ML14113A461
| ML14113A461 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/31/1980 |
| From: | Arlotti M, Skaritka J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML13316A521 | List: |
| References | |
| NUDOCS 8012120434 | |
| Download: ML14113A461 (67) | |
Text
RELOAD SAFETY EVALUATION SAN ONOFRE NUCLEAR GENERATINGSTATION UNIT 1, CYCLE 8 REVISION 1 October, 1980 Edited by J. Skaritka App rdved:
M. GCArlotti, Manager Fuel Licensing and Coordination Nuclear Fuel Division
TABLE OF CONTENTS Title
1.0 INTRODUCTION
1 2.0 REACTOR DESIGN 2
2.1 Mechanical Design 2
2.2 Nuclear Design 2
2.3 Thermal and Hydraulic Design 3
3.0 ACCIDENT EVALUATION 5
3.1 Power.Capability 5
3.2 Accident Evaluation 5
3.3 Incidents Reanalyzed 6
4.0 TECHNICAL SPECIFICATIONS 8
5.0 REFERENCES
10 APPENDIX A - LOCA Analysis for 20 Percent of A-1 Steam Generator Tubes Plugged APPENDIX B -
NON-LOCA Safety Evaluation for B-1 20 percent of Steam Generator Tubes Plugged
LIST OF TABLES Table Title Page 1
Fuel Assembly Design Parameters 2
Core Physics Parameters 12 3
Shutdown Requirements and Margins 13 LIST OF FIGURES Figure Title Page 1
Core Loading Pattern 14 F
Total Versus Axial Offset 3
Technical Specification Figure 2.1.1 -16 Safety Limits:
Temperature, Power, Pressure RCS Flow -
195,000 GPM
1.0 INTRODUCTION
AND
SUMMARY
The San Onofre Nuclear Generating Station Unit 1 is shutdown for Cycle 7/8 refueling and repairs of steam generator tubing. Cycle 8 startup is estimated for late 1980 or early 1981.
This report presents an evaluation for Cycle 8 operation which demon strates that the core reload will not adversely affect the safety of the plant. It is not the purpose of this report to present a reanalysis of all potential incidents. Those incidents analyzed and reported in the FSA(1) which could potentially be affected by fuel reload have been reviewed for Cycle 8 design described herein. The results of new analy ses have been included, and the justification for the applicability of previous results from the remaining analyses is presented. These analy ses assume that:
(1) Cycle 7 operation is terminated between 10030 and 11030 MWD/MTU, (2) Cycle 8 burnup is limited to the end-of-full power capability*, and (3) there is adherence to plant operating limitations given in the technical specifications, and (4) the proposed technical specification changes in Section 4 are implemented.
The San Onofre 1, Cycle 8 core loading pattern is shown in Figure 1.
The one Region 6 and 51 Region 7 fuel assemblies from Cycle 7 will be removed and replaced by 52 Region 10 fuel assemblies. A Region 7 fuel assembly will be reused in the central core position.
Nominal design parameters for Cycle 8 are 1347 Mwt (100 percent rated core power), 2100 psia system pressure, nominal core inlet temperature of 551.5 0F, 4.64 kw/ft average linear fuel power density, and a 195,000 gpm RCS Thermal Design Flow (93.4 percent of Cycle.7 TDF).
The Cycle 8 reduced TDF accounts for up to an equivalent 20 percent steam generator tube plugging.
This report replaces a January 1980 Cycle 8 RSE( 2) which used the same TDF as for Cycle 7. Changes from the January 1980 RSE are noted by bars in the margins.
- Definition:
Full rated power and temperature (approximately 575 0F Tavg), control rods fully withdrawn, and zero ppm of residual boron.
2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 10 fuel assemblies is the same.as the Region 9 assemblies. Table 1 compares pertinent design parameters of the various fuel regions. The Region 10 fuel has been designed according to the fuel performance model in Reference 3.
Clad flattening will not occur during Cycle 8. All fuel regions have a predicted clad flattening time equal to:or greater than 50,000 EFPH. No fuel region will receive this exposure.
2.2 NUCLEAR DESIGN Cycle 8 core loading statisfies an ECCS analysis limit of FT x P (2.89 as shown in Figure 2. The limitations on FT of 2.89 include the effects Q
of the local power peaking of Figure 3.1 in WCAP 8131(4) to assure that the allowable value for LOCA is.
satisfied. The points plotted on Figure 2 in clude maneuvers typically done at San Onofre Unit 1 and variants on these maneuvers done at a number of control rod insertions, times and burnups.
The new FT x P<2.89 limit results from the LOCA analysis for 20 percent steam generator tubes plugged, presented in Appendix A. Section 4 gives the technical specification changes needed.
The limiting FT has been determined for the combination of the most adverse FXY and the most.adverse F that will be experienced during operation in Cycle 8. The most adverse Fxy occurs at beginning of life and the most adverse FN occurs at end of life. The results shown for FT in Figure 2 include uncertainty factors of 15 percen.t for conservatism and 4 percent for manuf.act.uring tolerances.
The xenon transient analysis has been evaluated similarly to analyses of previous cycles. The most limiting F, including an uncertainty of 10 percent on F is 1.87 at 84 percent of core height. With the Cycle 8 FH of 1.55, an FN of 1.96 at 84 percent of. core height AH-2 would be required to reach a DNBR of 1,30 at this elevation and 118 percent power. This margin exists assuming a control rod withdrawal occurs with the rods moving to the fully withdrawn position.
Table.2 provides a comparison of Cycle 8 kinetics characteristics with the current limit based on previously submitted accident analysis. The effect of the.Table 2 parameters, including those that fall outside the current limits, are evaluated in Section 3. Table 3 provides the end of-life control rod worths and requirements at the most limiting condi tion during the cycle. The required shutdown margin is based on a previously submitted accident'analysis.(5)
The available shutdown margin exceeds the minimum required to meet the accident analysis.
2.3 THERMAL AND HYDRAULIC DESIGN A reduction in thermal margin for Cycle 8 will result due to a 6.6 per cent reduction in thermal design flow. New DNB Core Limits (See Section
- 4) are generated using the Cycle 5 through 7 design axial power shape with a DNB design FN of 2.07 at 85 percent core elevation. The z
N difference between the DNB design 2.07 FN and the 1.92 limiting F at 85 percent core elevation results in a DNBR margin of 5.5 percent. Reference 7 identified this margin as available to offset rod bow DNBR penalties. However, since the stainless steel clad fuel has less than a 50 percent gap closure, there is no rod bow DNBR penalty(6), and the 5.5 percent margin is available for other uses.
