ML17279A882
| ML17279A882 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/31/1988 |
| From: | Hibbard M, Krajicek J, Rawlings J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17279A877 | List: |
| References | |
| ANF-88-02, ANF-88-2, NUDOCS 8803150320 | |
| Download: ML17279A882 (56) | |
Text
SabSISOSSO SSOSO~-
PDR ADOCK 05000397, (
P PDR iq NUCLEARFUELS CORPORATION 'DVANCED ANF-88-01 Issue Date: 1/15/88 WNP-2 CYCLE 4 PLANT TRANSIENT ANALYSIS Prepared By:
. E. Krajicek/H. J. Hi.bbard BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services
'Prepared By:
J. C. Rawlings ENSA AN AFFIUATE OF KAAFTWEAK UNION Qxae'u
NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Advanced Nuclear Fuels Corporation. It Is being submitted by Ad-vanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to fac/litate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Advanced Nuclear Fuels Corporation.fabricated reload fuel or other technical services provided by Ad-vanced Nuclear Fuels Corporation for light water power reactors and It is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, informa-tion, and belief. The Information contained herein may be used by the U.S.
Nuclear Regulatory Commission ln its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Cor-poration in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warranties and representations concem-Ing the subject matter of this document are those set forth ln the agreement bet-ween Advanced Nuclear Fuels corporation and the custo/rter to vrhlch'this docb=
ment Is Issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:
A. Makes any warranty, or representation, express or im-plied, with respect to the accuracy, completeness, or use-fulness of the information contained in this document, or that the use of any Information, apparatus, method, or pro-cess disclosed in this document will not Infringe privately owned rights, or 8 Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, ap-paratus, method. or process disclosed In this document.
XN NF F00 768 (1/8
ANF-88-01 TABLE OF CONTENTS Section ~Pa e 1..0 INTRODUCTIONo ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 2 .0
SUMMARY
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN............................. 5 3 .1 Design Basis........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 3.2 Anticipated Transients................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 3.2. 1 Load Rejection Without Bypass.......... ~ ~ ~ ~ 6 3.2.2 Feedwater Controller Failure........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 3.2.3 Loss Of Feedwater Heating.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9 3.3 Calculational Model.................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9 3.4 Safety Limit........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9
. 3.5 4.1 Final Feedwater Temperature Reduction..
HAXIHUH D essgn OVERPRESSURIZATION...................
Bases......................................................
~ ~ ~ ~ ~ 10 32 32 4.2 Pressurization Transients......................................... 32 4.3 Closure Of All Hain Steam Isolation Valves........................ 33 5.0 RECIRCULATION FLOW RUN-UP......................................... 34 6 .0 REFERENCES........................................................ 37 A PPENDIX Ao ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A 1
ll 0 ~
ANF-88-01 LIST OF TABLES Table Pa<ac 2.1 Thermal Margin Summary For Cycle 4........................... ~ ~ ~ ~ ~ ~
3.1 Design Reactor And Plant Conditions For WNP-2................ ~ ~ ~ ~ ~ ~ 11 3.2 Significant Parameter Values Used In Analysis For WNP-2...... ~ ~ ~ ~ ~ ~ 12 3.3 Results Of System Plant Transient Analyses................... ~ ~ ~ ~ ~ ~ 15 5.1 Reduced Flow HCPR Operating Limit For WNP-2.... ~ ~ ~ ~ ~ ~ 35 LIST OF FIGURES Ficiure ~Pa e 3.1 Load Rejection Without Bypass Results, RPT Operable, Normal Scram Speed.................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 16 3.2 Load Rejection Without Bypass Results, RPT Operable, Normal Scram Speed.................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17 3.3 Load Rejection Without Bypass Results, RPT Inoperable, Normal Scram Speed.................... ~ ~ ~ ~ ~ ~ 18 3.4 Load Rejection Inoperable, Normal Scram Without Bypass Results, Speed....................
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '9 3.5 Load Rejection Without Bypass Results, RPT Operable, Tech. Spec. Scram Speed............. 20 3.6 Load Rejection Without Bypass Results, RPT Operable, Tech. Spec. Scram Speed....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 21 3.7 Load Rejection Without Bypass Results, RPT Inoperable, Tech. Spec. Scram Speed............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 22 3.8 Load Rejection Without Bypass Results, RPT Inoperable, Tech. Spec. Scram Speed.......... .. . 23 3.9 Load Rejection Without Bypass Results, End-Of-Cycle Minus 2000 MWD/HTU Exposure, RPT Inoperable, Tech . Spec. Scram Speed ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 24 Load Rejection Without Bypass Results, End-Of-Cycle Minus 2000 MWD/HTU Exposure, RPT Inoperable, Tech . Spec. Scram Speed ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 25
ANF-88-0' IS OF FIGURES (Continued)
~F3 ere Pacae
- 3. 11 Feedwater Controller Failure Results For 47% Power And 106% Flow With Normal Scram Speed. 26 3e 12 Feedwater Controller Failure Results For 47% Power And 106% Flow With Normal Scram Speed............. .. 27 3e 13 Feedwater Controller Failure Results, RPT Inoperable, N ormal Scram Speed..:..................................... 28 3e 14 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed......................................... 29
- 3. 15 Feedwater Controller Failure Results For 47% Power And 106% Flow With Tech. Spec. Scram Speed...................: 30 3e 16 feedwater Controller failure Results For 47% Power And 106% Flow With Tech. Spec. Scram Speed........
