ML20112H497

Cycle 2 Plant Transient Analysis

ML20112H497
Person / Time
Site:	Susquehanna
Issue date:	12/03/1984
From:	Collingham R, Jensen S, Keheley T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17139C813	List: ML20112H484 ML20112H497 ML20112H504 ML20112H515
References
XN-NF-84-118, NUDOCS 8501170166
Download: ML20112H497 (49)

Text

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XN -NF 118 l

SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS DECEMBER 1984 RICHLAND,WA 99352 ERON NUCLEAR COMPANY, INC.

wa8aema P PDR

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XN-NF-84-118 ISSUE DATE 12/3/84 SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS Prepare: .I_ - (4 lb LL T.H. Keheley~ g BWR Safety Analysis s l Concur: /z'w ybt

'S.E. Jensen, Lead Engineer h[/s-BWR Saf.efty Analysis Concur: -/F tef/u

'R.E. Collingham pManager f

BWR Safety Anajgsis Concur: '((M//u osn ,4-,

d.N. Morgan, A nager #

Customer Ser91ces Engineering

. ,1 Approve: ' (

R.B. Stout, Manager Licensing and Safety Engineering Approve: F ' b dC IY G.A. SofeWKanager/

Fuel Engineering and Technical Services

II XN-NF-84-118 i

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l NUCLEAR REGULATORY COMMISSION 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 Exxon Nuclear Company, Inc. It is being submitted by Exxon Nuclear to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the U.S. Nuclear Regulatory Commission in its ~ review of this report , -and by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Exxon Nuclear in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:

A.. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulrrss of the information contained in this document, or that to 'ise of any information, apparatus, method, or process disclosed in this document will not . infringe privately owned rights, or B. Assumes 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.

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XN-NF-84-118 TABLE OF CONTENTS Section Page Number Number

1.0 INTRODUCTION

............................................... 1 2.0 S U MM AR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN...................... 4 3.1 DESIGN BASIS.......................................... 4 3.2 ANTICIPATED TRANSIENTS................................ 5 3.2.1 Load Rejection Without Bypass.................. 5 3.2.2 Feedwater Cont rol l er Fail u re... ...... .. . ... ... . 6 3.2.3 Loss of Feedwater Heating...................... 7 3.3 CALCULATIONAL M0 DEL................................... 7 3.4 SAF ET Y L I M IT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.0 MAXIMUM OVER PRESSUR I ZAT 10N. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 4.1 DESIGN BASES.......................................... 20 4.2 PRESSURIZATION TRANSIENTS............................. 20 4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES............ 21

5.0 REFERENCES

................................................. 22 j

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XN-NF-84-118 i

I APPENDICES f I

A. MODIFICATIONS TO COTRANSA/PTSBWR3.......................... 23 B. PLANT SPECIFIC CHANGES TO COTRANSA/PTSBWR3................. 24 8.1 S U MMAR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 B.2 INSTRUMENTATION DELAYS................................ 25 i

B.3 THERMAL POWER M0NITOR................................. 26 B.4 HPCI L0GIC............................................ 27 B.5 RECIRCULATING PUMP TRIP............................... 28 B.6 FEEDWATER C0NTROLLER.................................. 29 B.7 RECIRCULATION FLOW CONTROL SYSTEM..................... 30 B.8 PRESSURE REGULATOR CONTROL SYSTEM..................... 31 B.9 COTRANSA HOT CHANNEL M0 DEL............................ 32 C. MC PR S AF ET Y L I M IT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 6 i C.1 INTRODUCTION.......................................... 36 C.2 AS S UM PT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 C.2.1 Design Basis Power Distribution................ 38 C.2.2 Hydraulic Demand Curve......................... 38 C.2.3 System Uncertainties........................... 38 j C.2.4 Fuel Rel ated Uncertai nti es ..... ....... ......... 38 l C.3 SAFETY LIMIT CALCULATION.............................. 40 l

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XN-NF-84-118 LIST OF TABLES Table Title Page 2.1 T h e rm a l Ma r g i n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l l

3.1 Design Reactor and Plant Conditions Susquehanna Unit 1..................................... 9 3.2 Significant Parameter Values Used in Ana lys i s Su squeha nna Uni t 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3 Results of Plant Transient Analysis....................13