The above design 2.07 FN is too restrictive for DNB limiting ac z
cident analyses.
Therefore, a DNB design axial power shape with an.
FN of 1.95 at 85 percent core elevation is used for DNB limiting z
accident analyses. This 1.95 value provides.a 1.1 percent DNBR margin N
when compared to the 1.92 limiting F.
This margin is part of the Cycle 3 total DNB margin of 4.4 percent shown below, which is a reduction from the previous 8.8 percent margin defined in Reference 7.
DNBR margins available for Cycle 8:
Pitch reduction 3.3 percent Adverse design axial power shape 1.1 percent Total DNBR margin 4.4 percent For all DNBR analyses, the local power spike due to fuel densification is not included, as justified in Reference 8.
3.0 ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability is evaluated considering the consequences of those incidents examined in the FSA,(1) using the previously accepted design basis. It is concluded that the' core reload will not adversely affect the ability to safely operate at 100 percent of rated power and 195,000 gpm RCS Thermal Design Flow (93.4 percent Cycle 7 TDF) during Cycle 8. For Condition II overpower transients, the fuel centerline temperature limit of 4700 0F can he accommodated with margin in the Cycle 8 core. The time dependent densification model( 8) was used for fuel temperature evaluations.
The LOCA limit at rated power can be met by maintaining FQ at or below 2.89.*
This limit is satisfied by the power control maneuvers allowed by the technical specifications, which assure that the Interim Acceptance Criteria (IAC) limits are met for a spectrum of small and large breaks.
3.2 ACCIDENT EVALUATION The effects of the reload and the reduced TDF on the design basis and postulated incidents analyzed in the FSA(1 ) are evaluated in Appen dices A (LOCA) and B (non-LOCA).
Most of the non-LOCA incidents are accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis. For those incidents which.were reanalyzed, it was determined that the applicable design bases are not
- exceeded, and, therefore, the conclusions presented. in the FSA are still valid.
- The new FQ limit of 2.89 is the result of the Cycle 8 reduction in TDF (93.7 percent Cycle TDF) which accounts for up to 20 percent equiva lent steam generator tube plugging. Appendix A presents the LQCA ana lysis which establishes the 2.89 F limit., Cycle 7 had a 2.95 F imit.
-5
A core reload can typically affect accident input parameters in the following areas:
core kinetic characteristics, control rod worths, and core peaking factors. Cycle 8 parameters in each of these three areas were examined as discussed below to ascertain whether new accident ana lyses were required.
A comparison of Cycle 8 core physics parameters with current limits is given in Table 2. The kinetic values fall within the bounds of the current limits.
Changes in control rod worths may affect differential rod worths, shut down margin, ejected rod worths, and trip reactivity. Tables 2 and 3 show that the maximum reactivity withdrawal rate, and the shutdown margin with the worst stuck RCCA are within the current limits. The ejected rod worths and trip reactivity curve are within the bounds of the previous Cycle 7 evaluation.
Peaking factor evaluations were performed for the rod out of position, dropped RCCA bank, dropped RCCA, and hypothetical steamline break accident to ensure that the minimum ONB ration remains above 1.30.
These evaluations were performed utilizing Cycle 8 transient statepoint information and peaking factors. In each case, it was found that the peaking factor for Cycle 8 was lower than the Value for which DNBR equals 1.30. Appendix B shows that the dropped rod and steambreak incidents satisfy safety limits with the Cycle 8 reduced TDF condi tions. The peaking factors following control rod ejection are within the limits of previous analysis for the EOL zero power and full power cases. Peaking factors for the Cycle 8 BOL zero and full power inci dents exceed previously analyzed values, and these cases are reanalyzed in Appendix B.
3.3 INCIDENTS REANALYZED Appendix A presents the LOCA analysis results for 100 percent rated power and a reduced TDF. Results show that the limiting EQ of 2.89 results in a PCT of 22720F, which satisfies the IAC PCT limit of 23000F.
The required technical specification changes are given in Section 4.
Appendix B presents. the non-LOCA evaluations for the reduced TDF condi tions and Cycle 8 core physics parameters (Table 2).
The following FSA postulated incidents are reanalyzed in Appendix B:
- 1. Uncontrolled RCCA Bank Withdrawal at Power.
- 2. Boron Dilution.
- 3. Control Rod Ejection Accidents.
- 4. Loss of Coolant Flow.
- 5. Steamline Break.
- 6. Dropped Rod The Appendix B reanalysis assumed protection setpoints consistent with those in the technical specifications and those proposed changes given in Section 4.
4.0 TECHNICAL SPECIFICATIONS This section contains the technical content of proposed changes to the San Onofre Plant Technical Specifications. These changes are consistent with the plant operation necessary for the design and safety evaluation conclusions stated previously to remain valid.
Due to the reduced thermal design flow, a number of changes to the present technical specifications are required. *These changes are sum marized below:
Section 2.1, Item (2);
Replace:
The combination of reactor system pressure and coolant temperature shall not exceed the locus of points established for the power level in Figures 2.1.1. If the actual pressure and temperature is above or to the left of the locus of points for the appropriate power level, the safety limit is exceeded.
With:
The combination of reactor power and coolant temperature shall not exceed the locus of points established for the RCS pressure in Figure 2.1.1. If the actual power and temperature is above the locus of points for the appropriate RCS pressure, the safety limit is exceeded.
Figure 2.1.1 -
Replace with Figure 3 in this report.
Section 2.1, Table 2.1 Item 4 Variable Low Pressure Replace > 14.45. (1.3130 AT + Ta) 7298.7 avg ith >26.15 (0.984 AT + Tavg 14341 Section 3.5.2:
Control Group Insertion Limits -
Basis (Item 1, 2nd para.)
Replace:
A more restrictive limit on the design maximum value of specific power, FNH and F is applied to operation Q
in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F NH and F are 13.97 kW/ft., 1.55 A
Q and 2.95, respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.
With:
A more restrictive limit on the design maximum value of specific power, F H and F is applied to operation A
Q in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F H and F are 13.7 kW/ft., 1.55 and 2.89 respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.
Section 3.11 A., Incore Axial Offset Limits (Positive and Negative Offsets) 2.95/P - 2.1225 -
3.0 Replace IAO =
0.03084 2.89/P-2.1225 - 3.0 with IAO =
0.03021 2.95/P - 2.1181 +3.0 Replace IAO =
-.03132 2.89/P - 2.1181 +3.0 with IAO
-.03068
5.0 REFERENCES
- 1. Docket Number 50-206, "San Onofre Nuclear Generating Station, Unit 1, Part 2, Final Safety Analysis".
