ML20039G956
| ML20039G956 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/23/1981 |
| From: | Cooke G, Fausz N, Kelley R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17194A407 | List: |
| References | |
| XN-NF-81-78, XN-NF-81-78-R01, XN-NF-81-78-R1, NUDOCS 8201190343 | |
| Download: ML20039G956 (45) | |
Text
{{#Wiki_filter:k XN NF 81-78 [ REVISION 1 1 l DRESDEN 3, CYCLE 8 i PLANT TRANSIENT ANALYSIS REPORT I I I DECEMBER 1981 l 1 7 , ,, ~. . g. _ ;6{ . - 4 t,: Q c < -u ~,
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I ' XN-NF-81-78 Revision 1 1Y/Y3/ ,I DRESDEN 3, CYCLE 8 PLANT TRANSIENT ANALYSIS REPORT Prepared by : fd //' n .T.' Fags'z ' () Pla Transient Analysis Prepared by R.H. Kelley D Plant Transient Analysis (/ Prepared by : . C G.C. Cooke, Manager Pla Transient Analysis Concur : /
"G.F. Owsley, Mgnager Reload Fuel Litensing 1 0 I Approve : Ow ~
JJJ. Morgan, Mana@r Licensing & Safety Engineering Approve : k cutan - ~ VIY/7,l{ G.A. Sofer, Manager Fuel Engineering & Technical Services I I
/mb I E(ON NUCLEAR COMPANY,Inc.
I
I I I U. S. CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLE ASE RE AD CAREFULLY I Exxon Nuclear Company's warranties and r+ presentations concerning the subject matter of this document are those set forth in the Agreer.ient between Exxon Nuclear Company, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such Agreement, neither Exxon Nuclear Company, Inc. nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method or process disclosed in this document. The information contained herein is for the snle use of Customer. In order to avoid impairment of rights of Exxon Nuclear Company, Inc. in patents or inventions which may be included in the information con-tained in this document, the recipient, by its acceptance of this document agrees not ta publish or make public use (in the patent sense of the term) of such information until so authorized in writing by Exxon Nuclear = Company, Inc. or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document. I XN-NF F00,765 I
1 i , > i i 1 i XN-NF-81 -78 Revision 1 i ll 4 I 4 i l11 1 i 1 i _T_ABLE OF CONTENTS 1 SECTION PAGE i l l l l.0 INTRODUCTION ................................................. 1 2.0
SUMMARY
...................................................... 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN ........................ 5 ll I
4.0 MAXIMUM OVERPRESSURIZATit,N .................................. 26 l ,
5.0 CONCLUSION
................................................... 30
6.0 REFERENCES
................................................... 32 1
~
!Il Appendix A ........................................................ A-1
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I i i,ll l !Il
- 11 i.ll l
ill 4
i i i XN-NF-81-78 lE .. Revision 1
- E ii LIST OF TABLES
; Table Page 2.1 Thermal hargin ............................................... 3 i
f 2.2 Results of Plant Transient Analyses ...................... ... 4 3.1 Des ign Reactor and Plant Cond it ions (Dresden 3) . . . . . . . . . . . . . .11 l l i 3.2 Significant Parameter Values Used ............................ 12/13 I 3.3 Control Characteristics ...................................... 14 'I
- l. A-1 Dresden-3 Cycle 8 Safety Limit Fuel Related Uncertainties ..... A-3 l
A-2 Dresden-3 Cycle 8 Safety Limit Nominal Input Parameters........ A-4 lI I lI 4 !I 1 1 il
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l I i ] ]' iii XN-NF-81-78 l 1 Revision 1 l LIST OF FIGURES - II 15 FIGURE PAGE l3 1 'I 3.1 S c r am R e a c t i v i t y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 !I 4 I 3.2 Axial Power Distribution ...................................... 16 i 3.3 Generator Load Rejection w/o Bypass ........................... 17 (Expected Power and Flows) J 25 3.4 Generator Load Rejection w/o 8ypass .......................... 18 (Expected Vessel Pressure and Level) 3.5 Generator Load Rejection w/o Bypass .......................... 19 lE !g (Expected CPR for a Typical Fuel Assembly r 3.6 Increase in Feedwater Flow (Power and Flows) .................. 20