. LIST OF FIGURES Figure Title Page 3.1 Load Rej ecti on Wi thout Byp as s ... ... .. . .. .. . .. . .. . .. . . . . 14 3.2 Load Rej ect i on Wi thout Bypa s s . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.3 Feedwater Controller Failure........................... 16 3.4 Feedwate r Cont rol l e r Fa i l u re .. . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.5 Loss of Feedwater Heating.............................. 18 3.6 Loss of Feedwater Heating.............................. 19

1 XN-NF-84-118

1.0 INTRODUCTION

This report presents the results of Exxon Nuclear Company's (ENC's)

Gvaluation of system transient events for Susquehanna Unit 1 during Cycle

,2 operation. The evaluation determines the necessary thermal margin (MCPR limits) to protect against the occurrence of boiling transition during the most limiting anticipated transient. The evaluation also demonstrates whether the vessel ' integrity would be protected during the most limiting pressurization event. The bases for these analyses have been provided in Reference 1.

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SUMMARY

Using ENC methodology and considering cycle 2 fuels, the most limiting plant system transient with regard to thermal marain at assumed bounding conditions was determined to be the turbine load rejection without bypass. The MCPR operating limits for potentially limiting plant system transient events are shown in Table 2.1 for comparison. The limiting transient was evaluated with all co-resident fuel types modelled and the most limiting fuel type was used to determine the reported MCPR's.

All the plant system transients reported herein were found to be less limiting than the control rod withdrawal error (CRWE) core transient event. The CRWE analysis and resulting MCPR operating limit are reported in Reference 2.

Maximum system pressure has been calculated for the containment isolation event, which is a rapid closure of all main steam isolation valves, using the senario as specified by the ASME Pressure Vessel Code.

This analysis shows that the safety valves of Susquehanna Unit I have sufficient capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110 per cent of design pressure (1.1 x 1250 = 1375 psig). The analysis also assumed six safety relief valves out of service. The maximum system pressures predicted -

during the event are shown in Table 2.1.

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3 XN-NF-84-118 1

Table 2.1 Thermal Margin (design basis conditions)

Transient ACPR/MCPR*

Load rejection without 0.19/1.25 bypass Turbine trip without 0.17/1.23 bypass Feedwater controller 0.17/1.23 failure Loss of feedwater 0.13/1.19 heating Inadvertent initiation 0.12/1.17 of HPCI

• Based on a safety limit MCPR of 1.06 Maximum Pressure (psig) l

+

Transient Vessel Dome Vessel Lower Plenum Steam Line MSIV closure 1303. 1318. 1302. ,

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a 4 XN-NF-84-118 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 DESIGN BASIS Consistent with the FSAR plant transient analysis, the thermal margin design basis condition was taken as the 104% power /100% flow operating point; The plant conditions for this point are as shown in Table 3.1.

t The most limiting point in cycle has been determined to be at end of full A

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.

Following requirements established in the Plant Operating License and associated Technical Specifications, observance of a MCPR operating limit aof.1.25 or greater protects against boiling transition during anticipated system plant transients from design basis conditions for Susquehanna Unit '

l- cycle 2.

The calculational models used to determine thermal margin include ENC's plant transient and core thermal-hydraulic codes as described in previous documentation (1,3-7).. Fuel pellet to clad gap conductances used in the analyses are based on calculations with RODEX2 (8). Recirculation pump coastdown was input based on measured Susquehanna Unit 1 startep test data, and the Susquehanna system transient model was benchmarked to

. appropriate Susquehanna Unit 1 startup test data. All transients were analyzed on a bounding basis. Table 3.2 summarizes the values used for

.important parameters to provide a bounding analysis.

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r- 1 5 XN-NF-84-118  !

3.2 ANTICIPATED TRANSIE J.i-l ENC considers eight categories of potential system transient occurrences for det Pump BWR's in XN-NF-79-71 (1). Three of the most limiting transients are described here in detail to show the thermal margin for Cycle 2 of Susquehanna Unit 1. These transients are:

o load rejection without bypass o feed water controller failure o loss of feedwater heating A summary of the transient analyses is shown in Table 3.3. Other plant transient events are inherently non-limiting or clearly bounded by one of the above events.

