ML20154A413

Extended Operating Domain/Equipment Out-of-Svc Analysis for Dresden Units 2 & 3

ML20154A413
Person / Time
Site:	Dresden
Issue date:	07/13/1988
From:	Keheley T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17201J425	List: ML20154A398 ML20154A402 ML20154A413
References
ANF-88-069, ANF-88-69, NUDOCS 8809120208
Download: ML20154A413 (53)

Text

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( ADVANCED NUCLEAR FUELS CORPORATION Q _

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EXTENDED OPER ATING DOM AIN / EQUIPMENT OUT OF SERVICE l

AN ALYSIS FOR DRESDEN

( UNITS 2 AND 3 l

JULY 1988 e

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( ADVANCEDNUCLEAR FUELS CORPORATION ANF-88-069 Issue Date: 7/13/88 EXTENDED OPERATING DOMAIN / EQUIPMENT ,

OUT OF SERVICE ANALYSIS FOR DRESDEN UNITS 2 AND 3

.i Prepared by

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T. H. Keheley V BWR Safety Analysis Licensing and Safety Engineering Fucl Engineering and Technical Services

)

I July 8, 1988

)

)

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CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARO NQ CONTENTS ANO USE OF THIS DOCUMENT

_PLEASE READ CAREPULLY Advanced Nucear Fuees Corporaton's warranties and reoresentatens cork comeg the suotect matter of inis docurnent are those set form in the Agreement between Advanced NucJeer Puote Corporspon and tne C stomer Dursuant to wrwen mas occurnent is is ec. Accoroegty, escoot as otherwise espreessy pro.

wood in sucn Agreement, neether Advanced Nucear Fueis Corporaten not any person acting on its cenait manoe any warranty or 'oorecentaten, encrossed or imched, entn resoect to the accuracy, comotetences. or usefumees of tr.4 infor.

mecon contamed in tnie occurnent. or that the use of any informaten. accaratus.

memod or process osecosed m this occument wW not etnnge pnvate<y owned ngnte: or assumes any Racihtee wem respect to tne use of any eformaton, ao.

paratus, momed or procese ciecomed in mis document.

The eformenon contamed nerem is for tne soie use of Catomer.

In orcer to avoid impairment of ngnts of .*4vanced Nuc ear Fuees Corporaton en potents or eventens wnich may os octuced m the informanon contamed m this documert, the remotent. ey its accootence of this document. agrees not to oucisen or mane puohc use (m the patent use of the term)of suen eformaten untd so autnormed in wrtog my Adve" l Nuc: ear Fuets Corocraten or untd after sin (6) months followeg termeate oesten of the aforesa,o Agreernent and any extensen thereof. uruess otrL 9 expressly crowded a tre Agreernent. No ngnts or nconsee e or to any paa. , are irnoped ey tee fumisneg of inis docu-ment.

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ANF-88-069 Page i TABLE OF CONTENTS Section Eagg

1.0 INTRODUCTION

.......................... ........................ 1

$ 2.0

SUMMARY

........................................................ 3 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN AT ICF AND FH00S/FFTR.... 4 3.1 Design Basis................................................... 4 3.2 Calculational Model............................................ 4 3.3 Anticipated Transients......................................... 5 3.3.1 Load Rejection Without Bypass Valve Operation.......... ....... 5 3.3.2 Feedwater Controller Failure................................... 6 4.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN DURING C0ASTDOWN......... 19 4.1 Design Bases................................................... 19 4.2 Calculational Mode 1............................................ 19 l 4.3 inticipated Transients......................................... 19 4.3.1 Load Rejection Without Bypass Valve Operation.................. 19

4.3.2 Feedwater Controller Failure................................... 20 5.0 TRANSIENT ANALYSIS FOR LOSS OF FEEDWATER HEATING............... 31 5.1 Design Basis................................................... 31 l 5.2 C al cul a t i o n al Mod e1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 5.3 Loss Of Feedwater Heating Transient............................ 31 6.0 MAXIHUM OVERPRESSURIZATION ANALYSIS............................ 34 6.1 Design Basis................................................... 34 l 6.2 Pressurization Events.......................................... 34 l 6.3 Closure Of All Main Steam Isolation Va1ves..................... 35

(

7.0 REFERENCES

............. ....................................... 37 l

l l

i ANF-88-069 Page 11 LIST OF TABLES Idtl1 EAEA 3.1 Reactor And Plant Conditions....................................... 7

)

3.2 Load Rejection Without Bypass...................................... 8 3.3 Feedwater Controller Fai?ure To Maximum Demant..................... 9 4.1 Load Rejection Without Bypass During Coastdown f Dresden Unit 3 Cycle 11............................................ 21 4.2 Load Rejection Without Bypass During Coastdown l

Dresden Unit 2 Cycle 12............................................ 22-4.3 Dresden 2 Cycle 9 Coastdown Core Follow Data....................... 23 j 4.4 Load Rejection Without Bypass During Coastdown 80% Powe r/10 8% F l ow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 4.5 Feedwater Controller Failure During Coastdown...................... 25 5.1 Loss Of Feedwater Heating.......................................... 33 s

