ML20076G957
| ML20076G957 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/31/1983 |
| From: | Chandler J, Collingham R, Kelley R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17194B696 | List: |
| References | |
| XN-NF-83-58, NUDOCS 8309010210 | |
| Download: ML20076G957 (34) | |
Text
XN NF 83 58 DRESDEN UNIT 3 CYCLE 9 PLANT TRANSIENT ANALYSIS AUGUST 1983 RICHLAND, WA 99352
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ERON NUCLEAR COMPANY,Inc.
I 830901C210 830825 PDR ADOCK 05000249 P pon
XN-NF-83-58 DRESDEN UNIT 3 CYCLE 9 PLANT TRANSIENT ANALYSIS Prepared By-
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- R.~H. Kelley (j
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Plant Transient Analysis Prepared By: ,
[ R. C. Collingham, Manager BWR Safety Analysis Concurred By: 7 Y f.f J. C. Chandler
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Reload Fuel Licensing
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Approved By: [C'. [f _ . 2- 9 Tut t, # 2 R. B. Stout, Manager
- Licensing & Safety Engineering 1
Approved By:
G. A. Sofer, Manager Fuel Engineering & Technical Services
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j E(ON NUCLEAR COMPANY,Inc.
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8 XN-NF-83-58 i
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was tierived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services I provnjed by Exxon Nuclear for licht water power reactors and it is true and correct to tne best of Exxon Nuclear's knowledge, information,
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and belief. The information contained herein may be esed by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting nn its behalf:
A. Makes any warranty, express or implied, witn respect to
! the accuracy, completeness, or usefulness of the infor-
[ mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not intnnge privately owned rights; or B. Assumes any liabilities with respect to the use of, or for j
darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
XN- NF F00,766
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XN-NF-83-58
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l TABLE OF CONTENTS SECTION PAGE l
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1.0 INTRODUCTION
............................................. 1 2.0
SUMMARY
................................................. 2 I 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN ................... 5 i
4.0 MAXIMUM OVERPRESSURIZATION ............................... 23
5.0 REFERENCES
............................................... 27 t
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XN-NF-83-58 i
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LIST OF TABLES I
l TABLE PAGE 2.1 Thermal Margin ............................................ 3 2.2 Results of Plant Transient Analyses ..................... . 4 3.1 Design Reactor and Plant Conditions (Dresden 3) ........... 11 I
3.2 Significant Parameter Values Used ......................... 12 3.3 Control Characteristics ................................... 14 i
iii XN-NF-83-58 LIST OF FIGURES FIGURE PAGE 3.1 Generator Load Rejection w/o Bypass ...................... 15 (Expected Power and Flows) 3.2 Generator Load Rejection w/o Bypass ...................... 16 (Expected Vessel Pressure and Level) 3.3 Generator Load Rejection w/o Bypass ...................... 17 (Expected CPR for a Typical Fuel Assembly) 3.4 Increase in Feedwater Flow (Power and Flows) ............. 18 3.5 Increase in Feedwater Flow (Vessel Pressure .............. 19 and Level) 3.6 Increase in Feedwater Flow (Typical CPR) ................. 20 3.7 Loss of Feedwater Heating (Power and Flows) .............. 21 3.8 Loss of Feedwater Heating (Vessel Pressure and Level) .............................. 22 4.1 MSIV Closure without Direct Scram (Power and Flows) ........................................ 25 4.2 MSIV Closure without Direct Scram (Vessel Pressure and Level) .............................. 26 s
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1 XN-NF-83-58
1.0 INTRODUCTION
This report presents the results of Exxon Nuclear Company's (ENC) evaluation of core-wide transient events for Dresden Station Unit 3 during Cycle 9 operation. Specifically, the evaluation determines the necessary thermal margin limit required to protect against the occurrence of boiling transition during the most limiting anticipated transient. Also, the evaluation demonstrates that vessei integrity will De protected during the most limiting pressurization event. The results are also incorporated in Reference 2.