3.2.1 Load Rejection Without Bypass 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 recirculating pump trip (RPT). The compression wave produced by the fast control valve closure travels through the steam lines into the 1

)

vessel and creates the vessel pressurization. Condenser bypass flow, which could mitigate the pressurization effect, is not allowed. The excursion of core power due co void collapse is primarily terminated by j reactor scram and void growth due to RPT. Figures 3.1 and 3.2 depict the f 1

time varience of critical reactor and plant parameters during the load l rejection with bounding assumptions. These assumptions are:

o Technical Specification control rod speed

. o Technical Specification scram delay time

6 XN-NF-84-118 o void reactivity increased by 10%

o scram reactivity decreased by 20%

This results in a ACPR of 0.19 for the load rejection without bypass when RPT is operable.

The load rejection transient was then analyzed assuming the same bounding conditions but with RPT inoperable. This resulted in a ACPR of 0.23.

3.2.2 Feedwater Controller Failure

. Failure of the feedwater control system is postulated to lead to a maximum increase in feedwater flow through 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 vessel trip setting is exceeded. . To protect against

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spillover of subcooled water to the turbine, the turbine trips, 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 with bounding values and resulted in a ACPR of 0.17.

< Figures 3.3 and 3.4 present critical variables for this event. This event ,

was !also examined for reduced power conditions at full. flow; a ACPR of up Lto 0.22 was calculated which is bounded by the conrol rod withdrawal error event. Sensitivity results also show that the calculated ACPR is insensitive to the rate of feedwater flow increase.

l 7 XN-NF-84-118 I

This transient event was also analyzed assuming bounding condtions and failure of the bypass valves to open. This resulted in a ACPR of -

l 0.24. {

1 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 thermal power monitor system trip setpoint. The gradual power change allows fuel thermal response to maintain pace with the increase in neutron flux. For this analysis it was assumed that the initial feedwater temperature dropped 100 F linearly over a one minute period. The magnitude of the void reactivity feedback was assumed to be 10 per cent lower than expected to assure that the power response to l subcooling was gradual. Scram performance was conservatively assumed at its Technical Specification limit with scram worth 20 per cent below expected. A ACPR of 0.13 was calculated.

The bounding analysis was then repeated, but the bypass valves were assumed to be inoperable. The resulting ACPR remcined the same at 0.13.

3.3 CALCULATIONAL MODEL The plant transient code used to evaluate the generator, load rejection and feedwater flow increase was ENC's advanced code COTRANSA (1). This axial one-dimensional model predicted reactor power' shifts toward the core middle and top as pressurization occured. This was accounted for explicitly in determining thermal margin changes in the transient. The loss of feedwater heating event was evaluated with the

8 XN-NF ll8 code PTSBWR3 (1) since rapid pressurization and void collapse does not occur in this event. Appendix A delineates the changes made to COTRANSA (1) to merge the PTSBWR3 code with the COTRANSA code, to refine numerical techniques and to improve input. Appendix B deliniates the plant related changes made to these codes for the Susquehanna analysis.

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 fuel rods in the core. The safety limit is the minimum critical power ratio (MCPR) which would be permitted to occur during the liaiting anticipated operational cccurance. The MCPR . operating limit is derived by adding the change in critical power ratio ( CPR) of the limiting anticipated operation cccurance to the safety limit.

The safety limit for all fuel types in Susquehanna Unit 1 Cycle 2 was

<!atermined by the methodology presented in Reference 2 to have a value of 1.06. The input parameters and uncertainties used to establish the safety limit are presented in Appendix C.

9 XN-NF-84-118 Table 3.1 Design Reactor and Plant Conditions Susquehanna Unit 1 Reactor Thermal Power (104%) 3439 MWt

. Total Recirculating Flow (100%) 100.0 Mlb/hr Core Channel Flow 89.6 Mlb/hr Core Bypass Flow 10.4 Mlb/hr Core Inlet Enthalpy 518.0 BTV/lbm Vessel Pressures Steam Dome 1031 psia Upper Plenum 1049 psia Core 1058 psia

-Lower Plenum 1067 psia Turbine Pressure 974.7 psia Feedwater/ Steam Flow 14.15 Mlb/hr Feedwater Enthalpy 360.8 BTU /lbm Recirculating Pump Flow (per pump) 15.7 Mlb/hr