5e I Reduced Flow MCPR Operating Limit.........................
A. I WNP-2 Cycle 4 Safety Limit Local Peaking Factors
( ANF XN 3 Fuel)........................................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A 5 A.2 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-I, -2 Fuel)...................................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
A.3 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (GE Fuel ) ~ ~ ~ ~ ~ ~ ~ ~ ~ A-7 A.4 Radial Power Histogram For I/4 Core Safety Limit Model... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
ANF-88-01
1.0 INTRODUCTION
This report presents the results of the Advanced Nuclear Fuels Corporation (ANF) evaluation of system transient events for the Supply System Nuclear Project Number 2 (WNP-2) during Cycle 4 operation. For this analysis the Cycle 4 core was assumed to contain 428 ANF 8x8 and 336 GE P8x8R fuel assemblies.
This evaluation together with the cycle operation extension achievable with final feedwater temperature reduction( ) (FFTR) and core transient events( )
determines the necessary thermal margin (HCPR limits) to protect against the occurrence of boiling transition during the most limiting anticipated transient. The evaluation also demonstrates the vessel integrity for the most limiting pressurization event. This evaluation is applicable to core flows up to the maximum attainable with the recirculation flow control valve in its
~ ~
fully open position which is 106 percent of the rated core flow value at 100%
~ ~ ~ ~
power. The methodology for these system transient analyses is detailed in References 3 and 4.
ANF-88-01 2.0
SUMMARY
The Minimum Critical Power Ratios (MCPR) for potentially limiting plant system transient events at increased core flow+ are shown in Table 2. 1 for powers that bound allowable values (47 to 104% power) at increased core flow. The system transient HCPR values of Table. 2.1 for the load rejection without bypass (LRNB) and feedwater controller failure (FWCF) transients were obtained using a scram time based on WNP-2 measured values. The loss of feedwater heating (LOFH) transient results shown in. Table 2. 1 were obtained from a bounding analysis which is discussed in Section 3.2.3. The limiting HCPR values for the cases of Table 2.1 are 1.31 for GE and 1.30 for ANF fuel.
Also, an analysis was performed for the LRNB event at a cycle exposure of EOC
-2000 HWd/HTU when a large number of control blades are still inserted in the core. ~ The analysis showed that this system transient was insignificant relative to the control rod withdrawal event (CRWE)(2). Thus, plant operating
~
limits can be based on CRWE event for, cycle exposures up to EOC -2000 MWd/HTU.
~ ~
For exposures beyond EOC -2000 HWd/HTU the limits in Table 2. 1 are applicable.
Additional transient analyses were performed assuming the recirculation pump trip (RPT) out of service and assuming technical specification scram speed (TSSS). The delta CPR results for these events are presented in Section 3.
The maximum system pressure was calculated for the containment isolation event, which is a rapid closure of all main steam isolation valves, using the scenario as specified by the ASME Pressure Vessel Code. This analysis shows that for WNP-2 Cycle 4 operation the safety valves have sufficient capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design pressure. The maximum
- The Cycle 2 transient events analyzed at the design basis power condition (104%) with increased core flow were found to bound the same transients analyzed at the design basis power (104%) and flow condition (100%)
for WNP-2 Cycle 2. These results are shown in Reference 5..
ANF-88-0 system pressures predicted during the event are below the ASHE limit of 1375 psig (110% of design pressure) and are shown in Table 2.1. The analysis conservatively assumed six safety relief valves out of 'service.
The continued applicability of the previously established HCPR safety limit of 1.06 in Cycle 4 was confirmed for all fuel types using the methodology of Reference 6.
ANF-88-01 TABLE 2.1 THERMAL MARGIN
SUMMARY
FOR CYCLE 4 Transient / Power / Flow Delta CPR MCPR*
GE Fuel ANF Fuel Load Rejection~ 104/106 0.25/1.31 0.24/1.30 Without Bypass Feedwater Controller** 47/106 0.12/1.18 0.11/1.17 Failure Loss of Feedwater*** Not Applicable 0.09/1.15 0.09/1.15 Heating MAXIMUM VESSEL PRESSURE (PSIG)
Transient Vessel Dome Vessel Lower Plenum Steam Line HSIV Closure 1286 1315 1289
- HCPR value using the 1.06 safety limit justified herein.
- These transients were evaluated with normal scram speed with RPT operable.
- WNP-2 plant specific bounding value, Reference 10.
ANF-88-01 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 Desi n Basis System transient analyses to determine the most limiting type of thermal margin transient were performed at the increased core flow condition of 106%..
As shown in Reference 5, system transients from the increased core flow condition bound thermal margin analyses transients from the nominal (100%)
flow condition. Analysis of the LRNB was performed at the rated design 104%
power/106% flow point. Since feedwater controller failure (FWCF) transients may be more severe at reduced power because of the larger change in feedwater flow, a FWCF transient was performed at the minimum- power (47%) that allowed for increased core flow. The initial conditions used in the analysis for transients at the 104% power/1065 flow point are as shown in Table 3. 1. The
~
most limiting exposure in cycle was determined to be at end of full power
~ ~ ~
~
capability when control rods are fully withdrawn from the core; the thermal
~ ~ ~
margin limit established for end of full power conditions is conservative in
~
relation to cases where control rods are partially inserted.