- 2.
Skaritka J., Editor, "Reload Safety Evaluation -
San Onofre Unit 1, Cycle 8", January 1980
- 3. Miller, J. V. (Ed.),
"Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations",
WCAP-8785, October 1976.
- 4. "Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station Unit 1, Cycle 4", WCAP-8131, May 1973.
- 5.
"SCE Report, "Steamline Break Accident Reanalysis, San Onofre Nuclear Generating Station Units 1, October 1976", Attachment to letter, K. P. Baskin to K. R. Goller, December 29, 1976.
- 6. Letters, J. F. Stolz (NRC) to T. M. Anderson (Westinghouse);
Subject:
Staff Review of WCAP-8691; April 5, 1979.
- 7. "Letter from K. P. Baskin (SCE) to K. R. Goller (NRC);
Subject:
Rod Bow Margin, San Onofre Unit 1; Docket No. 50-206; February 10, 1977.
- 8. Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8218-P-A, *March 1975 (Proprie tary) and WCAP-8219-A, March 1975 (Non-Proprietary).
- 9. Skaritka, J., Editor, "Reload Safety Evaluation -
San Onofre Unit 1, Cycle 7", August 1978.
- 10.
"Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 5", Attachment to letter from Jack B. Moore. to Edson G. Case, March 7, 1975.
TABLE 1 SAN ONOFRE UNIT 1 -
CYCLE 8 Fuel Assembly Design Parameters Region 7
8 9
10 Enrichment (w/o U-235)*
4.00 3.99 3.98 4.0 Density (percent Theoretical)*
94.65 94.59 94.66 95.0 Number of Assemblies 1
52 52 52 Approximate Burnup at 29050
- 20800, 8750 0
- All fuel region except Region 10 are as-built values. Region 10 values are nominal.
TABLE 2 SAN ONOFRE UNIT 1 -
CYCLE 8 Core Physics Parameters Current Limit Cycle 8 Moderator Temperature
-4.0 to 0(1)
-3.3 to -0.3 Coefficient, (Ap/OF) x 104 Doppler Coefficient,
-2.75 to -1.4(9)
-2.6 to -1.4 (Ap/OF).x 105 Delayed Neutron Fraction, 0.50 to 0.70(1) 0.55 to 0.62 eff, (%)
Maximum Prompt Neutron 26(10) 11.4 Lifetime (p sec),
Maximum Reactivity Withdrawal 40(9)
<40 Rate, (pcm/sec)*
- pcm 10-5 Ap TABLE 3 SAN-ONOFRE UNIT 1 -
CYCLES 7 and 8 Shutdown Requirements and Margins Cycle 7 Cycle 8 BOL EOL BOL EOL Control Rod Worth (% Ap)
All Rods Inserted 6.7 7.3 6.8 7.4 All Rods Inserted Less Worst 5.5 6.2 5.8 6.4 Stuck Rod (1) Less. 10%
4.9 5.6 5.2 5.8 Control Rod Requirements (% Ap)
Reactivity Defects (Doppler,
- Tavg, 1.9 2.6 1.9 2.6 Void, Redistribution Rod Insertion Allowance 0.8 0.8 0.9 0.9 (2) Total Requirements 2.7 3.4
.2.8 3.5 Shutdown Margin (1)-(2) (% Ap) 2.2 2.2 2.4 2.3 Required Shutdown Margin
% Ap) 1.25 1.9(5) 1.25 1.9 FIGURE 1 CORE LOADING PATTERN SAN ONOFRE UNIT 1 CYCLE 8 1
2 3
4 5
6 7
8 9
10
.11 1?
I3 14 1
A B
o lo9 9
9 9 fo C
D o0/t8?
ao89 10 0
E o
to9 9 8 78 8to
/o0 11 99 S
sy 10 o Ia?~~
q?
9 989 10 s 0o9 a fy99 Se sylo t
99 loa 8
9 7.9 8
9o89
>is IS 10o a
9999
/0 9
K 10 979889B?99"
-oL o o89 8
8 8,e41o lo S0 0 N
in to 9
9 9
to o
P 101 la 0 9 10 10 10 R
Bo8 Source Location in Fuel Region Number S4
Figure 2 F Total vs. Axial Ofl"set for San Onofre Q
Unit 1-Cycle 8 5.0 F -T 4.0 Cycl e 8 Des ign L mit of 2.89 aL
- 0.
0
-60
~~ _40_30_20
-10 0
_1_
203_40_0_6
-'--I
~
~
Aia Ofse (percentage)
--- j-,-~ ------.--
FIGURE 3
~~~~~
7--i I-Technical Specification Figure 2.1.1__
4-Sa fety Limits__
Temperature, Power, Pressure RCS Flow
-195,000.
GPM
- 7-.
~-
j77:
-4F L
L-0
=77----
W.
-- = =-
=
S i:
.~--.-7---
'3
- 0)
~
PSI_
L E
1 7 R S P e s r
.~--
- *~i2in -i 25 S
I --5 4
6 5
-P-
-77 77 7_
4 6
8 0
APPENDIX A LOCA ANALYSIS FOR 20 PERCENT OF STEAM GENERATOR TUBES PLUGGED
-Al -
APPENDIX A TABLE OF CONTENTS Title Page A.1 LOCA Summary A-3 A.2 RCS Blowdown Calculation A-3 A.3 Lower Plenum Refill/Core Reflood Calculations A-4 A.4 Hot Rod Thermal Transient Calculation A-4 A.5 Results and Conclusions A-5 LIST OF FIGURES Figure Title Page A.1 LOCA Transient Values:
Quality, Rod Film Coefficient, Core Flowrate A-6 A.2 LOCA Transient Values:
Core AP Core Pressure, Break Flowrate A-7 A.3 LOCA Transient Values:
Refill Water Inlet Velocity, Film Coefficient, Core Flow Rate A-8 A.4 LOCA Transient Values:
Peak Clad Temperature, Hot Spot Fluid Temperature A-9 A-2
A.1 LOCA
SUMMARY
The Emergency Core Cooling System (ECCS) performance following a Loss,.
of Coolant Accident (LOCA) has been reevaluated for the San Onofre Nuclear Generating Station (SONGS) Unit 1. For this analysis a uniform tube plugging level of 20% was analyzed for the following cases:
(Full power primary T&P) 100% license core power, 2100 psia system
- pressure, 195,000 gpm RCS flow, 13.7 kw/ft, 2.89 F Previous analyses have demonstrated that SONGS Unit 1 was in compliance with the AEC Interim Policy Statement, "Criteria for Emergency Core Cooling Systems for Light.Water Reactors," published in the Federal Register June 29, 1971. The limiting break has been a Double Ended Cold Leg Guilloting (DECLG) with a discharge coefficient of (CD) of 0.6.