6.1 ASME Overpressure Event............................................ 36 J'

I 4

L ANF-88-069 Page iii l

l l LIST OF FIGURES i

1 Fioure Eggg 1

1.1 Dresden Units 2/3 Operating Map.................................... 2 3.1 Load Rejection Without Bypass...................................... 10 3.2 Load Rejection Without Bypass...................................... 11 i 3.3 Load Rejection Without Bypass...................................... 12

)

3.4 Load Rejection Without Bypass...................................... 13

3.5 Feedwater Controller Failure....................................... 14 i 3.6 Feedwater Controller Failure....................................... 15 3.7 Feedwater Controller Failure....................................... 16 f 3.8 Feedwater Controller Failure............. ......................... 17 l 3.9 Feedwater Controller Failure....................................... 18 4.1 Load Rejection Without Bypass...................................... 26 4.2 Load Rejection Without Bypass...................................... 27 4.3 Load Rejection Without Bypass...................................... 28 4.4 Load Rejection Without Bypass...................................... 29 l

4.5 Load Rejection Without Bypass...................................... 30

)

l

)

ANF-88-069 Page 1

1.0 INTRODUCTION

I This report describes the plant transient analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of Increased Core Flow (ICF) to 108% of rated, Feedwater Heaters Out of Service and Final Feedwater Temperature Reduction (FH00S/FFTR), and Coastdown Operation for Dresden Units 2 and 3. Because Dresden Units 2 and 3 have equivalent physical systems from l a transient analysis viewpoint, conclusions drawn from these analyses are l generically applicable to both plants for present and future reloads of ANF fuel.

l The purpose of this document is to establish that the most limiting condition for operation of the Dresden units is at full power and increased f core flow. All future analyses will be performed at this operating condition.

i l

The analyses performed in this document were performed using the same l average core plant transient analysis methodology as used to calculate themal l rargin requirements for current operation of both Dresden units (Ref.1 and 2). The approved XCOBRA-T hot channel model determined the limiting change in l the Critical Power Ratio (delta CPR).

I

! This analysis supports operation in the expanded power and flow operating map shown in Figure 1.1.

ANF-88-069 Page 2 120 11 0 -

100/108 go. APRM Rod Line i

80-70 100% Rod Line E  : )

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FIGURE 1.1 ORESDEN UNITS 2/3 OPERATING MAP

(

ANF-88-069 Page 3 2.0 SUMARY The deterranation of thermal margin requirements for the Dresden units is based on the consideration of various operational transients. Reference 2 identifies tf.e limiting transients in each general category of events. The most limiting trans'ent avents for thermal margin in BWR/3 applications are the generator load rejection without bypass to the condenser, the loss of feedwater heating event, and the feedwater controller failure to maximum demand. The n.ost limiting event for the Dresden units is the generator load rejection without bypass valve operation.

Analyses assure that the MCPR Operating Limit protects all operating domains. The present operating map for the Dresden units allows flows up to 108% of rated. Operation with feedwater heaters out of service or with final feedwater temperature reduction increases operating flexibility. These two phenomena are functionally equivalent and are analyzed with a feedwater temperature reduction up to 100'F. Coastdown operation will extend the end of ,

the operating cycle. Each of these conditions were analyzed to determine the most limiting condition for operation for both Dresden units. By setting the Technical Specification limit on the most limiting condition, all others are bounded.

The most limiting condition for operation was established to be full power and increased core flow at aormal feedwater temperature. Therefore, all future analyses will be performed at these operating conditions. No specific pl , parameters need be checked on a per cycle basis to assure applicability of this report.

The closure of all main steam iso'ation valves (MSIVs) is the maximum system pressure event for ASME overpressure. MSIV closure is most limiting without activation of the MSIV position scram and without pressure relief credit for the four electromagnetic relief valves. The results of this analysis indicate that the ICF condition is the most limiting but stili within the requirements of the ASME code regarding vessel overpressure.

ANF-88-069 Page 4 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN AT ICF AND FH005/FFTR 3.1 D.gilan Basis Dresden Units 2 and 3 are sister plants with equivalent physical systems from a transient analysis viewpoint. Both plants contain cores of ANF 8x8 and ANF 9x9 fuel types. Limiting plant transient phenomena are a function of the

' physical characteristics of the p h.c' '.h e r than specific fuel types.

Therefore, the conclusions from .r ese analyses are applicable to both Dresden units. Reactor plant conditione for these analyses are shown in Table 3.1.

The most limiting point in the cycle is when the control rods are fully withdrawn from the core. The thermal margins established for the end oi /ull power capability are conservative for cases where control rods are partially inserted.

3.2 Calculational Maigl The average core plant transient methodology previously described in References 2 and 4 is updated in Appendix A of Reference 1 was used for the analysis reported ir this document. The delta CPRs were calculated with the approved XCOBRA-T (Rif. 5) hot channel model. A conservative integral power multiplier of 110% al plied in XCOBRA-T to pressurization transients accounted for COTRANSA code uncertainties.