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10 XN-NF-84-118 Table 3.2 Significant Parameter Values Used in Analysis Susquehanna Unit 1 High Neutron Flux Trip 125.3%

Control Rod Insertion Time 3.5 sec/90% inserted Control Rod Worth 00% below nominal Void Reactivity Feedback. 10% above nominal (1)

Time to Deenergized Pilot Scram Solenoid Valves 200 msec (maximum)

Time To Sense Fast Turbine Control Valve Closure 30 msec Time From High Neutron Flux Trip To Control Rod Motion 290 msec Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke Time (Total) 70 msec Fuel / Cladding Gap Conductance t Core Average.(Constant) 595.5 BTU /hr-ft2-F

Safety / Relief Valve Performance Settings Technical Specifications Relief Valve Capacity 225.4 lbm/sec (1110 psig)

Pilot Operated Valve Delay / Stroke 400/150 msec

.(1) For rapid pressurization transients

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11 XN.NF.84-118 I

l Table 3.2 (continued)

Significant Parameter Values Used in Analysis Susquehanna Unit 1 MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint 90% open Condenser Bypass Valve Performance Total Capacity 936.11 lbm/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 58.7 in Normal 36.5 in Low level trip 8 in Maximum Feedwater Runout Flow Three pumps 4118 lbm/sec Doppler Reactivity Coefficient (1) -0.00260 /F-void fraction Void Reactivity Coefficient (1) 13.98 ~/ void fraction Effective Delayed Neutron Fraction 0.0052 Prompt Neutron Lifetime 0.0479 msec Recirculating Pump Trip Setpoint 1150 psig Vessel Pressure (1) Nominal value L. :

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12 XN-NF-84-118 j Table 3.2 (continued)

Significant Parameter Values Used in Analysis Susquehanna Unit 1 I

l Control Characteristics Sensor Time Constants Pressure 500 msec Others 250 msec Feedwater Control Mode Three-Element Feedwater Master Controller Proportional Gain 50.0(%/%)(%/ft)

Reset Rate 1.70 (%/sec/ft)

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 3.33 %/psid A

m n ii=m . . ..s.ii.,ammi.

Table 3.3 Results of System Plant Transient Analyses Maximum Maximum Core Average Maximum Neutron Flux Heat Flux System Event (% Rated) (% Rated) Pressure ACPR Load Rejection without Bypass 252 108.9 1203 .19 140 109.3 1191 .17 l

Feedwater Controller Failure i Loss of Feedwater Heating 120 113.2 1074 .13 j

l 405 127.5 1318 NA r.

l MSIV Closure with Flux Scram 9 Note: All events are bounding case.

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20 XN-NF-84-118 4.0 MAXIMUM OVERPRESSURIZATION

. Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenerio as specified by the ASME Pressure Vessel Code. This analysis showed that the safety valves of Susquehanna Unit I have sufficient capacity and performance to prevent pressure from reaching the established transient pressure 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 DESIGN 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 MSIV closure) was assumed to fail during the transient. The calculation was performed with ENC's advanced plant simulator code COTRANSA (1), which includes an axial one-dimensional

.neutronics model.

4.2 PRESSURIZATION TRANSIENTS ENC has evaluated several pressurization events and has deteremined that closure of all main steam isolation valves (MSIV's) without direct scram is the most limiting. Though the closure rate of the MSIV's is substantually slower than the turbine stop valves or turbine control valves, the compressibility of the additional fluid in the steam lines causes the severity of these faster closures to be less. Essentially, the j

21 XN-NF-84-118 1

rate of steam velocity reduction is concentrated toward the end of the valve stroke, generating a substantual compression wave. Once the containment is isolated the subsequent core power production must be absorbed in a smaller volume than if a turbine isolation had occured.

Calculations have determined that the overall result is to cause isolation (MSIV closures) to be more limiting than turbine isolations.

4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES This calculation assumed that six relief valves were out of -service and that all four steam isolation valves were isolated at the containment boundary within 3 seconds. At about 2.6 seconds, the reactor scram is initiated by reaching the high flux trip set points. Since scram performance was degraded to its Technical Specification limit, effective power shutdown is -delayed until' after 6.3 seconds. Thus, substantial thermal power production enhances pressurization. Pressures reach the recirculating pump trip setpoint (1135 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 po+mr. The maximum pressure calculated'in the steam lines was 1302 psia occurring l

.near the vessel at about 6.8 seconds. The maximum vessel pressure was 1318 psia occurring in the lower plenum at about 6.8 seconds.

