800.0 27.0 to begin plant cooldown to shutdown cooling.
Shutdown cooling initiated.
11,600.0
SAFETY ANALYSIS RESULTS Using linear MSSV model:
Using MTC=-0.7xl0e-4 %Ap/*F (corresponds to current SONGS 3 MTC at 250 EFPD, SONGS 2 MTC is more negative), peak secondary pressure is 1208.5 psia Using MTC=0.0x10e-4b#0F (Technical Specification maximum), peak secondary pressure is 1214 psia Using best estimate MSSV model:
Peak secondary pressure is 1199.5 psia
LONG TERM MODIFICATIONS Change ring settings:
Nozzle ring to -100 notches Guide ring to -50 notches Measure spring rates of all MSSVs Modifications to be performed at cycle 5 refueling outages on Units 2
& 3 concurrent with valve overhauls
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1210 psig) of its design pressure of 1100 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accord ance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1974 Edition. The total relieving capacity for all valves on all of the steam lines is 15,473,628 lbs/hr which is 102.3 percent of the total secondary steam flow of 15,130,000 lbs/hr at 100% RATED THERMAL POWER. A minimum of I OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduc tion in secondary system steam flow and THERMAL POWER required by the reduced readtor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:
For two loop, four pump operation SP = (X) (Y)(V) x 111.3 where:
SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER.
V = maximum number of inoperable safety valves per steam line.
111.3 = Power Level-High Trip Setpoint for two-loop operation.
X = Total relieving capacity of all safety valves per steam line in lbs/hour (15,473,628 lbs/hr at 1190 psia).
Y = Maximum relieving capacity of any one'safety valve in lbs/hour (859,646 lbs/hr at 1190 psia).
SAN ONOFRE-UNIT 2 B 3/4 7-1