The calculational models used to determine thermal margin include .the ANF plant transient and core thermal-hydraulic codes as described in previous documentation( >>6~ ). Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with RODEX2(8). Recirculation pump trip (RPT) coastdown was input based on measured WNP-2 startup test data, and the COTRANSA system transient model for WNP-2 was benchmarked to appropriate WNP-2 startup test data. The hot channel performance is evaluated with XCOBRA-T(4) using COTRANSA supplied boundary conditions. Table 3.2 summarizes the values used for important parameters in the analysis.
ANF-88-0 3.2 Antici ated Transients ANF considers eight categories of potential system transient occurrences for Jet Pump BWRs in XN-NF-79-71( ). The three most limiting transients are described here in detail to show the thermal margin for Cycle 4 of WNP-2.
These transients are:
Load Rejection Without Bypass (LRNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LOFH)
A summary of the transient analyses is shown in Table 3.3. Other plant transient events are inherently nonlimiting or clearly bounded by one of th above events.
3.2.1 Load Re 'ection Without B ass This event is the most limiting of the class of transients characterized by rapid vessel pressurization. The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculation pump trip (RPT). The compression wave produced by the fast turbine control valve closure travels through the steam lines into the vessel and pressurizes the reactor vessel and core'. Bypass flow to the condenser, which would mitigate the pressurization effect, is conservatively not allowed. The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth due to RPT. Figures 3. 1 through 3. 10 depict the time variance of critical reactor and plant parameters from the analysis of the load rejection transient from the design basis power and increased core flow point for a matrix of cases which involve normal scram speed, technical speci'fication scram speed, and recirculation pump trip (RPT) in service and out of service.
ANF-88-01 Analysis assumptions are:
Control rod insertion time based on WNP-2 measured data (normal scram speed) and technical specification scram speed.
Integral power to the hot channel was increased by 10% for the pressurization transient, consistent with Refer ence 9.
Table 3.3 shows delta CPR values for a matrix of LRWB transients with the RPT out of service with both normal scram speed (NSS) and technical specification scram speed (TSSS).
Because a significant number of control rods are inserted into the core at exposures ..less than end-of-cycle'EOC) minus 2000 HWd/HTU, the system transients- are expected to be insignificant for cycle exposures less than this value. To confirm this, the LRWB was analyzed at the same 104% power/106%
~
flow condition point for the end-of-cycle (EOC). minus 2000 MWd/HTU exposure
~ ~
condition for the bounding case of the RPT inoperable with TSSS. The delta CPR values for the GE and ANF fuels for this EOC minus 2000 MWd/MTU case are both 0.05. These delta CPR values are also shown in Table 3.3 and are significantly less than the delta CPR values for the control rod withdrawal error (CRWE) event reported in Reference 2. This shows that the delta CPR for the CRWE bounds plant operation up to EOC minus 2000 MWd/HTU. For Cycle 4 exposures greater than EOC minus 2000 MWd/HTU, the other HCPR values defined in Table 3.3 are applicable.
3.2.2 Feedwater Controller Failure Failure of the feedwater control system is postulated to lead to a maximum increase in feedwater flow into the vessel. As the excessive feedwater flow subcools the recirculating water returning to the reactor core, the core power will rise and attain a new equilibrium
~ ~
if no other action is taken.
Eventually, the inventory of water in the downcomer will rise until the high
ANF-88-01 vessel level setting is exceeded. To protect against wet steam entering the turbine, the turbine trips upon reaching the high level setting, closing the turbine stop valves. The compression wave that is created, though mitigated by bypass flow, pressurizes the core and causes a power excursion. The power increase is terminated by reactor scram, RPT, and pressure relief from the bypass valves opening. The evaluation of this event was performed using the scram and integral power assumptions discussed in 3.2.1. Sensitivity results have shown that the calculated delta CPR is insensitive to the rate of feedwater flow increase, that EOC conditions are bounding because rods are inserted for lower cycle exposure, and that high flows are bounding because of higher axials in the core.
Reference 11 showed that the LRNB is more limiting at full power than the FWCF. Because the total change in feedwater flow is the greatest from reduced power condition, the FWCF was analyzed from reduced power conditions. The FWCF transient event was analyzed from the lowest allowed power (47%) a increased core flow. Figures 3.11 through 3. 16 present key variables. The delta CPR values for the co-resident fuel types for these three 47% power/106%
flow transients are shown in Table 3.3.
It has been determined that for a FWCF event that the control system signal to open the bypass valves passes directly to the bypass valves rather than be delayed by the pressure regulator. Thus, the bypass valves are opened earlier in the Cycle 4 analyses than in the earlier cycle analyses and results in a more realistic representation of the FWCF event. The effect on the FWCF analysis is that the compression wave produced by the turbine control valve closure is mitigated by the earlier opening of the bypass valves, and the core power excursion due to void collapse is diminished which reduces the magnitude of all calculated FWCF delta CPR's. Table 3.3 shows that all of the delta CPR/MCPR values are less than the delta CPR/MCPR value for the 104/106 LRNB event with RPT operable and normal scram speed.