- I 3./ Increase in feedwater Flow (Vessel Pressure and Level) . . . . . . . . 21 3.8 Increase in feedwater Flow (Typical CPR) ...................... 22 3.9 Loss of Feedwater Heating (Power and Flows) ................... 23 3.10 Loss of Feedwater Heating (Vessel Pressure and Level) . . . . . .. . . 24 3.11 Loss of Feedwater Heat ing (Typical CPR) . . . . . . . . . . . . . . . . . . . . . . . 25 I 4.1 MSIV Closure without Direct Scram (Power and Flows) ........... 28 4.2 MSIV Closure without Direct Scram (Vessel Pressure and Level).. 29 A-1 Dresden-3 Cycle 8 Safety Limit Radial Power Histogram ..........A-5 A-2 Dresden-3 Cycle 8 Safety Limit Local Peaking ...................A-6 I l I
I I l I
lI l XN-NF-81-78 l Revision 1 !I i i
1.0 INTRODUCTION
'I This report presents the results of Exxon Nuclear Company's (ENC) evaluation of core-wide transient events for Dresden Station Unit 3 during Cycle 8 operation. Specifically, the evaluation determines tra necessary thermal margin to protect against the occurrence of boiling l l transition during the most limiting anticipated transient. Also, the
- evaluation demonstrates that vessel integrity would be protected during the most limiting pressurization event. The bases for the analyses t
herein have been provided in XN-NF-79-71, Revision 2. The results are also incorporated in Reference 2. 1 i !I ,I I . !I
!I
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I 2 XN-NF-81-78 Revision 1 2.0
SUMMARY
The thermal margin limiting anticipated transient for Dresden Unit 3 is the generator load rejection event with failure of the condenser bypass system. The observance of a Minimum Critical Power Ratio (MCPR)(3) of 1.30 or greater during Cycle 8 at rated reactor conditions (and other restrictions as specified by the Dresden Unit 3 Operating License and associated Technical Specifications) adequately prevents fuel boiling transition during this as well as other (less limiting) anticipated transients considered. The MCPR operating limit required for potentially limiting events is shown in Table 2.1 for comparison. While minor fuel type differences were accounted for in the analyses of each event in Table 2.1, the MCPR operating limit was determined to be essentially the same for all fuel types. Maximum 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 ASME Pressure Vessel Code. The safety valves of the Dresden 3 unit have sufficient capacity and performance to prevent pressure from reaching the established transient safety limit of 110% of design pressure (1.1 x 1250 = 1375 psig). The maximum system pressures predicted during the event are shown in Table 2.1. A summary of results of the transient analyses is shown in Table 2.2. I I
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E 1 I l 3 XN-NF-81-78 Revision 1 Table 2.1 Thermal Margin Transient aCPR/MCPR 8x8(XN-1) P8x8R/8x8R(GE) 8x8(GE) I Generator .25/1.30 .75/1.30 .2b.'l . 30 II) Load Rejection (w/o bypass) I increase in Feedwater Flow
.21/l.26 .21/l.26 .21/l.26 Loss of .16/1.21 .16/1.21 .16/1.21 Feedwater Heating Maximum Pressure (psia)*
Transient Vessel Dome Vessel Lower Plenum Steam Lines MSIV Closure 1339 1361 1340
- Limit allowed is 1375 psig (1389.7 psia)
(1) See Section 3.2.1 for basis of this value I I
- _ - - - - - - - . . _ . _ - . - . . . . _ _ _ _ _ _ _ _ _ _ _ __ _ ~ ~ ~ ~
M M M M M M M M M M M m W M M M M M Table 2.2 Results of Plant Transient Analyses Event Maximum Maximum Maximum ACPR Neutron Flux Core Average System (% Rated) Heat Flux Pressure (% Rated) 1 A41 Load Rejection (l) 372% 114.3% DS9 psia .22 I w/o Bypass , i increase in 293% 116.8% 1214 psia .21 Feedwater Flow Loss of Feedwater 120% 119.2% 1056 psia .16 Heating MXIV Closure 490% 131.8% 1361 psia N/A w/ flux scram , (1) Nominal case, all other events are bounding case