To accomplish the reanalysis for increased steam generator tube plugging and reduced RCS condition, the Interim Acceptance Criteria (1AC) assump tions and the 1971 LAC analytical models were used.
The Double Ended Cold Leg Guillotine with a discharge coefficient of 0.6 was analyzed.
The following is a discursion of all the pertinent differences between the previous analysis and the new analysis, results of the new analysis, and justification that the scope of the new analysis is sufficient to satisfy licensing requirements.
A.2 RCS BLOWDOWN CALCULATION The reactor coolant system blowdown hydraulic transient was calculated A-3
using the SATAN-V computer code. The temperature of the fluid in the reactor vessel upper head was assumed to..be equal to the-hot leg tempera ture.
In the analysis, it was assumed that offsite power was lost coincident with the pipe break and that the reactor coolant pumps tripped.
It was also assumed, in the analysis, that 20 percent of the tubes in each steam generator were plugged.
A. 3 LOWER PLENUM REFILL/0ORE REFLOOD CALCULATIONS The lower plenum refill time was recalculated based on the modified ECCS.1 Consistent with the assumption that offsite power is lost coincident with the break, the modified ECCS will effectively deliver full safety injection flow (720 lb/sec) to the reactor vessel at 26.7 seconds. The "free fall" time required forthe safety injection water to drop from the cold leg nozzle to the lower plenum is approximately 0.9 seconds.
It was also assumed that the lower plenum was empty (no liquid water) at the end of blowdown. The lower plenum is therefore filled, and bottom of core recovery occurs,-at 68.4 seconds after the break.
The core reflood calculation was performed the same way as in the previous analysis with the exception that the resistance in the steam generators was increased to reflect the plugging of 20 percent of the steam generator tubes. As in the blowdown calculation (SATAN-V), the plugged tubes were assumed to be uniformly distributed among the three steam generators.
A.4 HOT ROD THERMAL TRANSIENT CALCULATION The hot rod thermal transient calculation was performed using the LOCTA-R2 computer program. There were no differences in the application of the LOCTA-R2 code between the previous analysis and the.new analysis.
Letter to D. Eisenhut, NRC from K. P. Bask'in, SCE, May. 30, 1978.
A-4.
A.5 RESULTS AND CONCLUSIONS Figures A.1 through A.4 show the transient behavior for key parameters during the accident. A limiting break size of 0.6 DECLG was used. As shown in Figure A.4, the peak-clad-temperature is 227ZoF, which satisfies the Interim Acceptance Criteria PCT of 2300*F. Therefore, even with an equivalent 20% tube plugging, the ECCS IAC PCT safety limit is satisfied by not exceeding an FQ of 2.89.
A-5
A!
dd n't~ 0.'6* UECL G B~v 77T 4177
- 7.
z$7 r-C)
____U S--~
4 u
__7_
I I-7 00L 08 09 Ot Oz 0
- UaDjad
-;Ln Fi gure A.1 LOCA Transient Values:
Quality Rod Film. Coefficient, Core Flow Rate A-6
- .7..
C) cCi c~co CD, cu 0-E
-7 7 7:=
7
'I c:~oo
.~~........
.......7
.C~
A-74
Figure A.3 LOCA Transient Values:
Refill Water Inlet Velocity,Film Coefficient, Core Flow Rate I~~~
jilt-ii I
-77 t/
1j L.
C)I n
I--III j
I I
7
- 7.
.7 I
I
-7 Ii'I F F-t1i 4ji1jipffb jl 0
20 40 60 80 100 120 140 160 Time After Break
-Seconds
LOCA: Acident: 0.6, DECLG B8reak'
&fl1.Jaw-j
-9
APPENDIX B NON-LOCA SAFETY EVALUATION FOR 20 PERCENT OF STEAM GENERATOR TUBES PLUGGED
APPENDIX B TABLE OF CONTENTS Title Page B.1 INTRODUCTION AND
SUMMARY
B-4 B.2 ACCIDENT ANALYSES/EVALUATIONS B-4 Control Rod Withdrawal From Subcriti.cal Condition B-6 Uncontrolled Control Rod Assembly Withdrawal at Power B-7 Boron Dilution B-9 Startup of an Inactive Reactor Coolant Loop B-10 Addition of Excess Feedwater B-11 Large Load Increases B-11 Dropped Rod B-12 Control Rod Ejection Accident B-13 Loss of Coolant Flow B-14 Steamline Break B-16 Loss of Load B-19 B.3 CONCLUSIONS B-20 B.4 APPENDIX B REFERENCES B-21 ZIST OF TABLES Table Title Page B.1 Summary of Rod Ejection Analysis Parameters and Results B22 B 2 Core Parameters Used in Steam Line break DNB Analysis B-23 B-2
APPENDIX B LIST OF FIGURES Figure Title B.1 Overpower and Overtemperature Protection Diagram B-24 B.2 Minimum DNBR for Rod Withdrawal at Power B-25 B.3 Minimum DNBR for Rod Withdrawal From Reduced Power B-26 B.4-Core. Flow Versus Time (Loss of Coolant Flow)
B-27 B.5 Nuclear Power Versus Time (Loss of Coolant Flow)
B-2-8 B.6 Heat Flux Versus Time (Loss of Coolant Flow)
B-29 B.7 DNBR Versus Time (Loss of Coolant Flow)'
B-30 B.8 RCS Pressure and Core Average Temperature (Case A-Steam Line Break)
B-31 B.9 Reactivity, Core Thermal Power and Steam.Flow (Case A -
SL Break)
B -32 B.10 RCS Pressure and core Temperature (Case B -
SL Break)
B.11 Reactivity, Core Power and Steam Flow (Case B - SL Break)
B-34 B.12 RCS Pressure and Core. Average Temperature (Case C -
SL Break)
B-35 B.13 Reactivity4Core Power and Steam Flow (Case C -
SL Break) 2-36 2.14 RCS Pressure and Core Average Temperature 23 (Case 0 -
SZ Break)B,3 2.15 Reactivity, Core Power and Steam Flow (Case D SL. Break)
B-38 B.16
- RCS Pressure and Core Average Temperature (Case E -SL Break)
B-39 Steam Line (SL) Break Cases Case A -Break Downstream of Flow Nozzle With Offsite Power Available Case B -Break' Exit of: SG With Offsite Power Available
- CaseC -
Same Break as Case A Without Offsite Power Case D*
- Same Break as Case B Without Offsite Power Case E -Break Equivalent *to 152 lbs/sec at 920. psia, With Off site Power B-3
B.1 INTRODUCTION AND PURPOSE This safety evaluation has been performed to address the non-LOCA safety considerations in allowing San Onofre, Unit 1, to operate with a sig nificant steam generator tube plugging, level.