ANF-88-01
'.2.3 Loss Of Feedwater Heat'n Loss of Feedwater Heating (LOFH) events were evaluated with the ANF core simulator model XTGBWR( ) by representing the reactor in equilibrium before and after the event. Actual and projected operating statepoints were used as initial conditions. Final conditions were determined by reducing the feedwater temperature by 100 F and increasing core power such that the calculated eigenvalue remain unchanged.
Based on a bounding value analysis, a MCPR operating limit of 1. 15 for WNP-2 with a MCPR safety limit of 1.06 is supported (i.e., a delta CPR of 0.09). As noted in Section 2.0 of this report, the WNP-2 MCPR safety li'mit for Cycle 4 continues to be 1.06; hence the LOFH transient requires a MCPR operating limit of 1.15 for WNP-2.
Calc lational Model The plant transient codes used to evaluate the pressurization transients (generator load rejection and feedwater flow increase) were the ANF advanced codes COTRANSA( ) and XCOBRA-T(4). This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred. This was accounted for explicitly in determining thermal margin changes in the transient. All pressurization transients were analyzed on a bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel model. The XCOBRA-T code was used consistent with the benchmarking methodology.
3.4 Safet Limit The MCPR safety limit is the minimum value of the critical power ratio (CPR) at which the fuel could be operated where the expected number of rods in boiling transition would not exceed 0. 1% of the fuel rods in the core. The
~ ~
~
operating limit MCPR is established such that in the event the most limiting
~ ~ ~ ~
10 ANF-88-0 anticipated operational transient occur s, the safety limit will not be violated.
The safety limit for all fuel types in WNP-2 Cycle 4 was confirmed by the methodology presented in Reference 6 to have the Cycle 2 value of 1.06. The input parameters and uncertainties used to establish the safety limit are presented in Appendix A of this report.
,3. 5 Final Feedwater Tem erature Reduction Reference 1 presents final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2. The FFTR analysis was performed for a 65 F temperature reduction. This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature. The FFTR analysis results show that delta CPR changes for the LRNB and FFTR transients of plu 0.02 and minus 0.01 are applicable to these respective anticipated operationa occurrence (AOO) events. That is, these LRNB and FFTR limit changes are applicable when Cycle 4 reactor operation is being extended with thermal coastdown at FFTR conditions.
ANF-88-01 TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 Reactor Thermal Power (104%) 3464 HWt.
Total Recirculating Flow (106%) 115.0 Mlb/hr Core Channel Flow 101.8 Hlb/hr Core Bypass Flow 13.2 Hlb/hr Core Inlet Enthalpy 529.2 BTU/ibm Vessel Pressures Steam Dome 1035. psia Upper Plenum 1049. psia Core 1055. psia Lower Plenum 1072. psia Pressure 'urbine 974. psia Feedwater/Steam Flow 15.0 Hlb/hr Feedwater Enthalpy 403.5 BTU/ibm Recirculating Pump Flow (per pump) 17.3 Hlb/hr
12 ANF-88-0.
TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 High Neutron Flux Trip 126.2%
Void Reactivity Feedback 10% above nominal*
Time to Deenergized Pilot Scram Solenoid Valves 200 msec Time to Sense Fast Turbine Control Valve Closure 80 msec Time from High Neutron Flux Time to Control Rod Motion 290 msec Scram Insertion Times (normal)"' 0.404 sec to Notch 45 0.660 sec to Notch 39 1.504 sec to Notch 25 2.624 sec to Notch 5 Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke Time (Total) 150 msec Fuel/Cladding Gap Conductance Core Average (Constant) 580. BTU/hr-ft2-F Safety/Relief Valve Performance Settings Technical Specifications Relief Valve Capacity 228.2 ibm/sec (1091 psig)
Pilot Operated Valve Delay/Stroke 400/100 msec
- For rapid pressurization transients a 10% multiplier on integral power is used; see Reference 9 for methodology description.
- Slowest measured average control rod insertion time to specified notches fo each group of 4 control rods arranged in a 2x2 array.