I 5 XN-NF-81-78 Revision 1 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 DESIGN BASIS The plant transient analysis determined that the most thermal margin limiting condition was operation at full reactor power. Reactor and plant conditions for this analysis are shown in Table 3.1. The most 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 conditions is conservative for cases where control rods are partially inserted or reactor power is less. Following requirements established in the Plant Operating License and associated Technical Specifications, observance of the MCPR operating limit of 1.30 or greater protects against boiling transition during all anticipated transients at the Dresden Unit 3 for Cycle 8. I 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 pellets to clad gap conductances used in the analyses are based on CdlCulations with RODEX2(6). All calculational models have been bench-marked 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 varia-bility of four critical parameters was statistically convoluted so that the calculated thermal margin bounds 95% of the possible outcomes. Table 3.2 summarizes the values used for important parameters. I
L 6 XN-NF-81-78 Revision 1 Ihe values of gap conductance calculated by RODEX2 for the fuel c- types in the Dresden-3 Cycle 8 core were equivalent for two main reasons. First, RODEX2 predicted gap closure for all fuel types. Second, no credit was taken for fission gas release. The gap conductance was therefore relatively insensitive to rod internal pressure, and conservatively high values for gap conductance were used. The essentially equal gap conductance values resulted in essentially equal changes in thermal margin for all the resident fuel types. Table 3.3 provides the feedwater flow, recirculating coolant flow, and pressure regulation system settings used in the tvaluation. 3.2 ANTICIPATED TRANSIENTS ENC considered eight categories of potential transient occurrences 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 8 at Dresden Unit 3. These transients are:
. genergtor' load rejection w/o bypass
. increase in feedwater flow I . 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 I characterized by rapid vessel pressurization. The turbine / generator control system causes a fast closure of the turbine control valves. The compression wave produced travels through the steam lines into the vessel while the reactor protection system scrams the reactor in response to l l
I XN-NF-81-78 I 7 Revision 1 I 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 is terminated by reactor i scram since other mechanisms of power shutdown (Doppler feedback, pressure relief, etc.) are only partly successful. Figures 3.3, 3.4 and 3.5 j depict the time variance of critical reactor and plant parameters during a load rejection event with expected void reactivity feedback and normal
- scram performance. ENC evaluated this event to determir,e a ACPR which i
l would not be exceeded in 95% of the possible outcomes of the event when i four variables were considered:
. void reactivity
. scram worth
. control rod speed
. 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) Rod speed - one standard deviation equals 10.4 centimeters /sec. (2) Scram time delays - one standard deviation equals 30 millisecs. The calculated results of the statistical evaluation were: mean ACPR - .22 standard deviation - .016 95% oCPR - .25 I I
8 XN-NF-81-78 Revision 1 3.2.2 Increase in Feedwater Flow { Failure of the feedwater control system is tnstulated 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 core power will rise and attain a new equilibrium if no other action is takcn. Eventually, the inventory of water in the downcomer will rise unt 1 the high vessel water trip setting is exceeded. To protect against spillover of subcooled water to the turbine, the turbine trips, closing the turbine stop valves. The compression wave created, though mitigated by condenser bypass flow, pressurizes the core,, causing a power excursion. The power increase is terminated by scram and pressure relief from the bypass valves opening. The present evaluation of this event assumed that all applicable 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 types against boiling transition. Figures 3.6, 3.7 and 3.8 display critical variables for this event. E 3.2.3 Loss of Feedwater Heating g . 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 allows fuel thermal response to maintain pace with the increase in reutron flux. For this analysis, it was assumed that the initial feedwater temperature dropped 1450F ;