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i 22 XN-NF-84-118

5.0 REFERENCES

1. R.H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2 (as supplemented), Exxon Nuclear Co., Inc., Richland, WA 99352 (November,1981)

2. S.E. Jensen, "Susquehanna Unit 1 Cycle 2 Reload Analysis, Design and Safety Analyses for ENC XN-1 8x8 Reload Fuel ," XN-NF-84-116, Exxon Nuclear Co., Inc., Richland, WA 99352 (December, 1984).

3. T.L. Krysinski and J.C. Chandler, " Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description," XN-NF-80-19(P), Volume 3, Revision 1, Exxon Nuclear Co., Inc., Richland, WA 99352 (April, 1981).

4. T.L. Krysinski et al ., " Exxon Nuclear Methodology for Boiling Water Reactors; Neutronics Methods for Design and Analysis,"

XN-NF-80-19(A), Volume 1, Exxon Nuclear Co., Inc., Richland, WA 99352 (May, 1980).

5. T.W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," XN-NF-524(P), Revision 0, Exxon Nuclear Co., Inc.,

Richland, WA 99352 (November, 1979).

6. R.H. Kelley, "Dresden Unit 3 Cycle 8 Plant Transient Analysis Report," XN-NF-81-78, Revision 1, Exxon Nuclear Co., Inc.,~ Richland, WA 99352 (December, 1981).

7. R.H. Kelley and N.F. Fausz, " Plant Transient Analysis for Dresden Unit 2, Cycle 9," XN-NF-82-84(P), Exxon Nuclear Co., Inc., Richland, WA 99352 (October, 1982).

8. K.R. Merckx, "RODEX2 Fuel Rod Mechanical Response Evaluation Model,"

XN-NF-81-58(A), Revision 2, Exxon Nuclear Co., Inc., Richland, WA 99352 (March, 1984).

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23 XN-NF-84-118 APPENDIX A MODIFICATIONS TO COTRANSA/PTSBWR3 COTRANSA originated with the coupling of a plant transient

' simulator code, PTSBWR3, and a one dimensional, coupled neutronic hydraulic code, COTRAN. Subsequent to the licensing of Dresden-2 the following modifications have been introduced into ENC's BWR plant transient model:

o. latest version of COTRAN replaced the original COTRAN o control system input module (CONTROL) introduced to coding o codes COTRAN, COTRANSA, PTSBWR3, and CONTROL all reside

'in the same program library The latest version of COTRAN (JUL83) replaced the original COTRAN because the numerical convergence

~ features.had been upgraded to increase code execution efficiency. A Control System Module has replaced the original control system model. so that til operations are handled through the input stream and may be easily tailored to the specific plant application. Having COTRAN, COTRANSA, PTSBWR3, and CONTROL in the same program library permits stand alone or grouped execution of each of the codes.

..m. _

24 XN-NF-84-113 APPENDIX B SUSQUEHANNA PLANT SPECIFIC FEATURES INCORPORATED INTO COTRANSA/PTSBWR3 0.1

SUMMARY

The ENC Diant transient analysis codes incorporate many plant specific features sucti as control systems through input and coding changes. This appendix discusses the required Susquehanna Unit 1 plant specific changes and control system input to the COTRANSA/PTSBWR3 plant simulator codes. These modifications were incorporated for exclusive use to Susquehanna Unit 1. Unless otherwise stated, modifications that were made to the PTSBWR3 portion of the code were paralleled in the COTRANSA portion.