This analysis was performed with the same methodology (l) used to esttblish thermal margin requirements for Dresden Unit 3 Cycle 8. The limiting expected transient, load rejection without condenser bypass, and maximum pressurization event, closure of all main steam isolation valves, were dete.imined to be the same for Cycle 9 as previously determined for Cycle 8(6),
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N 2 XN-NF-83-58 2.0
SUMMARY
The determination of the Minimum Critical Power Ratio (MCPR) for Dresden Unit 3 Cycle 9 was based upon the consideration of various possible operational transients (l). A MCPR of 1.30 or greater for all 8x8 fuel types during Cycle 9 assures that at least 99.9% of the fuel rods in the core vill avoid boiling transition during a full load rejection without condenser bypass at worst case (ena-of-cycle) conditions, as well as other less limiting anticipated operational transients. This analysis was based upon the current Dresden 3 Operating License and associated Technical Specifications. The MCPR operating limits required for the more poten-tially limiting events are shown in Table 2.1. These values are equal to those reported for Cycle 8.
The maximum system pressure has been calculated for the containment isolation event, which is a rapid closure of all main steam isolation valves without scram on valve position or relief through the four l
l electromatic relief valves. The safety valves of Dresden Unit 3 have sufficient flow c apac.i ty and opening rates to prevent pressure from reaching the established transient safety limit of 1375 psig, which is 110%
of design pressure. The maximum systein pressures predicted during the event are shown in Table 2.1.
A su:'raery of results of the transient analyses is shown in Table 2.2.
f This table shows the relative maximum fuel power levels, core averige heat fluxes, and maximum vessel pressures attained during the more limiting transient events.
l _ - - - - - - -
3 XN-NF-83-58 Table 2.1 Thermal Margin Summary DRESDEN UNIT 3, CYCLE 9 CPR/MCPR 8x8 (ENC)
Transient XN-1 & XN-2 8x8R(GE) 8x8(GE)
Generator .25/1.30 .25/1.30 .25/1.30 Load Rejection (w/o bypass)
Increase in .21/1.26 .21/1.26 .21/1.26 Feedwater Flow Loss of .16/1.21 .16/1.21 .16/1.21 Feedwater Heating ,
Maximum Pressure (psig)*
Transient Vessel Dome Vessel Lower Plenum Steam Lines MSIV Closure 1323.1 1347.6 1324.1
- Limit allowed is 1375 psig i I
1
Table 2.2 Results of Plant Transient Analyses Event Maximum Maximum Maximum Neutron Flux Core Average Vessel
(% Rated) Heat Flux Pressure
(% Rated) (psig)
Load Rejection (l) 300% 112.5% 1273 w/o Bypass increase in 260% 115.9% 1196 Feedwater Flow a
loss of Feed- 120% 118.5% 1039 water Heating MSIV Closure 490% 132.1% 1348 w/ flux scram (1) Nominal case, all other events are bounding case E
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f 5 XN-NF-83-58 I
l 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN
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3.1 DESIGN BASIS The plant transient analysis determined that the most thermal l
margin limiting condition was operation at full reactor power. Reactor and plant conditions for this analysis are shown in Table 3.1. The most I
l limiting point in cycle was end of full power capability when control rods are fully withdrawn from the core. The thermal margin limit established for end of full power capability is conservative for cases where control rods are partially inserted or reactor power is less than rated. Following requirements established in the Plant Operating License and associated Technical Specifications, observance of the MCPR operating limit of 1.30 or greater for all 8x8 fuel types protects against boiling transition during all anticipated transients at the Dresden Unit 3 for Cycle 9.
The calculational models used to determine thermal margin include ENC's plant transient (1), fuel performance (4), and core thermal-hydraulic (5) codes as described in previous documentation (l). Fuel pellet to clad gap conductances used in the analyses are based on previously submitted analyses (6). All calculational models have been benchmarked against appropriate measurement data, but the current evaluations are intentionally designed to provide a thermal margin which accounts for the random variability and uncertainty of critical parameters. For the limiting generator load rejection without bypass event, the variability of four critical parameters was statistically convoluted so that the calcu-lated thermal margin bounds 95% of the possible outcomes. Table 3.2 i summarizes the values used for important parameters. Table 3.3 provides
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XN-NF-83-58 the feedwater flow, recirculating coolant flow, and pressure regulation system settings used in the evaluation.
3.2 ANTICIPATED TRANSIENTS ENC considered eight categories of potential transient occur-rences for Jet Pump BWR's in XN-NF-79-71(1). Three of these transients have been evaluated here to determine the thermal margin for Cycle 9 at Dresden Unit 3. These transients are:
. generator load rejection w/o bypass
. increase in feedwater flow
. loss of feedwater heating Other plant transient events are inherently non-limiting or clearly bounded by one of the above.