Tube plugging in suffi cient numbers results in three effects:
Reactor coolant flow is reduced due to increased steam generator flow resistance.
The steam generator heat transfer is reduced.
Primary reactor coolant mass inventory is reduced.
The effect of the reduced Thermal Design Flow, on each of the accidents analyzed in the San Onofre FSA[11 has been evaluated and included in this report.
B.2 ACCIDENT ANALYSIS The impact of reduced Thermal Design Flow on the non-LOCA accident analyses presented in the San Onofre FSA has been assessed. In general, all of the transients are sensitive to steady state primary flow.. The approach used was to identify the impact of a reduction in Thermal Design Flow on each accident. A study was made of each currently appli cable accident analysis to identify margins to safety limits which could be used to offset penalties due to the reduced primary flow.
B-4
This evaluation was based on the following assumptions:
Maximum core thermal power, MWt 1347 Thermal design flow, gpm/loop 65,000 S.G. tube plugging level, percent 20 Tavg at 100 percent power, OF 575.1 AT at 100 percent power, OF 47.2 FN 1.55 AH Tnoloads F 535 RCS pressure, psia 2100
- This safety evaluation is based on a maximum effective steam generator tube plugging level of 20 percent. All combinations of tube plugging and sleeving yielding an effective plugging level of less than 20 per cent are bounded by this safety evaluation.
In general, reanalysis and evaluation techniques were based on the assumptions and methods employed in the FSAR; exceptions are noted in the discussion of each incident. Based on the above work, reanalysis of the limiting transients is performed to verify safety conclusions.
The most recently applicable analysis used in this report is indicated by the footnote after each accident title.
B-5
Control Rod Withdrawal From a Subcritical Condition[2]
A control rod assembly withdrawal incident when the reactor is sub critical results in an uncontrolled addition of reactivity leading to a power excursion (Section 7.1 of the FSA).
The nuclear power.response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion causes a heatup of the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator tem perature rise is small.
Thus, nuclear power response is primarily a function of the Doppler temperature coefficient. The increase in tem perature would result in more Doppler feedback reducing the nuclear power excursion as presented in Reference 2 which would partially com pensate for the flow reduction. Therefore, the transient is only moder ately sensitive to reactor coolant flow.
The most recent analysis[2] shows that for a 40 x 10-5 ak/sec reactivity insertion rate the peak heat flux achieved is 74 percent of nominal with a resultant peak fuel average temperature of 711OF and peak clad temperature of 560 0F. Sensitivity studies have shown that a 7 percent reduction of flow would degrade heat transfer from the fuel by a maximum of 7 percent; peak fuel and clad temperatures would increase by a maximum of 7 percent yielding maximum fuel and clad temperatures of approximately 724'F and 562oF, respectively.
These temperature increases will not result in violation of any safety limits.. Control rod withdrawal from a subcritical condition is a non-limiting transient and not significantly impacted by the above conditions.
B-6
Uncontrolled Control Rod Assembly Withdrawal at Power[1,2 1 An uncontrolled control rod assembly withdrawal at power produces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature (Section 7.2 of the FSA).
Reduced flows and elevated outlet temperature (Tavg is unchanged) will result in less initial margin to DNB. In addition, the reduced primary flow will increase loop transit time which could require new values of lead/lag time constants to be determined for the variable low pressure setpoint equation. Thus, to assure adequate core protection, the Reactor Core Thermal and Hydraulic Safety Limits have been recalculated consistent with the reduction in RCS flow. Based on these new protection lines the variable low pressure setpoint equation constants have been recalculated with new Core Limits. This accident has been reanalyzed to verify the adequacy of the protection setpoints and the lead/lag time constants.
Method of Analysis The transient was reanalyzed employing the LOFTRAN digital computer code and the same assumptions regarding initial conditions used for the FSA, including:
- 1. Power levels equal to 100 percent and 80 percent of 1347 MWT (plus 3 percent calorimetric error);
B-7
- 2.
Inlet temperature 40F above Tin corresponding to the initial power level;
- 3. Pressure (2070 psia) 30 psi below nominal;
- 4. Reactor trip on high. nuclear flux at 118 percent of nominal power, with trip delay time of 0.5 seconds;
- 5. The setpoints for the variable low pressure reactor trip function are those that presently appear in Section 4 of this report and Figure B.1 with allowances for instru mentation errors. A trip delay time of 2.32 seconds was used. This is consistent with the FSA analysis which included an additional 0.5 second delay in temperature detector response. A additional 4 second delay was, assumed in this analysis to account for the total RTD delay.
- 6. Nominal flow is 65,000 gpm/loop.
Results Figures B.2 and B.3 show the minimum DNBR as a function of inverse reac tivity insertion rate. The limiting case for DNB margin is an inverse insertion rate of 833.3 min/4k from full power initial conditions which results in a minimum DNBR of 1.313.
B-8
Conclusions These results demonstrate that the conclusions presented in the FSA are still valid. That is, the core and reactor coolant system are not ad versely affected since the nuclear flux and variable, low pressure trips prevent the core minimum DNB ratio from falling below 1.30 for this incident. Thus, the setpoint equation changes have adequately compen sated for the reduction in Thermal Design Flow.
Boron Dilution[1]
Reactivity can be added to the reactor with the Chemical and Volume Control System by feeding a more dilute solution of boric acid into the Reactor Coolant System than-is present in the reactor coolant.
Steam generator tube plugging or sleeving does not affect the boron dilution analysis as presented in Section 7.3 of the FSA except for the dilution at power with the reactor in manual control.
Reactivity insertion resulting from a boron dilution is greatest at beginning of life. At beginning of life the Technical Specifications require 1.25 percent Ap minimum shutdown capability. With a maximum reactivity insertion rate of 10-5 Ap/second, as assumed in the FSA, it would require a continuous dilution for approximately 21 minutes to lose the 1.25 percent Ap shutdown margin. This is conservative since the actual shutdown margin at beginning of life would be greater than.the 1.9 percent Ap required by the Technical Specifications at end of life.
B-9
For dilution at power. with the reactor under manual control, protection is provided by the variable low pressure trip.