I I 9 XN-NF-81-78 Revision 1 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 . Scram performance was assumed at its Technical Specification limit with scram worth 20% below expected. Reactor neutron flux reached 120.1% of rated and clad surface heat flux increased nearly as much. Calculation of thermal margin assumed that bundle ppwer increased by 20% which predicted a ACPR of 0.16 for each type. Figures 3.9, 3.10 and 3.11 depict the transient progression. 3.3 CA_CULATIONAL MODEL The plant transient model used to evaluate the load rejection and feedwater increase event was ENC's advanced code, COTRANSA. This one-dimensional neutronics model predicted reactor power shif ts toward the core middle and top as pressurization occurred. This was accounted for explicitly in detennining thermal margin changes in the transient. The loss of feedwater heating event was evaluated with the PTSBWR3 code since rapid pressurization 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 anti-cipated operational occurrence as previously calculated. The MCPR operating limit is derived by adding the change in critical power ratio (tfPR) of the limiting anticipated operational occurrence to the safety limit. I
!I 10 XN-NF-81-78 l Revision 1 i The safety limit for Dresden 3 Cycle 8 was detennined by the methodology presented in Reference 3 to have a value of 1.05. 1 The input parameter values and uncertainties used to establish the safety limit are presented in Appendix A. il i l lI l I I I lI l l l I l l I I I : I
- I I.
11 XN-NF-81-78 i Revision 1 Table 3.1 Design Reactor and Plant Conditions (Dresden 3) I Reactor Thermal Power (Mwt) 2527 Total Recirculating Flow (Mlb/hr) 98.0 i Core Channel Flcw (Mlb/hr) 87.8 Core Bypass Flow (Mlb/hr) 10.2 Core Inlet Enthalpy (BTU /lbm) 522.3 I Vessel Pressures (psia) Dome IC20 l Upper Plenum 1026 Core 1035 f l . Lower Plenum 1049 l l Turbine Pressure (psia) 965 1 Feedwater/ Steam Flow (Mlb/hr) 9.8 Feedwater Enthalpy (BTU /lbm) 304.1 Recirculating Pump Flow (Mlb/hr) 17.l(1) I I I (1) per pump I I . lI .
I 12 XN-NF-81-78 Revision 1 Table 3.2 Significant Parameter Values Used(I) High Neutron Flux Trip 3032.4 MW Control Rod Insertion Time 3.5 sec/90% inserted Control Rod Worth 10% below nominal (2) Void Reactivity Feedback I Time to Deenergized Pilot Scram 10% above nominal (3) 298 msec Solenoid Valves Time to Sense Fast Turbine 80 msec 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 Contro; Valve Stroke 150 msec (Total) Fuel / Clad Gap Conductance Core Average (Constant) 893 BTU /hr-ft2 OF Limiting Assembly 1430 BTU /hr-ft2 OF (variable *) (at 8.475 kw/ft) Safety / Relief Valve Performance Settings Technical Specifications I (1) Generator load rejection w/o bypass event was evaluated statist ically (see Section 3.2.1) (2) 20% for calculations with point kinetics model (3) 25% for calculations with point kinetics model , varies slightly with power and fuel type
il 13 XN-NF-81-78 Revision 1 Table 3.2 Significant Parameter Values Used (contd.) I Safety / Relief Valve Performance (contd.) Pilot Safety / Relief Valve Capacity 150.83 lbm/sec Power Relief Valves Capacity 603.6 lbm/sec Safety Valves Capacity 1356.8 lbm/sec Pilot Operated Valve Delay / Stroke 0.4/0.1 sec 5 Power Operated Valves Delay / Stroke 0.65/0.2 sec MSIV Stroke Time 3.0 seconds 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) N rm 3 inches E
- 5 Range of Operation + 10 inches High Level Trip 42 inches Maximum Feedwater Runout Flow (3 pumps) 4966 lbm/sec Maximum Fcedwater Runout flow (2 pumps) 3310.67 lbm/sec Doppler Reactivity Coefficient (Nominal) .0026$/0F/ void fraction Void Reactivity Coefficient (Nominal) -15.895/ void fraction Scram Reactivity Worth Figure 3.1 Axial Power Distribution Figure 3.2 Delayed Neutron Fraction .005175 Prompt Neutron Lifetime 4.58 x 10-5 sec Recirculating Pump Trip Setpoint 1240 psig (vessel pressure)
- I I .