3.2.1 Generator Load Rejection without Condenser Bypass This event is the most limiting of the class of transients characterized by rapid vessel pressurization. The turbine / generator control system causes a fast closure of the turbine control valves.
Closure of these valves causes the reactor syst 1 to be pressurized while the reactor protection system scrams the reactor in response to the sensing of the fast closure of the control valves. Condenser bypass flow, which can mitigate the pressurization effect, is not allowed. The excursion of core power due to void collapse (by pressurization) is terminated by reactor scram since other mechanisms of power shutdown (Doppler feedback, pressure relief, etc.) are only partly successful. Figures 3.1, 3.2 and 3.3 depict the time variance of critical reactor and plant parameters during a load rejection event with expected void reactivity feedback and
7 XN-NF-83-58 normal scram performance. ENC calculated that the thermal margin (ACPR) required to prevent boiling transition for the nominal case for Cycle 9 was slightly less than previously calculated for the same case for Cycle 8.
ENC had calculated this event for Cycle 8 to determine a ACPR which would not be exceeded in 95% of the possible outcomes of the event when four variables were considered:
. void reactivity
. scram worth
. control rod speed (average of all rods) l
. scram time delay.
The standard deviations of the first two variables were 5%
of their expected value. The standard deviations of the latter two variables were based upon plant test data:
(1) Average rod speed - one standard deviation equals 10.4 centimeters /sec.
(2) Scram time delays - one standard deviation equals 30 l millisecs.
l In the evaluation of Cycle 9, the cycle dependent neutronic
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and thermal hydraulic parameters were considered along with potential changes in control rod performance since the Cycle 8 analysis. While measured control rod speed and scam time delay slightly deteriorated between cycles, the void reactivity for Cycle 9 was less negative. The overall result was calculated 6CPR's for Cycle 9 being slightly less than calulated for Cycle 8. Since neither the mean or standard deviation of 6PR is expected to be greater for Cycle 9, the calculated results for Cycle 8 are retained: '
l 8
XN-NF-83-58 mean ACPR .22 standard deviation .016 95% ACPR .250 3.2.2 Increase in Feedwater Flow Failure of the feedwater control system is postulated to lead to a maximum increase of feedwater flow into the vessel . As the excessive feedwater flow subcools the recirculating water returning to the reactor core, the cere 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 vessel water trip setting is exceeded. To protect against spillover of subcooled water to the turbine, the turbine trips, with resultant closure of the turbine stop valves. The power increase is terminated by scram, and pressure relief is obtained from the bypass valves opening. The present evaluation of this event assumed that all the conservative conditions of Table 3.2 were concurrent; no statistical evaluation was considered, and the ACPR calculated represents a bounding result. Though small differences exist between G.E. and ENC fuel, the highest ACPR of 0.21 reported is adequate to protect all fuel tyoes against boiling transition. Figures 3.4, 3.5 and 3.6 display critical variables for this event.
3.2.3 Loss of Feedwater Heating ;
The loss of feedwater heating leads to a gradual increase in the subcooling of the water in the reactor lower plenum. Reactor power slowly rises to the overpower trip point (120% of rated power) . The gradual power change al'ows fuel thermal response to maintain pace with the I
l 9
XN-NF-83-58 increase in neutron flux. For this analysis, it was assumed that the initial feedwater temperature dropped 1450F linearly over a two minute period. The magnitude of the void reactivity feedback was assumed to be 25% lower than expected, so that the power response to subcooling was gradual, maximizing the thermal heat flux. Scram performance was assumed at its Technical Specification limit with scram worth 20% below expected.
Reactor neutron flux reached 120% of rated before surf ace heat flux increased nearly as much. For conservatism, the thermal margin cal-culation assumed that the heat flux increased 120%, resulting in a predicted tLPR of 0.16 for each fuel type. Figures 3.7 and 3.9 depict the transient progression.
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10 XN-NF-83-58 3.3 CALCULATIONAL MODEL The plant transient model used to evaluate the load rejection and feed water increase event was ENC's advanced code, COTRANSA(1). This one-dimensional neutronics 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. The loss of feedwater heating event was evaluated with the PTSBWR3(1) code since rapid pres-k surization and void collapse do not occur in this event.
3.4 SAFETY LIMIT The 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 heated rods in the core.
Thus, the safety limit is the minimum critical power ratio (MCPR) which would be permitted to occur during the limiting anticipated operational occurrence as previously calculated. The MCPR operating limit is derived by adding the ch3nge in critical power ratio ( CPR) of the limiting anticipated operational occurrence to the safety limit.