The variable low pres sure trip has been adjusted to account for the revised operating condi tions; the adequacy of the protection was verified in the rod withdrawal at power analysis. For the dilution rate of,10- Apfsecond, a vari able low pressure reactor trip will occur approximately 102 seconds after the dilution begins from full power.with minimum reactivity feedback conditions.
The conclusions of the FSA remain valid.
Startup of an Inactive Reactor Coolant Loop[ 1]
An inadvertent startup of an idle reactor coolant pump results in the injection of cold water into the core. This accident need not be ad dressed due to Technical Specification restrictions which prohibit operation with a loop out of service for power levels greater than 10 percent.
However, a brief discussion of the impact of a flow reduction on the FSA analysis is included. The analysis in FSA Section 7.4.2 shows that the plant is tripped on the overpower trip (100 percent of nominal).
This trip setpoint is unaffected by flow reductions and would.
be conservative for operation at < 10 percent power since the low over power reactor trip setpoint, with a nominal value of 25 percent, is activated below 10 percent power (permissive circuit 7).
Thus, the nuclear power and heat flux transients are conservative; the lower loop flow would also result in a slightly lower reactivity insertion rate.
The peak heat flux would not exceed 100 percent so adequate DNB margin is available.
B-10
Addition of Excess Feedwater[l]
The addition of excessive feedwater is an excessive heat removal inci dent which results in a power increase due to moderator feedback. FSA Section 7.4.3 presents two cases. The first case assumes that all three feedwater control valves fully open together at full load. The second case assumes the startup of a feedwater pump with one pump already run ning while at 50 percent power; the control valves are in manual.
For both cases the reduction in primary coolant flow will result in slower cooldown, hence, a lower reactivity insertion rate.
The reduction in steam flow results in a decrease in feedwater.flow and a longer time until the operator is alerted by the steam generator high water level alarms. The FSA assumes manual reactor trips on the alarms. However, operation at the new system conditions does not affect the consequences of this transient since the protection system setpoints are derived so as to terminate such transients before a DNB ratio of less than 1.30 or the limiting fuel centerline temperature is reached if no operator action is assumed. The adequacy of the protection was veri fied in the rod withdrawal at power analysis.
Large Load Increases[1]
An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor coolant system temperature.
The maximum thermal power level for this new operating condition being evaluated in this report corresponds to the situation with the turbine B-li
control valves fully open. This eliminates the possibility of a large load increase above this power level.
Therefore, as was shown in FSA Section 7.4.4, a step load increase of 30 percent.to the maximum achiev able steam flow, is not expected to result in a reactor trip since suf ficient margin to the safety limits will exist. Protection for this, accident is provided by the overpower and variable low pressure protec tion setpoints. These have been adjusted to be consistent with the new operating conditions and prevent violation of.the core limits. The adequacy of this protection was verified in the rod withdrawal at power reanalysis.
Dropped Rod[l]
The drop of a Control Rod Assembly results in a step decrease in reac tivity which produces a similar reduction in core power, thus reducing the coolant average temperature. A highly negative moderator tempera ture coefficient assumed in the FSA analysis (Section 7.5) results in a power increase (overshoot) above the turbine runback value of 75 percent power (but less than nominal full power) causing.a temporary imbalance between core power and secondary power extraction capability. The ef fect of a 7 percent reduction in initial RCS flow would be a smaller reduction in coolant average temperature and less of a power overshoot.
A DNB evaluation with a statepoiont based on a peak power overshoot of 103 percent of nominal power was evaluated in conjunction with a 7 per cent reduction in flow as a.limiting case. The results of this DNB evaluation showed that the DNBR limit of 1.30 can be accommodated with margin in the current cycle. The conclusions in.the FSA remain valid.
B-12
Control Rod Ejection Accident[1,3]
The rod ejection transient is analyzed at full power and hot standby for both beginning and end of life conditions (Section 7.6 of the FSA). A reduction in core flow will result in a reduction in heat transfer.to the coolant which will increase clad and fuel peak temperatures. All four cases were reanalyzed. Credit was taken, when applicable, for reduced initial fuel temperatures resulting from revised fuel tempera ture curves calculated for cycle 8 operation.
Method of Analysis Analysis methods and assumptions used in the reevaluation were consis tent with those employed in the most recent reload safety analysis.
The calculation of the RCCA ejection transient is performed in two.
stages, first an average core channel calculation and then a hot region calculation. The average core calculation is performed using spatial neutron kinetics methods to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity. Enthalpy and temperature tran sients in the hot spot are then determined by multiplying the average core energy generation by the hot channel factor and performing a fuel rod transient heat transfer calculation. The power distribution calcu lated without feedback is pessimistically assumed to persist throughout the transient. The DNB time is not calculated. DNB is conservatively assumed to occur near the start of the transient.
B-13
Results The analysis results and inputs are summarized in Table B.l.The condi tions at the hot spot fuel rod do not exceed the limiting fuel crite ria[4] for any case.
The conclusions of the FSA, therefore, are still valid.
Loss of Coolant Flow[2]
As discussed in the FSA, Section 8.2, the most severe loss of flow transient is caused by the simultaneous loss of electrical power to all three reactor coolant pumps. This transient was reanalyzed to determine the effect of steam generator tube plugging on the minimum DNBR reached during the incident. Tube plugging yields lower initial primary flows resulting in less margin to the 1.30 DNBR limit. Since no variable low pressure or overpower signal would be generated for a loss of flow event this is a limiting accident with respect to DNB and was reanalyzed.
Method of Analysis Analysis methods and assumptions used in the reevaluation were consis tent with those employed in the most recent reload safety analysis.
These assumptions included:
- 1. Initial operating conditions most adverse with respect to the margin to DNBR, i.e., maximum steady state power level (103 percent of nominal), minimum pressure (2070 psia), and maximum temperature (579.);
B-14
- 2. Maximum Doppler power and temperature coefficient and most positive moderator temperature coefficient;
- 3. A flow coastdown equal to that used in the FSA. This is acceptable since during the first few critical seconds of.the transient the increased loop resistance due to the steam generator tube plugging has a negligible impact on the flow coastdown;
- 4. The reactor trip is assumed to occur on the low reactor coolant flow signal generated at 82 percent of the new nominal flow. This, too, is conservative since a reactor trip on low bus voltage would pre cede the low flow trip; 5.. 4 percent ao trip reactivity from full power.
Results The minimum value of the DNBR for this incident was greater than the 1.30 limit. Figures B.4 through B.7 show the flow coastdown, nuclear power, heat flux, and minimum DNB ratio vs. time.