l !I i 14 XN-NF-81-78 j Revision 1 1 Table 3.3 Control Characteristics i !I l Sensor Time Constants i Pressure 0.1 sec Others 0.25 sec Feedwater Control Mode 1-element Feedwater Master Controller Proportional Band 100% Reset 5 repeats / min j 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% f Reset 8 repeats / min Speed Controller Settings i Proportional Band 350% Reset 20 repeats / min j Pressure Setpoint Adjustor Overall Gain 5 psi /% demand i
. Time Constant 15 sec Pressure Regulator Settings
- Lead 1.0 second l Lag 6.0 seconds l Gain 30 psid/100% demand e
!I
I 4 I 15 XN-NF-81-78 Revision 1
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co Figure 3.2 Axial Power Distribution
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- y. ., Figure 3.8 Increase in Feedwater Flow (Typical CPR)
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Figure 3.10 Loss of Feedwater Heating (Vessel Pressure and Level) 5
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!I 26 XN-NF-81-78 l Revision 1 I 4.0 MAXIMUM OVERPRESSURIZATION l 4.1 DESIGN BASIS l The reactor conditions used in the evaluation of the maximum 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 l power actuated relief valves were not available to vent steam as the lE ASME Pressure Vessel Code does not allow credit for power operated relief jE l valves. Also, the most critical active component (scram on MSIV closure) l was failed during the transient. 4.2 PRESSURIZATION TRANSIENTS ENC has evaluated several pressurization events, and has deter-mined that closure of all main steam isolation valves without direct l l scram is most limiting. Though the closure rate of the MSIV's is substantially I j slower than turbine stop or control valves, the compressibility of the fluid in the steam lines causes the severity of the slower closure to be greater. Essentially, the rate of steam velocity reduction is concentrated lI 1 toward the end of valve stroke, generating a substantial compression lg wave. Once the containment is isolated, the subsequent core power production
- E i
must be absorbed in a smaller volume than if turbine isolation occurred. Calculations have determined that the overall result is to cause containment ] i isolation to be more limiting than turbine isolation. 4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES This calculation assumed all four steam lines were isolated at 4 I the containment boundary within 3 seconds. Due to the valve characteristics 4 .