The axial power shifts associated with the system overpressurization in

.he generator load rejection and the feedwater controller failure transients were modeled using the COTRANSA one dimensional core model. R00EX2 (Ref. 3) calculations determined conservative fuel pellet to clad gap conductances based on the Dresden units core configuration.

In accordance with ANF methodology, consistent bounding input is used to evaluate possible limiting transients. From these bounding results, the limiting transient is a generator load rejection without bypass valve operation.

a i

ANF-88-069 Page 5 3.3 Anticioated Transients Reference 2 generically considers eight major categories of system transients. For both Dresden units the limiting events are the generator load rejection without bypass, the feedwater controller failure, and the loss of feedwater heating. The generator load rejection and the feedwater controller '

failure to maximum demand have been evaluated for effects of increased core i flow and FH00S/FFTR. Analysis of FH005 and FFTR at increased core flow conservatively bounds FHOOS/FFTR at nominal core flow. These analyses assumed that a relief valve was out of service.'

3.3.1 Load Re.iection Without 8voass Valve Ooeration The load rejection without bypass valve operation (LRWB) is the most limiting of the rapid pressurization transients and of all the system I transients for the Dresden units. In the load rejection transient, the abrupt closure of the turbine control valve rapidly stops steam flow. The resulting pressure increase causes a decrease in void level in the core, which in turn creates a power excursion. This excursion is mitigated by Doppler broariening and pressure relief. However, the primary mechanisms for termination of the event are rod insertion and revoiding of the core.

The important parameters for this transient include the power transient (integral power) determined b/ the void reactivity and the control rod worth.

The void viactivity effects the power excursion rate and part of the intrinsic sht.tdown mechanism. The control rod worth determines the value of scram reactivity. Table 3.2 is a comparison of generator load rejection transients analyzed for a Dresden unit. The ICF case has the highest maximum neutron flux during the transient. Figures 3.1 through 3.4 compare important parameters for the ICF, FH00S/FFTR, and nominal rated cases. Figure 3.1 shows that the ICF case also has the largest integral power. The total core power produced during the transient for the ICF case was 3883 MW-sec and for the FH005/FFTR case it was 3535 MW sec. For comparison, the nominal case had a core power p 9 duction of 3640 MW-sec. This is because the ICF case results in a higher positive reactivity insertion rate during the pressurization portion

ANF-88-069 Page 6 of the transient. This results in the ICF case being the limiting pressurization evr at. This conclusion is valid for both Dresden Unit 2 and Unit 3.

3.3.2 ffJtdwater Controller Fathtg i Failure of the feedwater cof.rol system could lead to a maximum increase of feedwater flow into the reacter vessel. The excessive feedwater flow increases the subcooling in the recirculating water returning to the reactor core. This reduction in average moderator temperature results in a core power increase. Eventually, the increasing water level in the downcomer region will reach the high water level trip. The high level trip initiates a turbine trip to provent water from reaching the turbine. The turbine trip closes the turbine stop valves and the resulting scram arrests the power increase. The pressure pulse resulting from the stop valve closure is mitigated by opening the bypass valves to the condenser.

Figures 3.5 through 3.9 compare important parameters for the ICF, FH00S/FFTR, and the nominal cases. The total core power produced during the transient for the ICF case was 4421 MW-sec and for the FH005/FFTR case it was 3835 MW-sec. For comparison, the nominal case had a core power production of 4200 MW sec. The FH005/FFTR case has a lower integral power because of the lower steam dome pressure at the beginning of the transient. This lower steam dome pressu*e results in a larger rate of change of the feedwate flow.

Therefore, the high reactor vessel water level trip is reached earl:. .a in the ICF and nominal cases. Table 3.3 shows that the ICF case also has the maximum peak neutronic power. The ICF case is the limiting feedwater controller failure to maximum demand for both Dresden Units 2 and 3.

ANF-88-069 Page 7 l

TABLE 3.1 REACTOR AND PLANT CONDITIONS l

l PARAMETER NOMINAL FHOOS/FFTR ICE Reactor Power (MWt) 2527 2527 2527 Total Recirculating Flow (Mlb/hr) 98.0 105.8 105.8 Core Inlet Enthalpy (Btu /lbm) 522.3 513.0 524.0 B

Steam Dome Pressure (psia) 1020 1005 1020 Steam Flov (M1b/hr) 9.8 8.7 9.8 Feedwater Enthalpy (Btu /lb) 312.1 201.4 312.9 Recirculating Pump Flow (Mlb/hr) 17.1 18.5 18.5 ,

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ANF-88-069 Page 8 l

TABLE 3.2 LOAD REJECTION WITHOUT BYPASS l

Maximum Maximum Maximum Core Average Vessel Limiting Neutron Flux Heat Flux Pressure Fuel

% of Rated  % of Rated (osia) ACPR Nominal Conditions 350.9 121.0 1259 0.32 Increased Core Flow (ICF) 377.9 121.5 1260 0.33 ICF and Feedwater Heaters Out of Service (FHOOS) 236.6 120.1 1229 0.30 Note: All analyses performed with bounding (not statistically based) input.