B-15
Conclusions Steam generator tube plugging appreciably affects the results of the complete loss of flow transient, however, the minimum ONBR remains above 1.30 for this incident. The complete loss of flow case was analyzed since it is the most limiting one presented in the FSAR. Loss of a single pump with all loops in service is less limiting.
Steam Line Break[5]
The steamline break transient is analyzed for hot zero power, end of life conditions (Section 8.4 of the FSA) for the following cases:
Hypothetical Break (steam pipe rupture)
Inside the Flow Restrictor, with and without offsite power available Outside the Flow Restrictor, with and without offsite power available
- Credible Break (dump valve opening)
A hypothetical steamline break results in a rapid depressurization of the steam generators which causes a large reactivity insertion to the core via primary cooldown. The acceptance criterion for this accident B-16
is that no DNB must occur following a return to power.
This limit, how ever, is highly conservative since a hypothetical steamline break is classified as a Condition IV event. As.such, the occurrence of DNB in small regions of the core would not violate NRC acceptance criteria.
The credible steamline break, a Condition II event, results in a much slower depressurization of the steam generators and, hence, a slower reactivity insertion. The acceptance criterion for this incident is that the core remain subcritical throughout the transient.
A reduction in core flow will result in a reduction in heat transfer from the fuel to the coolant. Thus, the return to power for the hypo thetical break and the closest approach to criticality for the credible break generated in previous analyses would be conservative with respect to the lower initial flow conditions. However, incorporating the standard Westinghouse steamline rupture analysis assumption of a con servatively large feedwater flow rate even for a zero power analysis, not previously included in the San Onofre steamline break analyses, results in a quicker system cooldown. This accident was reanalyzed to verify the conclusions of the FSA with conservative feedwater assumptions and conservative Cycle.8 core physics parameters.
Method of Analysis Analysis methods and assumptions used in the reanalysis were consistent B-17
with those employed in the most recent safety analysis (with the excep tion of the feedwater assumption, discussed be-low).
These assumptions included:
- 1. Most pessimistic initial condition corresponding to hot standby conditions with minimum required shutdown margin at no-load Tavg
- 2. Cycle 8 core parameters including allowances for the most reactive RCCA stuck in its fully withdrawn postion;
- 3. An end of life shutdown margin of 1.9 percent Ao;
- 4. Nominal main feedwater flow at full power for the hypothetical breaks plus an additional 10 percent flow to simulate an auxiliary feedwater runout flow. For the failed open dump valve case only the auxiliary feedwater runout flow is assumed;
- 5. Minimum capability for injection of boric acid via the safety injec tion system.
B-18
Results The minimum value of the DNBR for the hypothetical breaks was greater than the 1.30 limit. Results for the credible confirmed that the core remained subcritical throughout the transient. Table B.2 presents the core parameters for the 4 hypothetical break cases used in ONB evaluations.
Figures B.8 through B.15 present the transient results for those cases summarized in Table B.2.
Figure B.16 presents the transient results for the credible break.
Conclusions The steamline rupture-accident, including conservatively high values of feedwater flow and conservative core physics parameters, has been shown to meet the DNB design basis for the hypothetical breaks and remain subcritical for the credible breaks for the case with 20 percent tube plugging.
Loss of Load[1]
The result of a loss of load is a core power level which momentarily exceeds the secondary system power extraction causing an increase in core water temperature. When not assuming credit for a direct reactor trip on turbine trip, this increase will lead to a significant increase in pressurizer pressure and water volume. The adequacy of the pressure relieving devices, both for the primary and secondary systems, was demonstrated in Section 8.6 of the FSA. A reduction in loop flow and RCS mass inventory will result in a more rapid pressure rise. This B-19
effect will be minor, however, since the reactor is tripped on high pressurizer pressure, maintained at a constant falue of 150 psi above the system operating pressure.
Thus, the time to reactor trip will be decreased which will result in a lower total energy input to the cool ant. Following a loss of load, the DNB ratio never falls below its initial value. In addition, the revised variable low pressure trip setpoint will assure adequate margin to DNB. Therefore, the conclusions of the FSA remain valid.
B.3 CONCLUSIONS To assess the effect on accident analysis of operation of San Onofre, Unit 1, with significant levels of steam generator tube plugging, a safety evaluation was performed.
The transients analyzed were control rod assembly at power, dropped rod, control rod ejection, loss of reactor coolant flow, boron dilution, and steamline break. In addition, an evaluation was performed to identify the effect of a flow reduction on the remaining transients and to quantify margins available to offset penalties. Based on this evalu ation, a 7 percent reduction in Thermal Design Flow corresponding to a maximum effective steam generator tube plugging level of 20 percent will not result in violation of safety limits for the transients evaluated.
B-20
B.4 APPENDIX B REFERENCES
- 1.
Docket Number 50-206, "San Onofre Nuclear Generating Station, Un-it 1, Part 2, Final Safety Analysis."
- 2.
Skaritka, J., editor, "Reload Safety Evaluation, San Onofre Nuclear Generating Station, Unit 1, Cycle 7," August, 1978.
- 3.
Skaritka, J., editor, "Reload Safety Evaluation, San Onofre Nuclear Generating Station, Unit 1, Cycle 8," January, 1980.
- 4.
Risher, 0. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP 7588, Revision 1-A, January, 1975.
5.-
SCE Report, "Steamline Break Accident Reanalysis, San Onofre Nuclear Generating Station Unit 1, October 1976," Attachment to letter K. P. Baskin to K. R. Goller, December 29, 1976.
B-21
TABLE B.l
SUMMARY
OF ROD EJECTION ANALYSIS PARAMETERS AND RESULTS BOL BOL EOL BOL Power Level, percent 0
103.
0 103 Ejected rod worth, percent ak 0.68 0.21 0.58 0.15 Delayed neutron fraction, percent
.55
.55
.50
.50 FQ after rod ejection 8.95 6.11 9.17 7.13 Number of operating pumps 2
3 2
3, Maximum fuel pellet center temperature, OF 2831 4491 1792 4440 Maximum fuel pellet average temperature, OF 2462 3402 1580 3342 Maximum clad average temperature, F 1894 2109 1305 2075 Fuel Pellet Melting, percent 0
0 0
0 Maximum fuel enthalpy (cal/gm) 98.4 143.7 59.2 140.8 B-22
TABLE B.2 CORE PARAMETERS USED IN STEAM LINE BREAK DNB ANALYSIS Outside Break with Power Outside Break without Power Inside Break with Power Inside Break without Power lime (Sec.)