I 27 XN-NF-81-78 i Revision 1 I and steam compressibility, the vessel pressure response is not noted until about 3 seconds after beginning of valve stroke. Since scram performance was degraded to its Technical Specification limit for this analysis, effective power shutdown is delayed until after 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 pressurization. Pressures reach the recirculating pump trip setpoint (1240 psig) before i 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 1340 psia occurring near the vessel at about 6.75 seconds. The j maximum vessel pressure was 1361 psia occurring in the lower plenum at t about 6.75 seconds. Figures 4.1 and 4.2 illustrate the progression of the transient. The calculation was performed with ENC's advanced plant simulator t code, COTRANSA, which includes a one-dimensional neutronics model. l l I ,I t l 1 I 1
g-- g g3m mmm mm m M m-m M M M W"M ECO' i eu a' ;_ k LtVLL 1 (f1EUIYUTT1bx) 7 2 HEA FLUX 3 RECIRCUt.ATIpN FLOW
- 4. VESSEL STEnM FLOW l 500 .; . p. a APTER FL O'w i
\
100 f i a i kJ t-- C T300 LL O ru g-I co j 200 ^ T 1 I uJ 2 2 100 1 2 3 4 5 1 2 3 4 5 12 3 JM _5 5 \ 5 _ 5 5 5
~
4 4 4 2 3 _ 4 i l ~ MM \ E. .__ 1 1 1 i v : u i k
-lGL v.0 10 - N I 20 3.0 40 5.G G,0 70 9.0 9.0 10.C m TIME, SEC SEQ. CPTRN37 12/10/31 11.01.36 Figure 4.1 MSIV Closure without Direct Scram (Power and Flows)
M M M M M M WM M~M~ M M~M~M M~M~~M M M I I 250' ! ; 1 vtse. c t entbbUrn. LHLAL ibl j 2 KS EEL WATER LEVEL t IN 1 300 / 4 250 !
/ %
ECC j
/
1
=
150
- /
100 l l l 50 2 a a :o x 7 2 g 2 2 J_ z 2
- z 7.. ,
0 1 1 1 d l 8 8 0.C
~
10 2.0 30 40 5.G C.0 70 8.0 9.0 10.0 - co O TIME. SEC GEO. CPTRN3/ 12/10/31 11.01 36 Figure 4.2 MSIV Closure without Direct Scram (Vessel Pressure and Level)
I 30 XN-NF-81-78 Revision 1 - I
5.0 CONCLUSION
5.1 THERMAL MARGIN CALCULATIONS In the case of Dresden Unit 3, rapid turbine isolation events currently require the greatest thermal margin to prevent transient boiling transition. Turbine trip w/o bypass, though less limiting than load rejection w/o bypass, was examined. While not reported here to avoid redundasy, ENC will continue to examine turbine trip in future cycles. The feedwater flow increase event was ideltified for detailed evaluation due to the change in reactor conditions prior to turbine trip on high water level. Condenser bypass relief greatly mitigated the transient severity. ENC will continue to evaluate this transient in future cycles until i+ is clear that it is inherently less limiting than I turbine isolation events. Loss of feedwater heating, if of sufficient degree, should I provide enough reactivity to drive reactor core power to its maximum overpower trip setpoint. If heating loss occurs slowly over an extended period, the fuel clad heat flux will increase to nearly the same relative level as neutron flux. Thus, in this case, cycle-to-cycle variations in the results for this transient will be relatively small. Since the variations of the results of the other limiting events are expected to be larger, ENC will continue to examaine the loss of feedwater heating event in future cycles, to assure that it remains a non-limiting transient. 5.2 MAXIMUM PRESSURIZATION ENC expects to perform cycle specific evaluation of the capacity of Dresden-3's safety valves to protect against overpressurization. I
i !I i 31 XN-NF-81-78 j Revision 1 i !I Because the existing margin is small, ENL will evaluate this transient with COTRANSA, rattar than its simpler PTSBWR3 model. ig II I lI il !I !.I .I I il 4
l
\,
iI l 32 XN-NF-81-78 Revision 1 i l l
6.0 REFERENCES
'E (1) R.H. Kelley, " Exxon Nuclear Plant Transient Model for Jet Pump l5 Boiling Water Reactors", XN-NF-79-71(P), (plus supplements 1 and 2)
J.C. Chandler, "Dresden Unit 3 Cycle 8 Reload Analysis", XN-NF-81-76, t I (2) Revision 1, December 1981.
- (3) T.W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling
- WaterReactors",XN-NF-524(P), November 1979.
! (4) L .H. Steves, et. al ., "HUXY: A Generalized Multirod Heatup Code with E 10 CFR 50 Appendix K Heatup Option", XN-CC-33(A), Revision 1,
,5 November 1975.