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ANF-88-069 Page 9

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TABLE 3.3 FEEDWATER CONTROLLER FAILURE TO MAXIMUM DEMAND Maximum Maximum Maximum Core Average Vessel Limiting Neutron Flux Heat Flux Pressure Fuel 4

% of Rated  % of Rated (osia) ACPR Nominal Conditions 226.3 116.2 1198 0.21 Increased Core Flow (ICF) 240.5 117.0 1199 0.23 i ICF and Feedwater Heaters Out of Service (FH005) 226.5 117.6 1157 0.21 t

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ANF-88-069 -

Page 19

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4.0 TRANSIENT ANALYSIS FOR THEFJ4AL MARGIN DURING COASTDOWN 4.1 Desian Bases Economic considerations make operation of the plant past the end of full power capability desirable: this operation is coastdown. Because it occurs at the end of cycle, all rods are fully withdrawn. However, as the power decreases and flow is at its maximum, the axial power distribution shifts toward the top of the core. Two operational modes may exist, hold the reactor dome pressure at its rated condition or allow it to decrease. Both conditions are discussed.

4.2 Calculatignal Model The assumptions and models described in Sections 3.1 and 3.2 are applicable to the following discussions. The coastdown analyses used bounding input.

4.3 Anticioated Trantients -

As discussed in Section 3.3, the limiting events for the Dresden units are the load rejection without bypass and the feedwater controller failure to m3ximum demand. Because the phenomena that cause these events to be limiting are system related, changing core power will not affect which transients are limiting within each category. Comparisons of the load rejection without bypass and the feedwater controller failure transients show that the load rejection without bypass remains the limiting transient for coastdown '1 operation.

4.3.1 Load Reiection Without Bvoass Valve Ooeration Section 3.3.1 describes the phenomena occurring during a generator load rejection without bypass valve operation. The previous analyses demonstrated .

that the ICF case is the most limiting; therefore, this analysis was performed at ICF. Table 4.1 and Table 4.2 are comparisons of generator load rejection transients at increased core flow during coastdown operation allowing dome pressure to vary for Dresden Unit 3 Cycle 11 and for Dresden Unit 2 Cycle 12, ,

respectively. The analysis trend for the two plants is the same; there is a slight increase in delta CPR as the power decreases to 60%.

r ANF-88-069 Page 20 This trend is counter intuitive. That is, logic would say that because of the lower steam flow the pressurization portion of the transient should be less severe. Analyses at less than rated power and flow have always shown this (Ref. 1). Figures 4.1 through 4.5 present normalized comparisons of the generator load rejection at the 100% power and 80% power with 108% flow cases. The important factor to consider, however, is the axial power shift.

Shifting the axial power higher in the core results in a less effective scram. Figure 4.2 shows that on a normalized to initial power basis the 80%

power caso hh: a larger power increase in the top of the core than the 100%

power case. This results in the larger delta CPR when the hot channel is forced to reach critical heat flux. However, it should be noted that as the power decreases during coastdown, the margin to limits increase. Table 4.3 presents actual core follow data for Oresden Unit 2 Cycle 9. It is seen that although the delta CPR increases by 3.0% as the power decreases to 80%, the ,

margin increases almost 11% for about the same power change. Therefore, no special limits are required for coastdown operation.

Table 4.4 presents a comparison of the LRWB transient performed at 80%

power and 108% flow at rated dome pressure and reduced dome pressure. The reduced dome pressure in fact has a slightly larger delta CPR than the rated pressure case (0.005), but this is not seen due to the conservative rounding up of all delta CPR results. Therefore, maintaining the coastdown limits based on the reduced pressure case bounds the rated pressure case.

i 4.3.2 Feedwater Controller Failun Section 3.3.2 describes the feedwater controller failure to maximum demand. The analyses for coastdown were performed at ICF and ICF with FH005/FFTR. The results of these analyses are presented in Table 4.5.

Because the bypass valve operation mitigates the impact of the pressurization event, the feedwater controller failure is bounded by the load rejection without bypass for all cases.

mm . - -

ANF-88-069 Page 21 TABLE 4.1 LOAD REJECTION WITHOUT BYPASS DURING COASTDOWN ORESDEN UNIT 3 CYCLE 11 Maximus Maximum Maximum Core Average Vessel Limiting Neutron Flux Heat Flux Pressure Fuel Pover/ Flow  % of Rated  % of Rated (osia) ACPR 100%/108% 377.9 121.5 1260 0.33 80%/108% 286.4 98.1 1195 0.34 60%/108% 192.3 74.2 1125 0.35 40%/108% 97.4 48.7 1062 0.31

i ANF-88 069 Page 22

, TABLE 4.2 LOAD REJECTION WITHOUT BYPASS DURING COASTDOWN DRESDEN UNIT 2 CYCLE 12 Maximum Maximum Maximum Core Average Vessel Limiting Neutron Flux Heat Flux Pressure Fuel Power / Flow  % of Rated  % of Rated fosia) ACPR 100%/108% 392.5 120.3 1305 0.33 80%/108% 301.6 97.1 1194 0.34 60%/108% 194.2 73.8 1124 0.34 40%/108% 98.4 48.2 1062 0.30