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25 XN-NF-84-118 B.2 INSTRUMENTATION DELAYS Instrumentation delay refers to the finite time lag between a variable change and the sensing of that change by an instrument. The following instrument time constants were incorporated into the plant simulator codes for Susquehanna:

1) APRM - 30 milliseconds

2) Dome Pressure - 500 milliseconds

3) Turbine Pressure - 250 milliseconds

4) Downcomer Water Level - 250 milliseconds

5) Steam Line Flow Rate - 250 milliseconds

26 XN-NF-84-118 B.3 THERMAL POWER MONITOR The thermal power monitor (TPM) is designed to prevent reactor power increases above a conservative analytical limit of 120.0% of rated power, or, if operation is at off rated flow, then Pscram = (0.58

• Wpct + 0.647) where Pscram = flow biased scram setpoint Wpct = recirculating flow through drive lines as %

of rated The measured power is intentionally time delayed with a six second time constant to prevent spurious trips. If monitored power exceeds Pscram or 120% of rated power, the reactor scram systems trips. The logic used in the plant transient simulator is summarized below.

Aprm TPM Trip g TPM g Apr.m  :

g n'"8* Scram Drive *sa Flow 7

l u _ - _ - - -- -

27 XN-NF-84-118 B.4 HPCI LOGIC The plant transient simulator models allow the user to assume an startup of the High Pressure Coolant Injection System (HPCI) if the vessel water level falls below a setpoint specified by input. If so, the flow of HPCI water through the feed lines will increase to its maximum rate exponentially in four seconds. The maximum flow rate achieved will be 695 lb/sec at an enthalpy of 35 BTU /lb. The flow will be assumed to mix perfectly with the feedwater flow and enter the vessel at the top of the downcomer.

28 XN-NF-84-118 8.5 RECIRCULATING PUMP TRIP The recirculating pumps will automatically trip off on high vessel pressure or low downcomer level regardless of options chosen by the user.

Additionally, the user may select an input option which would trip the

~

pumps on indication of a turbine trip or a generator load rejection. When using this option, the user specifies the process and mechanical delay time between the initiating event of the pump trip (turbine trip or load rejection) and the time at which the pump rotor begins to physically decelerate. The user also specifies by input the setpoints for tripping the pumps on vessel pressure and level. .

29 XN-NF-84-118 B.6 FEEDWATER CONTROLLER Because of the modifications made to COTRANSA as dicussed in Appendix I A, the feedwater control system is incorporated into the code as part of the input stream. Figure B-1 gives a schematic of the control system used.

The feedwater control system maintains a pre-established level in the reactor vessel during normal plant operation by varying the speed of the  !

steam turbine driven feed pumps. Steam flow and feed flow are compared and an error signal is sent to the mismatch gain amplifier. The sensed reactor water level is compared to the level setpoint, this error signal is summed with the mismatch gain amplfier signal to provide the input signal to the flow controller. The flow controller provides the input to the function generator after going through an output limiter and a lead / lag compensator. The function generator signal is then sent to the turbine feed pumps.

p .

30 XN-NF-84-118 B.7 RECIRCULATION FLOW CONTROL SYSTEM Susquehanna Unit 1 is a based loaded plant, so the recirculation flow control' system will always be in manual control. With the new control system module no flow control system was modelled, the pump speed was manually controlled through input.

I i

_ _~ __ . - . =

31 XN-NF-84-118 8.8 PRESSURE-REGULATOR CONTROL SYSTEM As is discussed in Appendix A, the pressure regulator control system is entered. into COTRANSA/PTSBWR3 as input data. The model as it is input i

is shown in Figure B-2. Functionally, the pressure regulator adjusts turbine and bypass flow to maintain turbine pressure at a desired setpoint. Essentially, the system produces an error signal by comparing a sensed pressure with a pressure setpoint. This error signal is conditionea by the lead / lag characteristics of the control valve and produces a steam flow based on the pressure setpoint.

l 4

?:

r

s:  ;

i

,r 32 XN-NF-84-118 8.9 THE COTRANSA HOT CHANNEL MODEL l 1

L Q. , '

The existing COTRANSA hot channel model was modified to automate the iterative calculations using XCOBRA, R0DEX2, and HUXY which are normally used to evaluate the fuel thermal response to the transient conditions calculated by COTRANSA. This modification was made in order to expedite

,a the determination of the limiting transient for the cycle under

(" evaluation ; operating thermal margin requirements were determined through

g use of the procedures described in the ENC topical report on thermal limits methodology, XN-NF-80-19(P), Volume 3. Use of the hot channel

' /model to determine the limiting transient for a given application is i

? described in XN-NF-79-71(P), " Exxon Nuclear Plant - Transient Methodology for Boiling Water Reactors," Revision 2, dated November 1981.