T.W. Patten and F.T. Adams, " Methodology for Calculation of Pressure I (5) Drop in BWR Fuel Assemblies", XN-NF-79-59(P), October 1975.
! (6) K.R. Merckx, "RODEX2 Fuel Rod Thennal-Mechanical Response Evaluation Model, XN-NF-81-58(P), August 1981.
1 iI i
!I iI i
4
)I lI lI i
- I
!I . ; A-1 XN-N F-81 -78 Revision 1 I APPENDIX A i
DRESDEN-3 CYCLE 8 i ! SAFETY LIMIT CALCULATION PARAMETERS, l I INPUT VALUES, AND UNCERTAINTIES !I A.1 REACTOR SYSTEM UNCERTAINTIES l The reactor system uncertainties used in the Dresden-3 Cycle 8 safety limit calculation are the generic values listed in Table 5.1 of XN-NF-524(P)(3), A.2 FUEL RELATED UNCERTAINTIES l Fuel related uncertainties used in the Dresden-3 Cycle 8 safety limit calculation are listed in Table A-1. The values listed in Table A-1 are for Dresden-3 Cycle 8 with the exception of the XN-3 correlation 2 I uncertainty, which is generic. A.3 NOMINAL INPUT PARAMETER VALUES i Nominal values of input parameters used in the Dresden-3 Cycle 8 safety limit calculation are listed in Table A-2. I A.3.1 RADIAL POWER HISTOGRAM The radial power histogram used in the Drasden-3 Cycle 8 safety limit calculations is given in Figure A-1. The radial power histogram was chosen from a representative group of
- histograms. The histogram was then biased in a manner which would
! produce a worse (larger) value of the predicted safety limit. The peak
- I t
value for the histogram was chosen such that the limiting bundle MCPR would conservatively remain greater than the expected MCPR operating I
i A-2 XN-NF-81-78 Revision 1 limit under steady-state, full-power, full-flow conditions. f A.3.2 LOCAL PEAKING DISTRIBUTION The local peaking distribution used in the Dresden-3 Cycle 8 safety limit calculation is shown in Figure A-2. The local i j peaking distribution was chosen from the predicted distributions covering i l the range of Dresden-3 Cycle 8 exposures. The chosen distribution was used because it was found to produce the worst (largest) value of the safety limit of the group of distributions. P A.3.3 AXIAL POWER DISTRIBUTION f The axial power distribution used in the Dresden-3 Cycle I 8 safety limit calculations was: FA (X/L) = 0.30 + 1.10 sin (nX/L) l where X/L = relative axial position. This axial power distribution was chosen because it is conservative with i respect to the predicted axial power distributions of MCPR limiting i bundles. A.4 SAFETY LIMIT RESULTS l The final Dresden-3 Cycle 8 safety limit calculation used 500 Monte Carlo trials. The MCPR of the safety limit calculation using the nominal j input parameters was 1.05. With those conditions, the number of rods in i i the core which are expected to avoid boiling transition is greater than
- 99.9%. Thus, a safety limit of 1.05 for Dresden-3 Cycle 8 satisfies the requirement that at least 99.9% of the rods in the core must be expected to avoid boiling transition when the reactor is at the safety limit.
l ll .