. ~ . . - - - . _ _ _

k ANF-88 969 Page 23 f

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TABLE 4.3 DRESDEN 2 CYCLE 9 C0ASTDOWN CORE FOLLOW DATA [

p mwd /MT [Mt D1u.. Pressure Flow (M1b/hr) [Efi 7597 2334 1008 psia 98.0 1.65 7831 2188 1020 psia 97.3 1.74 8016 2131 1016 psia 97.7 1.79 8351 2050 1019 psia 93.3 1.83 L

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r ANF-88-069 Page 24 TABLE 4.4 LOAD REJECTION WITHOUT BYPASS DURING CDASTDOWN 80% POWER /108% FLOW 7 t

Maximum Maximum Maximum Core Avwrage Vessel Limiting '

Neutron Flux Heat Flux Pressure fuel Dome Pressure  % of Rated 1_gf Rated (nsia) ACPR i

1020 psia 261.0 97.7 1219 0.302 992 psia 286.4 98.1 1195 0.337 I

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ANF-88-069 Page 25 )

TABLE 4.5 FEEDWATER CONTROLLER FAILURE DURING COASTDOWN Maximum Maximum Maximum Core Average Vessel Limiting Neutron Flux Heat Flux Pressure Fuel Power / Flow  % of Rated  % of Rated fosial ACPR 100%/108% 240.5 117.0 1199 0.23 80%/108% 193.5 94.5 1120 0.23 60%/108% 116.3 65.8 1043 0.14 40%/108% 57.6 42.2 993 0.06 100%/108%* 226.5 117.6 1157 0.21 80%/108%* 193.8 94.7 1084 0.23 60%/108%* 102.5 64.4 1024 0.09 40%/108%* 53.7 41.9 990 0.04

• With FH00S/FFTR

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ANF 88 069 Page 31 5.0 TRANSIENT ANALYSIS FOR LOSS OF FEEDWATC MEATING 5.1 Desian Basis Plant transient analysis for the Dresden Units has shown that the most limiting event for an increase of recirculating vessel coolant subcooling is the loss of feedwater heating transient. The reactor plant conditions for the analysis are the nominal conditions shown in Table 3.1.

5.2 Calculational Model The core average plant transient methodology described in References 2 and 4 as updated in Appendix A of Reference I was used for the analysis of the loss of feedwater heating transient. The PTSBWR point kinetics model was used for core ar.d :ystem response. The fuel pellet to clad gap conductance values used in the analysis are based on RODEX2 for Dresden core configurations.

Because of the slow natwe of this event, the delta CPRs are determined using a quar'-steady-state analysis with XCOBRA.

This analysis is then compared to the XTGBWR analysis for the loss of i feedwater heating.

)

. 5.3 Loss Of Feedwater Heatino Transient The loss of feedwater heating leads to a gradual subcooling of the water in the lower planum. Core power slowly increases to the overpower trip set' point. The gradual power changc allows the fuel thermal response to '

I maintain pace with the increase in neutron flux. This analysis conservatively (

assumed that the feedwater temperature dropped 200'F over a two-minute period.

Void reactivity is assumed to be 25% more negative than the nominal value, which results in a maximum value of power and heat flux. Scram performance is assumed to be 20% less than the nomina' value.

Table 5.1 presents a sumary of the loss of feedwater heati' g analysis results using PTSBWR. The Dresden Unit 2 Cycle 12 analysis also was performed using the three dimensional nodal core simulator code XTGBWR (Ref 6). The result of this analysis 's a delta CPR of 0.14. As is seen from Table 5.1, if

1 ANF-88-069 Page 3.'

the delta CPR for the loss of feedwater heating is set to 0.20 for both Dresden Units, all past and future cycles will be bounded.

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ANF-88 069 Page 33 TABLE 5.1 LOSS OF FEEDWATER HEATING Dresden Unit Cycle Void Reactivity (1/ void fraction)* Limitina ACPR 3 8 -15.89 0.16 3 9 -15.81 0.16 3 10 -15.14 0.20 3 11 -16.55 0.19 2 9 -16.40 0.16 2 10 -16.40 0.20

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6.0 MAXIMUM OVERPRESSURIZATION ANALYSIS This section describes the analysis of the maximum overpressurization accident performed to assure maximum vess31 pressure will not exceed 110% of the design value for compliance with the ASME code.

C.1 Desian Basis The evaluation of the maximum pressurization event was performed with the recctor conditions summarized in Table 3.1. The same bounding conditions as those used in the transient analysis were assumed. In addition, fitrther conservatism was added by not allowing the operation of the four power actuated relief valves as required by the ASME.

I 6.2 Pressurization Event 1 The general catecories of expected maximum pressurization events are partial or total isolation of the turbine or containment and loss of offsite power. Generally, the condition in which the greatest energy is generated within the smallest confinement will result in the maximum pressurization.

Previous analyses have determined that the maximum pressurization transients for the Dresden units is the inadvertent closure of all MSIVs with failure of direct scram (Ref. 7). The position scram, which comands reactor shutdown upon MSly movement, mitigates the effects of this event to the point that it does not contribute to the determination of thermal margins. Delaying the scram until the high pressure trip setpoint is reached results in a substantially more severe transient.