. The original COTRANSA hot channel model, which was retained from the  ;

COTRAN core model integrated into the transient code, provided an

approximation of the transient critical-power ratio for a particular fuel assembly using the XN-3 correlation. Initial' conditions for the assembly were defined in the input.. The hot channel modification consists -of a determination of the initial power defining the limiting assembly, a transient csiculation of the coolant flow to the limiting assembly, and a calculation' of the critical power ratio of the limiting assembly. In this

+

applicstion, the limiting assembly is defined as' one which momentarily experiences calculated critical heat- flux (MCPR =1.00) during the transient, which is consistent with. the definition of the limiting assembly used in the plant transient methodology topical report.

L

33 XN-NF-84-118 The figure-of-ma-'t thermal margin requirements calculated by the hot channel model were all based on deterministic uncertainty values.

Comparison between COTRANSA hot channel analyses and iterative calculations with XCOBRA, RODEX2, and HUXY have indicated that both methods yield similar results. Any differences in critical power ratio values calculated by the two methods are bounded by the deterministic treatment of plant measurement uncertainties. Further comparative ,j analyses have shown the consistency between the results obtained with the

'XCOBRA-T code and results calculated by using the COTRANSA hot channel model. (XCOBRA-T is currently being documented in an ENC topical report.)

The use of the COTRANSA hot channel model with deterministic uncertainties provides a reasonable basis for~the determination of limiting transient conditions for further thermal margin analysis, s

C _ _ _ . _.. . .. . . . _ _ _ . . . . _ . . _ _ . . _ _ _ _ . . , ..

i f.evel Setpoint

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+ f+ ~

+

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m S

g'r3 +1  %- y 7

r,s + 1 V i r.S+1 y , l

, hs g i /-- + l Feed now Js

%tpt N Limit er g rp + i ,,,___ ,7 ,

Pumps c  : J3 T,S + 1 Lead / Log Output ,

ragg, Componector Limiter 5 Cenerator k l

Figure B-1 Feedwater Control System E j

t )lli 1 l , l

. ' ii il!, ! h - i , 9 E " b =m rL

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dr G M n e V E at L D i

L m

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36 XN-NF-84-118 I

l APPENDIX C MCPR SAFETY LIMIT C.1 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 99.9% 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 be establishing a minimum margin.to critical heat flux limits for steady state operation and calculating a transient effects allowance to assure that the steady state limit is protected during anticipated off-normal conditions. This appendix addresses the calculation of the minimum margin to steady state critical heat flux limits, which is implemented as the MCPR safety limit in the plant technical specifications. The transient effects allowance, or'the limiting transient del ta-C PR , is treated in the body of this riport.

The MCPR safety limit is established through statistical ccnsideration 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 c:rrelation, power peaking, and core coolant distribution, are fuel

1 37 XN-NF-84-118 related. When ENC fuel is introduced into a core where it will coreside 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 ENC-fabricated reload batch is

used to replace a group of dissimilar fuel assemblies, the core averagae fuel dependent parameters change because of the difference in the relative number of each type of bundle in the core, and the MCPR safety limit is ,

I again reevaluated.

, i The design basis power distribution is made up of components I

corresponding to representative radial, axial, and local peaking factors.

Where such data are available, these factors are determined through examination of operating data for previous ' cycles and predictions of operating conditions during the cycle being evaluated for the MCPR s'afety limit. If operating data are not available, either because the reactor has not been operated or because appropriate data cannot be supplied to ENC, the safety limit' power distribution can be determined strictly from the predicted operating conditions during the cycle being evaluated. Data

-for Susquehanna Unit 1 are limited to Cycle 1 operation,. which is' not considered typical of later cycle operation because of the differences in design between G.E. initial core fuel and both ENC and G.E. reload fuel designs.

n ,, , n .- - - - ,

-m. . - . -n,,,.,m

38 XN-NF-84-118 C.2 ASSUMPTIONS C.2.1 DESIGN BASIS POWER DISTRIBUTION The local power distribution used in the safety limit analysis is shown in Figure C-1. The axial power shape was conservatively assumed to be a center-peaked, truncated cosine distribution. The radial power distribution used in the safety limit analysis is shown in Figure C-2.