a
- XN-NF-81-78 A-3 Revision 1 i
Tatile A-1 Dresden-3 Cycle 8 Safety Limit Fuel Related Uncertaintics lg Parameter Standard Deviation Assumed Probability !E (% of Nominal) Distribution Type XN-3 correlation 4.1 f.or mal l Assembly Radial Peaking 5.18 Normal Factor i Rod Local Peaking Factor 2.46 Normal Asserrbly Flow Rate 2.8 Normal 4 i lI lI 1 lI \1 !I lI
_ _ _ - _ . . - - - _ _ _ _ . _ _ _ . . . -. . - . . . . _ - - . _ - . = _ _ _. _ . _ _ . _ - - 4I 1
; A-4 XN-NF-81-13 1
l W '.sion 1 i j Table A-2 Dresden-3 Cycle 8 Safety Limit Nominal Input Parameters !I i I l Parameter Value j Core Pressure 1035 psia Core Power 3277 MW t , L I Core inlet Enthalpy 521.8 BTU /lbm i Total Core Flow 98.0 M1bm/hr l Feedwater Temperature 3200F Feedwater Flow Rate 12.4 Mlbm/hr 1 Hydraulic Demand Curve
- l G = 1.540 + (-8.851 x 10 )2 x LHGR 4
(1.908 x 10-3) X LHGR2 r I where G = Assembly Mass Flux [Mlbm/ft2 -hr] LHGR = Assembly power [kw/ft] I , I I I lI
)
5%
?~ -
M >a Eh8 s~ 5k8 - M 6
- 1 M 5 1
4 M 1 m a 3 r g o M 1 t s i 2 H 1 r e w o 1 P l 1 a M i d R a 0 O R 1 T C t A M 9 F i i m G N L 0 I y K t M 8 A e E P f a 0 S L A I 8 M 7 D A l e 0 R c y C 3 M 6 0 n e d s 5 e r M ,0 D 1 4 - M 0 A e r 3 u g i M 0 F
- 2
- 0 M
1
- 0 M
0 0 0 9 3 7 5 5 d 3 2 0 M 1 1 1 g 0%5m g bS5 M
I A-6 XN-NF-81-78 Revision 1 , I
- L : ML : ftL : !'
- fl : M : ML : ML :
- 1."6 : 1.92 : 0.02 : 1. W : 1.'6 : 1.08 : 0 . o .? : 1.01 :
I ____
- PL
- ML*
1 : H
- H : "La :
M
- ML I :
- 1. '] 8 : 0.90 : 1.03 : 1.30 : 0.98 : 0.73 : 1.C3 : J.92 :
I
- ML : N' ; H : h : H : H
~
ML* : r1 :
- 1.33 : 1 09 : 1.00 : 0.94 : 0.95 : 0.96 : 0.73 : 1.08 :
I . PL
- fi
- H
- H
- U
- H H
- M
- 1 01 : 1.Gn : 0.97 : 0.94 : 0.00 : 0.05 : 0.98 : 1.06 :
- ML : 1 : H : H : H : H
H : M :
- 1.C3 : 1.GA : 0.99 : 3.94 : 0.94 : 0.c4 : 1.00 : 1.08 :
I __ ____ ..._____
- ML
- ML
- M
- H
- H
- H M
ML : I : 1 08 : 0.M : 1.13 : 0.99 : 3.97 : 1.C0 :
- 1. 3 : 0.02 :
I L : "La : ML : M : M : M : ML* : ML :
- 1 07 : 1 01 : 0.73 : 1.:H : 1 06 : 1.09 : 0.oD : 1.02 :
I y I D
- LL : L : ML
- ML :
ML
- ML : ML :
L
- 1 01 : 1.?7 ; 1.98 : 1 03 : 1.01 : 1.03 : 1.08 : 1.06 :
I E
....._e._ __ e. au. _ _ . .._ _____ . . _ _
WIDE Figure A-2 Dresden-3 Cycle 8 Safety Limit Local Peaking I I .
1 I I XN-NF-81-78 Revision 1 15N3fdf* e l DRESDEN 3, CYCLE 8 PLANT TRANSIENT ANALYSIS REPORT , i DISTRIBUTION J l J.C. Chandler R.E. Collingham l G.C. Cooke , N.F. Fausz l L.J. Federico S.E. Jensen W.V. K,1yser R.l! Kelley J.E. Krajicek T.L. Krysinski J.L. Maryott J.N. Morgan I G.F. Owsley R.B. Stout L.C. O'Malley/ CECO (60) Document Control (5) I w I
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