Although the closure rate of the MSIVs is substantially slower than that of the turbine stop or control valves, the compressibility of the fluid in the steam lines provides significant damping of the compression wave associated with the turbine trip events to the point that the slower MSIV closure without the direct scram results in nearly as severe a compression wave. Once the ,

containment is isolated, the subsequent core power production must be contained in a smaller volume than the associated turbiae trip events.

ANF 88 069 Page 35 Analyses have demonstrated that the containment isolation event under these conservative assumptions results in a higher overpressure than total isolation of the turbine.

6.3 Closure Of All Main Steam Isolation Valves This calculation assumed that all four steam lines were isolated at the containment boundary within three seconds. The valve characteristics and steam compressibility combine to delay the arrival of the compression wave at the core until approximately three seconds from the initiation of the MSIV stroke.

Table 6.1 presents the results of the comparison case for ICF, ICF and FHOOS, and full power part flow. The most limiting conditions for this event are the increased core flow and normal feedwater temperature case. This conclusion is valid for both Dresden units.

ANF-88 069 Page 36 TABLE 6.1 ASME OVERPRESSURE EVENT Power / Flow Maximum Neutronic Maximum Heat Maximum Vessel -

% Flux (% rated) Flux (% rated) Pressure fosia) 100/108 439.0 133.9 1324 100/108 FH005 360.5 129.2 1304 100/87 412.2 129.2 1324 l l I i

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ANF 88 069 Page 37 7.0 REFERENr.ES

1. T. H. Keheley, "Dresden Unit 3 Cycle 10 Plant Transient Analysis," XN NF.

85 62, Exxon Nuclear Co., Inc., Richland, WA 99352 September 1985.

2. 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.

3. K. R. Merckx, "R00EX2 Fuel Rod Thermal Mechanical Response Evaluation Model," XN-NF 81-58(A), Revision 2. Exxon Nuclear Co., Inc., Richland, WA 99352, March 1984.

4. J. A. White, "Exxon Nuclear Methodology For Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description," XN NF 80-19fPifA), Volume 3 Revision 2, Exxon Nuclear Co., Inc., Richland, WA 99352, January 1987.

5. "XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," XN-NF-84-105f P), Volume 1 and Supplements 1 and 2 Exxon Nuclear Co., Inc., Richland, WA 99352, June 1985.

6. D. J. Braun, "Dresden Unit 2 Cycle 12 Plant Transient Analysis," ANF 88-Qil, Advanced Nuclear Fuels Corporation, Richland, WA 99352, May 1988.

7. 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.

ANF-88 069 Issue Date: 7/13/88 EXTENDED OPERATING DOMAIN / EQUIPMENT OUT OF SERVICE ANALYSIS FOR DRESDEN UNITS 2 AND 3 DISTRIBUTION O J. Braun G. J. Busselman M. E. Byram R. E. Collingham M. J. Hibbard T. H. Keheley T. L. Krysinski J. L. Maryott R. S. Reynolds D. F. Richey G. L. Ritter R. H. Schutt G. A. Sofer H. E. Williamson Ceco /J. M. Ross (

Document Control

. s Ahncnmerit '

g s GE Nuclear Energy

+.,i.  :- cr.

1:  :.

July 26, 1988 '

cc: W. F. Naughton REP:88 161 Mr. R. A. Roehl Supervising Fuel Buyer COPMONWEALTH EDISON COMPANY Fuel Department, 234 E P. O. Box 767 Chicago, IL 60690

SUBJECT:

Correction to Dresden 2 Cycle 12 Alternate Water Chemistry LTA's MPLHGR Curve

REFERENCES:

1. Test and Inspection Agreement between Commonwealth Edison Company and General Electric Company dated May  ;

28, 1975.

2. Letter from R. A. Roehl to R. E. Parr, "Dresden 2 Cycle 12 Alternate Water Chemistry LTA's Exposure Limits " RAR:88-193, May 19, 1988.

3. Letter from R. E. Parr to k. A. Roehl, (Telecopied July 19,1988) ' Reevaluation of Dresden 2 Cycle 12 Alternate Water Chemistry LTA's Exposure Limits,' REP:88 159, July 19, 1988.

4. Loss of Coolant Accident Analysis For Quad Cities Units 1 & 2 and Dresden Units 2 & 3, NEDO 24146 Rev 1. April 1979.

t ATTACHMENT $:

Corrected MAPLHGR Curve for Dresden 2 Alternate Water Chemistry LTA Bundles  ;

Dear Mr. Roehl:

I The attached MAPLHGR curve is the corrected version of the composite l limiting MAPLHGR curve for the Dresden HWC LTA's. Please replace the MAPLHGR curve which was sent to you in Reference 3 with the attached  !

curve. The corrected MAPLHGR curve is consistent with the GE LOCA analysis for the Dresden Unit 2 HWC LTA's (Reference 4).