C.2.2 HYDRAULIC DEMAND CURVE Hydraulic demand curves based on calculations with XCOBRA were used in the safety limit analysis. The XCOBRA calculation is described in ENC topical reports XN-NF-79-59(A), " Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," anu XN-NF-512(A), "The XN-3 Critical Power Correlation."

C.2.3 SYSTEM UNCERTAINTIES System measurement uncertainties are not fuel dependent. The values reported by the NSSS supplier for these parameters remain valid for the insertion of ENC fuel . The values used in the safety limit analysis are tabulated in ENC topical report XN-NF-524(A), " Exxon Nuclear Critical Pcwer Methodology for Boiling Water Reactors."

, C.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 ENC topical report XN-NF-524(A), " Exxon

m-39 XN-NF-84-118 Nuclear Critical Power Methodology for Boiling Water Reactors." Power 1

measurement uncertainties are established in ENC topical report XN-NF-80-19(A), Volume 1, " Exxon Nuclear Methodolog'y for Boiling Water Reactors; Neutronics Methods for Design and Analysis." l

40 XN-NF-84-118 C.3 SAFETY LIMIT CALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ENC topical report XN-NF-524(A), " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." A minimum CPR value of 1.06 was ased in the analysis, and after 500 Monte Carlo trials it was determined that 6t least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95%.

r

.. .. ........ .... .. . .. . . ~ . . . . ~ .. . . _ . _ . _ . . . . . . . . ..., .

41 XN-NF-84-ll8

:  :  :  :  :  :  : i

LL : L : LM  : HM  : IM : HM : LM : L : l

0.93 0.97 1.02  : 1.06 : 1.06 : 1.09 : 1.08 : 1.06 :

L : HM : IM* H : IM : H : IM* a IM : l

0.97 : 1.04 0.88 : 1.03 : 0.92 : 1.06 : 0.94 : 1.08 :  !

:  :  :  :  :  :  :  : )

_________________________________________________________ )

a  :  :  :  :  :  :  : I

IM : H  : H  : H  : H  : H : E  : HM  :

1.01 : 1.07 : 0.99 : 0.97 0.98 : 0.99 : 1.08 : 1.09 : l 194  : E  : H : HM : W H : M  : E  :

1.06 : 1.03 : 0.97 : 0.89 : 0.00 : 0.98 4 0.92 : 1.06 :

HM : LM* H : W : HM : H : H : HM :

1.07 : 0.85 : 0.99 : 0.00 : 0.89 : 0.97 : 1.03 : 1.06 :

IM : H : H : H  : H  : H : Ih*

• LM  :

1.02 : 1.08 : 1.00 : 0.99 : 0.97 : 0.99 : 0.88 : 1.02 :

L : LM : H  : LM*

• H : H : HM : L :

0.97 : 0.97 : 1.08 : 0.85 : 1.03 : 1.07 : 1.04 : 0.97_:

W _________________________________. ______________________

I  :  :  :  :  :  :  :  :  :

t-t D  : LL : L  : IM : IE : HM : LM : L : LL :

E  : 0.93 0.97  : 1.02 : 1.07 : 1.06 : 1.01 : 0.97 : 0.93 :

WIDE Figure C-1 Susquehanna Unit 1 Cycle 2 Safety Limit Local Peaking Factors

y j DESIGN BASIS RADIRL POWER

80 -

4 4

60 -

i 1

v. -

i *

, g .

c l S -

l t 40 - 2 b

o 5 -

z l'

20 - -

i M ,

. . . . I k.

2  ;

L -

0 , , , , , , , -

t i

D.2 0.1 0.6 0. 8 1 1.2 1.4 1.6 -

i Bundle Power Factor  ;

Figure C-2 Susquehanna Unit 1 Cycle 2 Safety Limit Radial Power Histogram j i

i

l XN-NF-84-118.

ISSUE DATE 12/3/84 SUSQUEHANNA UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS Distribution J.C. Chandler R.E. Collingham S.F. Gaines R.G. Grummer S.E. Jensen T.H. Keheley J.E. Krajicek T.L. Krysinski J.N. Morgan L.A. Nielson H.G. Shaw/PP&L (40)

G.A. Sofer R.B. Stout H.E. Williamson C.H. Wu Document Control (5) i t

+

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