The safety limit for Dresden Unit 3 Cycle 9 was determined by the methodology presented in Reference 3, and used to determine the MCPR safety limit for Cycle 8 operation of Dresden Unit 3, to have the following value:
Dresden Unit 3 Cycle 9 MCPR Safety Limit = 1.05.
The input parameter values and uncertainties used to establish the safety limit are as presented in Reference 6.
s
11 XN-NF-83-58 Table 3.1 Design Reactor and Plant Conditions (Dresden 2)
Reactor Thermal Power (Mwt) 2527.0 Total Recirculating Flow (Mlb/hr) 98.0 l I
l Core Channel Flow (Mlb/hr) 87.6 f
Core Bypass Flow (Mlb/hr) 10.4 Core Inlet Enthalpy (BTV/lbm) 522.3 Vessel Pressures (psia)
Dome 1020.0 i Upper Plenum 1026.0 Core 1035.0 Lower Plenum 1049.0 Turbine Pressure (psia) 964.7 l
Feedwater/ Steam Flow (Mlb/hr) 9.8 Feedwater Enthalpy (BTU /lbm) 304.1 Recirculating Pump Flow (Mlb/hr) 17.1 (1)
(1) Per pump l
12 XN-NF-83-58 Table 3.2 Significant Parameter Values Used (1)
High Neutron Flux Trip 3032.4 MW Control Rod Insertion Time 3.5 sec/90% inserted Control Rod Worth 20% below nominal Void Reactivity Feedback 10% above nominal (2)
Time to Deenergized Pilot Scrom 298 msec (maximum)
Solenoid Valves Time to Sense Fast Turbine 80 msec (maximum)
Control Valve Closure Time from High Neutron Flux 290 msec Trip to Control Rod Motion Turbine Stop Valve Stroke 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke 150 msec (Total)
Fuel / Clad Gap Conductance Core Average (Constant) 893 BTU /hr-ft2 OF Limiting Assembly 1430 BTV/hr-ft2OF (variable *) (at 8.475 kw/ft)
Safety / Relief Valve Performance Settings Technical Specifications (1) Generator load rejection w/o bypass event was evaluated statistically (see Section 3.2.1)
(2) 25% for calculations with point kinetics model
- Varies : lightly with power and fuel type
13 XN-NF-83-58 Table 3.2 Significant Parameter Values Used (cont.)
Safety / Relief Valve Performance (cont.)
Pilot Safety / Relief Valve Capacity 166.1 lbm/sec (at 1080 psig)
Power Relief Valves Capacity 620.0 lbm/sec (at 1120 psig)
Safety Val"es Capacity 1432.0 lbm/sec (at 1240 psig)
Pilot Operated Valve Delay / Stroke 0.4/0.1 sec Power Operated Valves Delay / Stroke 0.65/0.2 sec MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint 90% open Condenser Bypass Valve Performance Total Capacity 1085.2 lbm/sec Delay to Opening (from demand) 0.1 sec Opening Time (Entire Bank with 1.0 sec
, (Maximum Demand)
% Energy Generated in Fuel 96.5%
Vessel Water Level (above Separator Skirt)
Normal 30 inches Range of Operation +10 inches High Level Trip 48 inches Maximum Feedwater Runcut Flow (3 pumps) 4966 lbm/sec Maximum Feedwater Runout Flow (2 pumps) 3310.67 lbm/sec Doppler Reactivity Coefficient (nominal) -0.002285/0F/ void fraction Void Reactivity Coefficient (nominal) -15.81$/ void fraction Scram Reactivity Worth -36.639$*
! Axial Power Distribution (Peak / average) 1.191 at x/L = .375 Delayed Neutron Fraction .00519 l Prompt Neutron Lifetime 4.66 x 10-5 sec Recirculating Pump Trip Setpoint 1240 psig (vessel pressure)
The value used in the analysis was 80% of the nominal l
_ _ _ _ _ \
14 XN-NF-83-58 Table 3.3 Control Characteristics Sensor Time Constants h Pressure 0.1 sec Others 0.25 sec Feedwater Control Mode 1-element Feedwater Master Controller Proportional Band 100%
Reset 5 repeats / min Feedwater 100% Mismatch Water Level Error 60 inches Steam Flow (not used) 12 in equivalent Flow Control Mode Master Manual Master Flow Control Settings Proportional Band 200%
Reset 8 repeats / min Speed Controller Settings Proportional Band 350%
Reset 20 repeats / min Pressure Setpoint Adjustor Overall Gain 5 psi /% demand Time Constant 15 sec Pressure Regulator Settings Lead 1.0 sec Lag 6.0 sec Gain 30 psid/100% demand
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23 XN-NF-83-58 4.0 MAXIMUM OVERPRESSURIZATION 4 .1 DESIGN BASIS
(
The reactor conditions used in the evaluation of the maximum k pressurization event are those shown in Table 3.1. In addition to the conservative assumptions shown in Table 3.2, ENC assumed that the four power J
actuated relief valves were not available to vent steam as the ASME Pressure p Vessel Code does not allow credit for power operated relief valves. Also, the
\
most critical active component (scram on MSIV closure) was f ailed during the transient.