Please note that the attached information is proprietary to the General Electric Company and should be controlled pursuant to Article XVI!! >

(Proprietary Data And Access) of the January 6,1986 Contract.  !

n _ _ - - - - _ _-_ ~ . - - - , . , . _ _ _. A

u Very truly yours,

6 ^

R. E. Parr Senior Fuel Project Manager Edison Projects WC 174; (408) 925 6526

. s, SENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE DRESDEN UNIT 2 CYCLE 12 LTA BUNOLES LY5455, LY5456, LY5457, LY5458 AVERAGE PLANAR MAPLHGR EXPOSURE (KW/FT)

(GWd/ST) 0.2 11.5 1.0 11.6 5.0 11.9 10.0 12.1 15.0 12.1 20.0 11.9 25.0 11.3 30.0 10.7 35.0 10.2 41.6 4.8

4 ,

4 ? .

Attachment 6 Discussion of Previous SER Topics i

The following discusses several topics raised in previous generic and reload SERs for Dresden and is based on information previded by Advanced Nuclear Tuels.  ;

1. Rod Bow Considerations I During the review of the previous reload submittal, the NRC SER placed an exposure cap on ANF 8x8 and 9x9 fuel due to rod bow considerations. The limit was set at 30,000 mwd /Mt for ANF 8x8 fuel and 23,000 ffWd/Mt for ANF 9x9 fuel (batch average exposure). This condition was eliminated for both ANT 6xS and 9x9 fuel when ANF received the NRC SER for XN NT-82-06, Sup-plement 1, Revision 2 in May of 1988. This SER, which references mechanical design analyses of 35,000 mwd /Mt peak assembly exposure for ANF 8x8 and i 40,000 mwd /Mt peak assembly expcsure for ANT 9x9 fuel in XN NF-85-67(P)(A), [

Revision 1, establishes ANF's currently approved mechanical design limits.

2. Extended Burnup Approval Status (XN NX-82-06)

Qualification of Exxon Nuclear Fuel for Extended Burnup (XN NF-82-06(P)(A) w/ Supplements 2, 4, and 5, Revision 1) was approved in July of 1986 by the NRC. Extended burnup qualification for 9x9 fuel (XN NT-82-06(P)(A) Sup-plement 1, Revision 2) was approved in May of 1988 by the NRC.

3. Conditions on Critical Power Methods (XN NF-524, Rev. 1) f In the Safety Evaluation prepared by the Core Performance Branch covering Revision i to XN NF-524(P)(A), "Exxon Nuclear critical Power Methodology  ;

for Boiling Water Reactors", the NRC Staff identified four conditions to i be met during the application of the subject methology under the generic  !

approval granted by the SER. The steps taken during the analysis to assure l compliance with these conditions are described below.  !

l CONDITION 1: Each plant specific application must contain the data used i to generate the uncertainties employed in the methology. i The uncertainties used in the Dresden Unit 2 MCPR safety limit calculations are the same as the uncertainties which have been used in pervious Dresden  ;

analyses. The two loop uncertainties are discussed below; the single loop ,

uncertainties are the same except as described in the Cycle 11 reports. l Plant measurement uncertainties which are not fuel-dependent were taken from approved NS$$ supplier generic documents applicable to Dresden. As ,

identified in XN NT-524(P)(A), Revision 1, specific uncertainty values used  ;

in the analysis were a feedwater flow rate uncertainty of 1.76), a feedwater '

terperature uncertainty of 0.76%, a core pressure uncertainty of 0.5%, and i

1 I

a total core flow rate uncertainty of 2.5%. The generic core inlet tem-perature uncertainty of 2.0% was conservatively replaced with an uncer-tainty of 2.4% on the core inlet enthalpy. These approved values were used as one-sigma uncertainties consistent with NEDO-24011. The nominal values, oncertainties, and statistical treatment of these measured plant parameters are summarized in Table 1.

The uncertainties associated with the XN-3 Critical Power Correlation are based on data contained in MiNF-512(P)(A), Revision 1, and XN NF 734(P)(A). The safety limit analysis was based on a one sigma un-certainty value of 4.11% for the XN 3 correlation, consistent with the source documents noted above and with XN NT-80-19(P)(A), Volume 4, Revision 1, which provides a generie description of the overall reload analysis.

The correlation statistics were develeped from the ANF's CHF data base, which includes test geometries which encompass the Dresden 8x8 and 9x9 fuel designs. XN NF-734(P)(A) was issued explicitly to validate the XN-3 sta-tistics for application to 9x9 fuel.

Power distribution measurement uncertainties are based on data contained in XN NF 80-19(P)(A), Volume 1. The safety limit calculation was based en one-sigma uncertainies of 5.28% on radial peaking factor and 2.46% on local peaking factor consistent with the reference report. These uncertainties were develeped based on analytical predictions of measured data for BWR fuel. The same methods used for the analytical predictions were used for the nuclear design analyses for Dresden, hence the generic uncertainty values are applicable to Dresden.

The correlation and power distribution measurement uncertainties and their statistical treatment for the Dresden analysis are summarized in Table 2.

CONDITION 2: All plant parameters that are not statistically convoluted must be placed at their limiting value.