13 ANF-88-01 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint 85% open Condenser Bypass Valve Performance Total Capacity 990. ibm/sec Delay to Opening (80% open) 300 msec Fraction of Energy Generated in Fuel 0.965 Vessel Water Level (above Separator Skirt)
High Level Trip (L8) 73 in Normal 49.5 in Low Level Trip (L3) 21 in Maximum Feedwater Runout Flow Two Pumps 5799. ibm/sec Recirculating Pump Trip Setpoint 1170 psig Vessel Pressure
14 ANF-88-0 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
Control Characteristics Sensor Time Constants Steam Flow 1.0 sec Pressure 500 msec Others 250 msec Feedwater Control Mode Three-Element Feedwater 100% Mismatch Water Level Error 48 in Steam Flow Equiv. 100%
Flow Control Mode Manual Pressure Regulator Settings Lead 3.0 sec Lag 7.0 sec Gain 3.3%/psid
15 ANF-88-01 TABLE 3.3 RESULTS OF SYSTEH PLANT TRANSIENT ANALYSES Haximum Haximum Haximum Core Average System Delta CPR Neutron Flux Heat Flux Pressure GE ANF Event % Rated % Rated Qsi~ Fuel Fuel LRNB 373 119 1170 0.25 0.24 RPT Operable, NSS*
LRNB 125 1181 0.32 0.29 RPT Inoperable, NSS LRNB 442 125 1175 0.32 0.30 RPT Operable, TSSS**
LRNB 574 131 1189 0.38 0.35 RPT Inoperable, TSSS LRNB EOC -2000 HWD/HTU 284 110 1168 0.05 0.05 RPT Inoperable, TSSS FWCF (47% Power/106% 103 50 1010 0.12 0.11 Flow), NSS RPT Operable FWCF (47% Power/106% 129 52 1020 0.15 0.14 Flow), NSS RPT Inoperable FWCF (47% Power/106% 110 51 1013 0.14 0.12 Flow), TSSS RPT Operable HSIV Closure With 669 133 1315 N/A Flux Scram NOTE: All results are for the design power and increased flow point (104%
power/106% flow) unless- otherwise noted.
- Technical Specification Scram Speed (TSSS)
CI lO
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX CI 3. RECIRCULATION FLOM.
CI IA 4. VESSEL STEAH FLOW
- 5. FEEDHATER FLOH CI Ci uj D
+D W m D
D CII O
CC ILJ Q.
CI 12345 123 CI I
CI ii 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50. ll I
TIHE, SEC 00 4
CO I
CD Figure 3.1 Load Rejection M' Bypass Results, RPT Operable, Normal Scram Spe
- i. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL WATER LEVEL (IN)
CI Ol i Ch CI Cl Cll 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 TIHE, SEC Figure 3.2 Load Rejection Without Bypass Results, RPT Operable, Normal Scram Speed
D lO I. NEUTRON FLUX LEVEL
- 2. HEAT FLUX D 3. RECIRCULATION FLOW
- 4. VESSEL STEAH FLOW
- 5. FEEDWATER FLOW D
Cl WD I- D CO lE D
D cu W
D lK W
CL f2345 123 23 34 D
I 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 ll TIME, SEC I CO CO I
Figure 3.'3 Load Rejection With 'ypass Results, RPT Inoperable, Normal Scram Speed
- 1. VESSEL PRESSURE CHANGE (PSI )
- 2. VESSEL WATER LEVEL (IN) .
i 1
0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 I
TIHE, SEC CO CO I
C)
Figure 3.4 Load Rejection Without Bypass Results, RPT Inoperable, Normal Scram Speed
Cl Cl 4D J. NEUTRON FLUX LEVEL
- 2. HEAT FLUX CI
- 3. RECIRCULATION FLOW lA 4.-VESSEL STEAH FLOW
- 5. FEEOWATER FLOW CI uj I- oo M cn CK O
z O Al CJ lZ hJ 2 CL 12345 l23 2 5 3 234 CI 4
0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 I
TIHE, SEC CO CO I
C)
Figure 3.5 Load Rejection M' Bypass Results, RPT Operable, Tech. Spec. Sera ed
CI
- 1. VESSEL PRESSURE CHANGE (PSI )
- 2. VESSEL WATER LEVEL (IN) i 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 I
TIHE, SEC CO CO I
Figure 3.6 Load Rejection Without Bypass Results, RPT Operable, Tech. Spec. Scram Speed
CI ED
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX CI 3. RECIRCULATION FLOW CI LD
- 4. VESSEL STEAM FLOW
- 6. FEEDWATER FLOW CI
~o I- CI W m LX,'I Z
LLj c4 o
CK LLj 0
CI 12345 f23 5 CI I
CI CI 0.25 0.50 0. 5 1.00 1.25 . 0 1.75 2.00 2.25 2.50 rl I
TIME, SEC CO 00 I
C)
Figure 3.7 Load Rejection Withou ass Results, RPT Inoperable, Tech. Spec. Scram Sp
CI
- 1. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL WATER LEVEL (IN) 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 I
TIHE, SEC CO CO I
Figure 3.8 Rejection Without Bypass Results, C)
Load RPT Inoperable, Tech. Spec. Scram Speed
CI
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAH FLOW
- 6. FEEDWATER FLOW 345 3 23 34 0.2b U.bU U./0 L.UU l.2b i.bU ./0 2.00 .50 I
TIME, SEC CO CO I
figure 3.9 C)
Load Rejection Without ass Results, End-Of-Cycle Minus 2000 NWD/MTU Exposure, RPT -erable, Tech. Spec. Scram Speed
- i. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL HATER LEVEL (IN) 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 TIME, SEC Figure 3.10 Load Rejection Without Bypass Results, End-Of-Cycle Minus 2000 MWD/MTU Exposure, RPT Inoperable, Tech. Spec. Scram Speed
- l. HEAT NEUTRON FLUX LEVEI
- 2. FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAH FLOW
- 5. FEEOWATER FLOW 2
4 io i2 is i8 20 1l TIHE. SEC I Co Co I
figure 3.11 feedwater Controller ilure Results For 47K Power And C) 106K flow With Nor ram Speed
CI Ol
- l. VESSEL 2.