35 45 66 70 37 40 76 80 57 75 77 79 41.5 45 88 92 Core Inlet 420 401 370 366 416 407 352 347 398 377 375 373 422 414 358 354 Temperature (OF)
RCS Flow 100 100 100 100
.45
.40
.19
.18 100 100 100 100
.41
.37
.17
.16 (percent of nominal)
Heat Flux 27.6 28.0 34.1 29.8 23.9 22.6 19.3 18.4 24.1 27.0 30.1 29.2 20.6 19.2 18.3 17.6 (percent of 1347 Mwt)
RCS Pressure 1044 1032 1063 1069 1065 1052 1078 1084 1065 1082 1086 1087 1073 1063 1097 1102 (psia)
60 60 P, 2 P3 P4 P5 DNBR and Vessel. Exit Boiling Limits Overpower Protecti on Liit Steam
,4§ Generator Safety 50
/
Valve LLine OP 40 30 I-4
~40 Variable Low Pressure Trip Lines/i P2 = 1775 psia P3 = 1900 psia P4 = 2100 psia P5 = 2250 psia 3.0k 24 520 540 560 580 600 620 640 VESSEL AVERAGE TEMPERATURE (OF)
FIGURE B.1 OVERPOWER AND OVERTEMPERATURE PROTECTION DIAGRAM B-24
FIGURE-B.2 MINIMUM DNBR VERSUS INVERSE REACTIVITY INSERTION RATE FOR ROD WITHDRAWAL AT POWER TRANSIENTS 4.0
'l'l' Initial Core Power = 1347 MWt 3.5 3.0 2.5 2.0 Minimum ZFeedback 1.5 Maximum 1.5 Feedback 10 100 1000 INVERSE REACTIVITY INSERTION RATE (MIN/6K)
FIGURE B.3 MINIMUM DNBR VERSUS INVERSE REACTIVITY INSERTION RATE FOR ROD WITHDRAWAL FROM REDUCED POWER TRANSIENTS I
I III I
I Initial Core Power 1077.6 MWt 3.5 3.0 2.5 Maximum Feedback 2.0 Minimum Feedback 1.5 1.3 l
10 100 1000 INVERSE REACTIVITY INSERTION RATE (MIN/6K)
1.2000I 1.0000
.80000 a
C
.60000 LL
.40000 U.
U CD
.20000 TIME (SEC)
FIGURE B.4 CORE FLOW VERSUS TIME B-27
1 2000
- 1.
0000
-J
.80000 LL.
c(
S.60000 C
.40000 2000 0.0' CD CDU C
o
)
TIME (SEC)
FIGURE B.5 NUCLEAR POWER VERSUS TIME B-28
HEAT FLUX (FIRAC. OF 103% POWER) o 0.0(
0 CoC'-4)
C C
-I 20.00
-n
'-4 m
8.0000
- t. 000
2, 4000 2.2000.
2 0000
- 1. 8000 16000 1.4000 1 2000 5
0 o
a 0
0 TIME (SEC)
FIGURE B. 7 DNBR VERSUS TIME B-30
CORE AVERAGE TEMPERATURE (OF)
RCS PRESSURE (PSIA) nW J OA LA cO OO
-n LA 0D LA 0
LA 0
C 00000 0
ma 0
0 0
C 0
0 0D 0D 0CD 0
0 0
0 00.0
-n
-1 25.000 no C) 50.000 ri 75.000 C) rri
-A ca 100.00 m
-0 me o
X 125. 00 1 0 0 125.00
CORE THERMAL POWER AND STEAM FLOW (FRAC OF NOM)
REACTIVITY (PERCENT) e-c o
m e
s o
~
C3~~
0 0D00§C)
C3 0
C 00 Cooeo a
onCDC 0.0
- z 50.000 m
rn 75.00(D 100 00 2500 5000 +
2200.0 2000.0 1800.0 1600.0 V) 1400.0 a
1200.0 1000.00 600 00 550.00 500.00
-L 450.00 Lj 0 400-00 Lu 350.00 U
C-)
300.00 CD CD D
CD CD CD TIME (SEC)
FIGURE B.10 TRANSIENT RESPONSE TO STEAMLINE BREAK AT EXIT OF.
STEAM GENERATOR WITH OFFSITE POWER AVAILABLE (CASE B)
B-33
CORE THERMAL POWER AND STEAM FLOW (FRAC OF NOM)
REACTIVITY (PERCENT)
Q Q c
0 C
0
. 0 Q 0 Q
0.0 VI) 25 00 5000 CD m
Ln C-,
75.000 100.00 125.00O
+f I
I
2200.0 I
I 2
2000.0 1800.0 1600.0 V)
L.
1200.0 V) 1200.0 1000-00 LI 550.00
>ii 500.00 I-Li 0 400.00 350.00 300.00 o
a TINE (SEC)
FIGURE B.12 TRANSIENT RESPONSE TO STEAMLINE BREAK DOWNSTREAM OF FLOW MEASURING NOZZLE WITHOUT OFFSITE POWER AVAILABLE (CASE C)
B-35
CORE THERMAL POUER AND STEAM FLOW (FRAC OF NOM)
REACTIVITY (PERCENT) o rr w
0o C C LA C
LA CD cm (3
CD C m
00o 00o o
a g
0 00 0
0 0
0 0
0 0
500 0000 000 0
0 00
-~~
0.~0 00 -
II (s0.0 C-)
C,)
25.000 C)
C S10.0-00 12 -0 iH
~
oJ i
i i
CORE AVERAGE TEMPERATURE (OF)
RCS PRESSURE (PSIA) 0 4w 0 w
LA 0 LA 0
0 0
0p m
0o 0
0o
- o c
ooo ooo oo 0
0 I 0 I
m 0
0D 0
0 0
m P
0 D 0 CD 0
CD 0D C
2 00 0 0
0 0
0.0
-. n T.
25.00 50.000 (A
m
~m r- ><
125.00 150.00
CORE THERMAL POWER AND STEAM FLOW (FRAC OF NOM)
REACTIVITY (PERCENT) fl
~~rj frJ U) r~j
~
n c
oo o
0.00 Ul
)
25.000 C-)
50.00 m
c; 75.000 m
C1 100.00 125.00 1500 i
4
CORE AVERAGE TEMPERATURE
(*F)
RCS PRESSURE (PSIA) w CAr 4w LA LA c 0D LA 01 L"
0 CA1
- 0) c r\\)
rn~~
~
CD um C)
CD n
C 3
0 0
Mo u
e 0 5
C
.0.
o 0
0 o
0 0
00 400 00 0.0 v) 100.00 m L 700.00 m m r
300.00 rn C00.00 C) 5n 00.00 c:C)
(AA Ln m
800.00 n-c:)
C:)
.900.00
-m cn 1000.00--
.