f 4.2 PRESSURIZATION TRANSIENTS ENC has evaluated several pressurization events, and has determined that closure of all main steam isolation valves without direct scram is most limiting for maximum vessel pressure. Though the closure rate of the MSIVs is substantially slower than turbine stop or control valves, the compressibility of the fluid in the steam lines causes the severity of the compression wave of the slower closure to be nearly as great as the f aster turbine stop or control valves closures. Essentially, the rate and magnitude of steam velocity reduction is concentrated toward the end of valve stroke, generating a substantial compression wave. Once the containment is isolated, the subsequent core power production must be absorbed in a smaller volume than if the turbine isolation occurred. Calculations have determined that the overall result is to cause containment isolation to be more limiting than turbine isolation.
24 XN-NF 58 4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES This calculation assumed all four steam lines were isolated at the containment boundary within 3 seconds. Due to the valve characteristics and steam compressibility, the vessel pressure response is not noted until about )
3 seconds af ter beginning of valve stroke. Since scram performance was ,
degraded to its Technical Specification limit for this analysis, effective power shutdown is delayed until af ter 5 seconds. Due to limitations in steam venting capacity, (i.e. power operated relief valves failures), significant pressure relief is not realized until after 5 seconds, preventing that mechanism from assisting in power shutdown. Thus, substantial thermal power production enhances the pressurization. Pressures reach the recirculating )
pump trip setpoint (1240 psig) before the pressurization has been reversed by the lif ting of the safety valves. 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 1324.1 psig occurring near the vessel at about 6.75 seconds. The maximum vessel pressure was 1347.6 }
psig occurring in the lower plenum at about 6.5 seconds. Figures 4.1 and 4.2 illustrate the progression of the transient.
The calculation was performed with ENC's advanced plant simulator code, COTRANSA, which includes a one-dimensional neutronics model.
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5.0 REFERENCES
f L (1) " Exxon Nuclear Plant Transient Methodology for Boiling Water React-ors", XN-NF-79-71(P), Revision 2, Exxon Nuclear Company Inc., Rich-land, Washington 99352, November 1981.
(2) "Dresden Unit 3 Cycle 9 Reload Analysis", XN-NF-83_47, Exxon Nuclear Company Inc., Richland, Washington 99352, August, 1983.
/
[ (3) " Exxon Nuclear Critical Power Methodology for Boiling Water Rea-ctors", XN-NF-524(P), Exxon Nuclear Company Inc., Richland, Wash-ington 99352, November 1979.
(4) "HUXY : A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option", XN-CC-33(A), Revision 1, Exxon Nuclear Company Inc., Richland, Washington 99352, November 1975.
(5) " Methodology for Calculation of Pressure Drop in BWR Fuel Assem-blies", XN-NF-79-59(P), Exxon Nuclear Company Inc., Richland, Wash-ington 99352, 1979.
(6) "Dresden-3 Cycle 8 Plant Transient Analysis Report", XN-NF-81-78, Revision 1, Exxon Nuclear Company Inc., Richland, Washington 99352, December 1981.
i 4
I L
[ XN-NF-83-58 s
Issue Date: 8/2/83 1
DRESDEN UNIT 3 CYCLE 9 PLANT TRANSIENT ANALYSIS
-f Distribution J. C. Chandler -
I R. E. Collingham
\
G. C. Cooke f S. L. Garrett S. E. Jensen -
W. V. Kayser -
R. H. Kelley I T. L. Krysinski J. L. Maryott G. F. Owsley -
R. B. Stout R. I. Wescott H. E. Williamson <
CECO /L. C. O'Malley (60)
Document Control (5)
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