PRESSURE CHANGE (PSI)
VESSEL MATER LEVEL (IN)
CI ID
~ i CI Cll
~t CI I
CI I
I0 10 12 is 18 20 TIHE, SEC Figure 3.12 Feedwater Controller Failure Results For 47K Power And 106$ Flow With Normal Scram Speed
CI LO
%4
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX lA 3. RECIRCULATION FLOH Cll 4. VESSEL STEAH FLOW
- 5. FEEDHATER FLOH 3 3 3 CI CI I- i2 i2 w~ 2 CD lK w
0 2
2 lO i2 i6 i8 20 ll TIHE, SEC I CO CO I
C)
Figure 3.13 Feedwater Control le ilure Results, RPT Inoperable, Normal Scram Speed
- i. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL HATER LEVEL (IN) 6 . 8 io i2 i6 i8 20 TIHE, SEC Figure 3.14 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed
CI IO
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAM FLOW
- 5. FEEOWATER FLOW Cl i2 i2 i2 2 ~
4 2 2 4 IQ i0 i2 i4 i6 iB 20 ll TIME, SEC I CO 00 I
Figure 3.15 Feedwater Control Failure Results For 47K Power And CD 106% Flow With T pec. Scram Speed
Cl CV
- i. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL HATER LEVEL (IN)
CI I0 io i2 i4 i6 ie 20 ll TIME, SEC I CO CO I
Figure 3.16 Feedwater Controller Failure Results For 47K Power And C) 106K Flow With Tech. Spec. Scram Speed
32 ANF-88-01 4.0 HAXIHUH OVERPRESSURIZATION Haximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified by the ASHE Pressure Vessel Code. This analysis showed that the safety valves of WNP-2 have sufficient capacity and performance to prevent pressure from reaching the established transient pressur e safety limit of 110% of the design pressure. The maximum system pressures predicted during the event are shown in Table 2. 1. This analysis, also assumed six safety relief valves out of service.
4.1 Desi n Bases The reactor conditions used in the evaluation of the maximum pressurization event are those shown in Table 3. 1. The most critical active component (scram on HSIV closure) was assumed to fail during the transient. The calculation was performed with the ANF advanced plant simulator code COTRANSA( ), which includes an axial one-dimensional neutronics model.
4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that closure of all main steam isolation valves (HSIVs) without direct scram is the most limiting. Since the HSIVs are closer to the reactor vessel than the turbine stop or turbine control valves, significantly less volume is available to absorb the pressurization phenomena when the HSIVs are closed than when turbine valves are closed. The closure rate of the HSIVs is substantially slower than the turbine stop valves or turbine control valves. The impact of this smaller volume is more important to this event than the slower closure speed of the HSIV valves relative to turbine valves. Calculations have determined that the overall result is to cause HSIV closures to be more limiting than turbine isolations.
~ ~
33 ANF-88-01 4.3 Closure Of All Main Steam Isolation Valves This calculation also assumed that six relief valves were out of service and that all four main steam isolation valves were isolated at the containment boundary within 3 seconds. At about 3.3 seconds, the reactor scram is initiated by reaching the high flux trip setpoints. Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization has been reversed. Loss of coolant flow leads to enhanced steam production as less subcooled water is available to absorb core thermal power. The maximum pressure calculated in the steam lines was 1289 psig occurring near the vessel at about 5 seconds. The maximum vessel pressure was 1315 psig occurring in the lower plenum at about 5 seconds. These results are presented in Table 2. 1 and 3.3 for the design basis point.
34 ANF-88-01 5.0 RECIRCULATION FLOW RUN-UP The HCPR full flow operating limit is established through evaluation of anticipated transients at the design basis state. Due to the potential for large reactor power increases should an uncontrolled recirculation flow increase occur from a less than rated core flow state, the need exists for an augmentation of the operating limit HCPR (full flow) for operation at lower flow conditions.
Advanced Nuclear Fuels Corporation determined the required reduced flow HCPR operating limit by evaluating a bounding slow flow increase event. The calculations assume the event was initiated from the 104% rod line at minimum flow and terminates at 120% power at 103% flow (flow control valve wide open).
~
This power flow relationship bounds that calculated for a constant xenon assumption. ~ It was conservatively assumed that the event was quasi-steady and a flow biased scram does not occur.
~
~
The power distribution was chosen such that the HCPR equals the safety limit at the final power/flow run-up point. The reduced flow HCPRs were then calculated by XCOBRA( ) at discrete flow points.
The recirculation flow run-up analysis performed for WNP-2 Cycle 2 was reviewed, and the assumptions and conditions used for Cycle 2 are applicable to Cycle 4. Thus, the reduced flow HCPR operating limit for WNP-2 Cycle 2 is applicable to Cycle 4. This reduced flow HCPR operating limit is presented in Figure 5.1 and tabulated in Table 5. 1. The HCPR operating limit for WNP-2 shall be the maximum of this reduced flow HCPR operating limit and the full flow HCPR operating limit as summarized in Reference 2.