In the performance of plant transient analyses, ANT uses design values for major process parameters for consistency with the TSAR analyses which are superseded by the ANT transient analyses. Design values are established by the plant designer as conservative predictions of the boundaries of the plant operating envelope, and may not be accurate predictions of actual plant cperation. These vilues are used to assure a censervative calculation of the transient effects. Nominal values are best estimate predictions of plant operating conditions. The use of nominal conditions is appropriate for the statistically treated parameters in the Monte Carlo analysis.

Input to the Monte Carlo calculation censists of three major classifications of data heat balance information, pswer distribution information, and fuel geometric information.

Heat balance information consists of feedwater temperature and flow rate, core pressure and total flow rate, and core inlet enthalpy. All of these variables are considered statistically in the Mente Carlo analysis.

Fower distribution information is taken from the fuel management analysis and consists of radial, axial, and local peaking factors. Radial and local 2

peaking factors are considered statistically in the Monte Carlo analysis.

For power distributions characterized by bottom peaked core average axial power shapes, a limiting center peaked axial distribution is used. 1 Fuel geometric information consists of fuel dimensions and hydraulic demand cu rve s. Small variations in fuel dimensions with manufacturing tolerances are considered in the ANT pressure drop methodology and contribute to the flow distributien uncertainty. Hydraulic demand curves are used to deter-mine fuel assembly flow rates as a function of bundle powers individual assembly flow rates are treated statistically in the Monte Carlo analysis.

CONDITION 3: Each application should demonstrate that the uncertainties in plant parameters are treated with at least a 95% probability at a 95%

confidence level in accordance with Acceptance Criterion 1.0 of Standard Review Plan Section 4.4 The magnitude and nature of the uncertainties used in the Monte Carlo analysis have been established generically during the staff review of ANF topical report KN-NT 524(P)(A), Revision 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors". A detailed review of the XN 3 correlation statistics was included in the review of ANT topical report XN NT-80-19(P)(A), Volume 1, "Neutronies Methods for Design and Analysis".

The conclusion that these uncertainties may be conservatively treated as normally distributed was addressed during the generic review.

Uncertainties in the measurement of plant parameters were taken from the NSSS supplier's generic reload submittal. Based on the Staff's approval of these uncertainties for use in the MCFR safety limit calculation, the ANF analyses used the published values as 95% confidence statistics.

Process measurement uncertainties are generally characterised by a normal distribution thetsfore, a normal distribution was used in the ANT analy-sis.

The Monte Carlo analysis was performed to demonstrate that during sustained operation at the M0FR safety limit, at least 99.9% of the fuel rods in the core would be expected to Lvoid boiling transition at a confidence level of 95%. This conclusion conservatively assures that the boiling transition limitation will be protected during anticipated operational occurrences in which the MCPR safety limit is protected. The reft enced Standard Review Plan section identifies this method as an acceptable approach to the 95/95 treatment of uncertainties.

CORDIT!cN 4 Each application must present a goodness-of tit analysis for the fitting of the Fearson curve in order to assure that the number of Monte Carlo trials used in establishing the safety limit MCFR are sufficient.

In the original ANT MCPR safety limit methodology, the first four statis-tical moments of the Monte Carlo output were used to define an output frequency distributien through fitting of Fearson functions. This approach was take to minimize the number of trials necessary in the Monte Carlo analysis. Revision 1 to XN-NF-524 abandoned this approach in favor of a distribution-independent eethod of assigning tolerance limits. The new 3

1 approach required a larger number of Monte Carlo trials, but the end result was a conclusion which was independent of the Pearson functions.

Since the statistical analysis involved no fitting of standard functions to the Monte Carlo output, no goodness-of-fit analysis was provided. In the case of the Dresden analysis, 500 Monte Carlo trials were provided.

In the non-parametric tables, an expected value may be established at a confidence level of 95% with as few as 50 trials.

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Table 1 Plant Measurement Uncertainties Nominal Uncertainty Statistical Parameter Units value  % Nominal Treatment Feedwater Flowrate M1bs/hr 12.41* 1.76 Convoluted Feedwater Temperature deg F 340.1 0.76 Convoluted Core Pressure psia 1035 0.50 Convoluted Total Core Flow MLBM/hr 98.0 2.50 Convoluted Core Initt Temperature 0.20 Replaced by core inlet enthalpy Core Inlet Enthalpy Stu/lbm 522.3 0.24 convoluted Core Power MW 3200* Allowed to vary with heat balance

• Feedwater flowrate and core power were increased above design values to attain desired core MCPR for safety limit evaluation, consistent with XN NF-524(P)(A) Revision 1.

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i Table 2 Fuel-Related Uncertainties Source Uncertainty Statistical Parameter Document  % Nominal Treatment XN-3 Correlation XN NF 512(P)(A) 4.11 Convoluted XN NF-734(P)(A)

Radial Peaking Factor XN NF-80-19(P)(A)- 5.24 Convoluted Volume 1 Local Peaking Factor XN NT-8C 19(P)(A) 2.46 Convoluted Volume 1 Axial Peaking Factor XN NF-80 19(P)(A) 2.99 Limiting Volume ! Value Assembly Flowrate XN NF-79-59(P)(A) 2.80 Convoluted l

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