35 ANF-88-01 TABLE 5.1 REDUCED FLOW HCPR OPERATING LIHIT FOR WNP-2 Core Flow Reduced Flow HCPR
~Rated 0 eratin Limit 100 90 1.12 80 1.17 70 1.23 60 1.32 50 1.42 40 1.55
l.6 NOTE: The MCPR operating limit shall be the greater of the rated condition MCPR operating limit or the value for reduced flow from this curve.
30 io 50 60 70 80 90 100 L10 TQTl-IL CQAE AECIACULFITING FLQN (/ AFITFD)
Figure 5. 1 Reduced Flow MCPR Operating Limit I CO CO I
CD
37 ANF-88-01
6.0 REFERENCES
J. E. Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction," XN-NF-87-92, Advanced Nuclear Fuels Corporation, Richland, MA 99352, June 1987.
- 2. J. E. Krajicek, "Supply System Nuclear Project Number 2 (WNP-2) Cycle 4 Reload Analysis," ANF-88-02, Advanced Nuclear Fuels Corporation, Richland, MA 99352, January 1988.
- 3. R. H. Kelley, "Exxon Nuclear Plant Transient Hethodology for Boiling INt R t,'25-Ny-y -71 P, R I I 2 ( 001 t dt, E Nuclear Company, Inc., Richland, WA 99352, November 1981.
- 4. J. Ades and B. C. Fryer, "XCOBRA-T: Computer for Transient H.
it Eppl I-Ryd I,
17 Ill 0 A lyl," ~N-NF-I Eppl t 2 A
d Code
-I RA, 71 X~N-NF-V-IEP, BWR I, Vi Vl Supplement 4, Advanced Nuclear Fuels Corporation, Richland, WA 99352, February 1987 and July 1987.
J. B. Edgar, Letter to Supplemental Licensing Analysis Results, WPPSS, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.
T.
R M.
t,"
Richland, Patten, "Exxon Nuclear Critical Power Hethodology WA
~XN-NF-5240, R 99352, November 1983.
I I I, E N I for Boiling C p Water
- 7. T. L. Krysinski and J. C. Chandler, "Exxon Nuclear Hethodology for Boiling Mater Reactors; THERHEX Thermal Limits Hethodology; Summary 0 Iptl," X~N-57-EE-15 A, II I 3, R I I 2, E N Company, Inc., Richland, WA 99352, January 1987.
- 8. K. R.
~NF--A,R Herckx, Harch 1984.
1*1 I.,ENHechanical "RODEX2 Fuel Rod I 0 p,l Response Evaluation Hodel," XN-
.,IIIII d,NA 957.,
- 9. S. E. Jensen, "Exxon Nuclear Plant Transient Hethodology for Boiling Water Reactors: Revised Hethodology for Including Code Uncertainties in Determining Operating Limits for Rapid Pressurization Transients in RIIII," XN-NF IfN, R I I 2, 2 ppl t I,'2, d 3, E N Company, Inc., Richland, WA 99352, Harch 1986.
- 10. "Exxon Nuclear Hethodology for Boiling water Reactors Neutronics Hethods 7 0 10 A Exxon Nuclear Company, lyl," Inc.,'ichland,
~XN-Np- -I A, 71 I, Rppl I d I, WA 99352, Harch 1983.
- 11. J. E. Krajicek, "WNP-2 Cycle 2 Plant Transient Analysis," XN-NF-85-143, Revision 1, Exxon Nuclear Company, Inc., Richland, WA 99352, February 1986.
A-I ANF-88-01 APPENDIX A MCPR SAFETY LIMIT A.l INTRODUCTION Bundle power limits in a boiling water reactor (BWR) are determined through evaluation of critical heat flux phenomena. The basic criterion used in establishing critical power ratio (CPR) limits is that at least 99.91. of the fuel rods in the core will be expected to avoid boiling transition (critical heat flux) during normal operation and anticipated operational occurrences.
Operating margins are defined by establishing a minimum margin to the onset of boiling transition condition for steady state operation and calculating a
~ ~ ~ ~
transient effects allowance, thereby assuring that the steady state limit is protected during anticipated off-normal conditions. This appendix addresses
~
the calculation of the minimum margin to the steady state boiling transition condition, which is implemented as the MCPR safety limit in the plant technical specifications. The transient effects allowance, or the limiting transient change in CPR (i.e., delta CPR), is treated in the body of this report.
The MCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associated with the thermal hydraulic state of the reactor using design basis radial, axial, and local power distributions. Some of the calculational uncertainties, including those introduced by the critical power correlation, power peaking, and core coolant distribution, are fuel related. When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the MCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core. Similarly, when an ANF-fabricated reload batch is used to replace a group of dissimilar fuel assemblies, the core average fuel dependent parameters change because of the difference in the
A-2 ANF-88-01 relative number of each type of bundle in the core, and the HCPR safety limit is again reevaluated..
The design basis power distribution is made up of components corresponding to representative radial, axial, and local peaking factors. Where such data are appropriately available from the previous cycle, these factors are determined through examination of operating data for the previous cycle and predictions of operating conditions during the cycle being evaluated for the HCPR safety limit. If operating data are not available, either because the reactor has not been operated or because appropriate data cannot be supplied to ANF, the safety limit power distribution is determined strictly from the predicted operating conditions during the cycle being evaluated. Operating data for WNP-2 during Cycle 3 and the predicted operating conditions for Cycle 4 were evaluated to identify the design basis power distributions used in the Cycle 4 HCPR safety limit analysis.
A-3 ANF-88-01 A.2 ASSUMPTIONS A.2. 1 Desi Basis Power Dist ibutio The local and radial power distributions which were determined to be conservative for use in the safety limit analysis are shown in Figures A-1 through A-4.
A.2.2 H draulic Demand Curve Hydraulic demand curves based on calculations with XCOBRA were used in the safety limit analysis. The XCOBRA calculation is described in ANF topical reports XN-NF-79-59(A), "Methodology for Calculation of Pressure Drop, in BWR Fuel Assemblies," and XN-NF-512(A), "The XN-3 Critical Power Correlation."
A. 2.3 S stem Uncertainties System measurement uncertainties are not fuel dependent. The values reported by the NSSS supplier for these parameters remain valid for the insertion of ANF fuel. The values used in the safety limit analysis are tabulated in the topical report XN-NF-524(A), "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
A.2.4 Fuel Related Uncertainties Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty. The values used in the safety limit analysis are tabulated in the topical report XN-NF-524(A), "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." Power measurement uncertainties are established in the topical report XN-NF-80-19(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water. Rectors; Neutronics Methods for. Design and Analysis."
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ANF-88-0 A.3 SAF TY LIMIT CALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." With 500 Monte Carlo trials it was determined that for a minimum CPR value of 1.06 at least 99.91. of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95/..
A-5 ANF-88-01 LL L ML M ML L LL 0.93 0.95 1.02 1.06 1.02 0.95 0.92 L ML H ML H H M L 0.95 0.97 1.08 0.87 1.04 1.07 1.04 0.95 ML H H H H H ML ML 1.02 1.08 1.01 1.00 0.98 1.00 0.90 1.02 M ML H W M H H M 1.06 0.87 1.00 0.00 0.90 0.97 1.03 1.06 M H H M W H M M 1.06 1.04 0.98 0.90 0.00 0.99 0.93 1.05 ML H H H H H H M 1.02 1.07 1.00 0.97 0.99 1.00 1.06 1.08 L M ML H M H ML ML 0.95 1.04 0.90 1.03 0.93 1.06 0.96 1.07 LL L ML M M M ML L 0.92 0.95 1.02 1.06 1.05 1.08 1.07 1.03 Figure A.l WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-3 Fuel)
A-6 ANF-88-01 LL L HL H H HL L LL 0.95 0.96 1.00 1.03 1.03 1.00 0.96 0.95
~
L HL H HL H L 0.96 0.98 1.05 0.92 1.03 0.96 HL H H H H. H HL HL 1.00 1.05 1.02 1.01 1.00 1.01 0.94 1.00 H HL H W H H H H 1.03 0.92 1.01 0.00 0.93 1.00 1.03 1.03 M H H H W H H 1.03 1.03 1.00 0.93 0.00 0.97 1.03 ML H H H H H H. H 1.00 1.05 1.01 1.00 1.00 1.02 1.05 1.04 L ML H H H HL HL 0.96 0.94 1.03 0.97 1.05 0.97 1.03 LL L HL H H ML 0.95 0.96 1.00 1.03 1.03 1.03 Figure A.2 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-1, -2 Fuel)
A-7 ANF-88-01 LL L ML M M ML L LL 1.03 1.00 0.99 0.99 0.99 0.99 1.00 1.03 L M H H MH MH ML L 1.00 0.99 1.03 1.02 0.99 0.99 0.97 1.00 ML H L H H MH MH ML 0.99 1.03 0.91 1.02 1.01 0.98 0.99 0.99 M H H W H H MH M 0.99 1.03 1.02 0.00 1.02 1.01 0.99 0.99 M H H L W H H M 0.99 1.02 1.01 0.91 0.00 1.02 1.02 0.99 ML MH H H H L H ML 0.99 0.99 1.02 1.01 1.02 0.91 1.03 0.99 L
1'.00 '.97 ML MH 0.99 H
1.02 H
1.03 H
1.03 M
0.99 L
1.00 LL L ML M M ML L LL 1.03 1.00 0.99 0.99 0.99 0.99 1.00 1.03 Figure A.3 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (GE Fuel)
WNP-2 CYCLE 4 15 DESIGN BASIS RADIAL POWER 12.5 (A
10 n I Co 7.5 C) lL hl Kl 5
2.5 I
CO CO 0.25 0.50 0.75 1 1.25 1.50 1.75 CD BUNDLE PONER FFlCTOR figure A.4 Radial Power ogram For I/4 Core Safety Limit Model
ANF-88-01 Issue Date: I/I5/88 WNP-2 CYCLE 4 PLANT TRANSIENT ANALYSIS Distribution:
- 0. C. Brown R. E. Collingham R. A. Copeland N. J. Hibbard J. G. Ingham S. E. Jensen J. E. Krajicek J. L. Haryott J. N. Morgan J. C. Rawlings (ENSA)
G. L. Ritter H. E. Williamson J. B. Edgar/WPPSS (50)
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