ML20106C437

Partial Feedwater Heating Operation Analysis

ML20106C437
Person / Time
Site:	Perry
Issue date:	01/31/1985
From:	CLEVELAND ELECTRIC ILLUMINATING CO.
To:
Shared Package
ML20106C428	List: ML20106C425 ML20106C437
References
NUDOCS 8502120261
Download: ML20106C437 (34)

Text

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ATTACHMENT PNPP PARTIAL FEEDWATER HEATING OPERATION ANALYSIS s

January 1985 4

Cleveland Electric Illuminating Company Perry 1 & 2 Nuclear Power Stations mi

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PY-CEI/NRR-0174L ATTACHMENT APPENDIX 15.D TABLE OF CONTENTS

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PAGE 15.D Partial Feedwater Temperature' Heating Operation 15.D.1-1

'15.D.1 Introduction and Summary 15.D.1 2 15.D.2 Fuel Integrity MCPR Operating Limit 15.D.2-1 15.D.2.1 . Abnormal Operating Transient 15.D.2-1

'15.D.2.2' Rod Withdrawal Error 15.D.2-2 15.D.3. Fuel Integrity - Stability.- 15.D.3-1

.15.D.4 Loss-of-Coolant Analysis 15.D.4-1 15.D.5 Containment Response Analysis 15.D.5-1

- 15.D.6.. Acoustic and Flow Induced Loads Impact on Internals 15.D.6-1 115.D.7 Feedwater Nozzle Fatigue Usage 15.D.7-1

'15.D-8 Feedwater Sparger-Impact Evaluation 15.D.8-1

, 15.D.9: Reactor. Protection System Setpoint 15.D.9-1 15.D.'10, Miscellaneous Impact Evaluation 15.D.10-1 15.D.10.1 - Feedwater System Piping 15.D.10-1 15.D.10.2- Impact on Anticipated Transients Without Scram'(ATWS) 15.D.10-1

.15.D.10.3 . Annulus Pressurization Load Impact- 15.D.10-1 15.D.10.4 Fuel Mechanical Performance .15.D.10-1 o 15.DJ11 Reference 15.D.il -

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LIST OF TABLES n- .

s NUMBER TITLE PfE iL

p 15.D-11 Sumary ~ of Transient Peak .Value Results - 15.D.2-3 PFH Operation E 15 '.D . Sumary of Critical Power Ratio Results - 15.D.2-4

'PFH Operation <

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.15.D-3 .Sumary'of Feedwater Sparger Fatigue Analysis 15.D.8-2 w

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PY-CEI/NRR-0174L ATTACHMENT LIST OF FIGURES NUMBER-- TITLE PAGE Load Rejection.With Bypass Failure,

- 15 2 D-l'- 15.D.2-5 370'F Rated WT

.i' 15.D-2 Load Rejection With Bypass Failure, 15.D.2-7 320*F Rated FWT

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15.D-3 -Load Rejection With Bypass Failure, 15.D.2-9 250*F Rated FWT

'15.D -Feedwater Controller Failure, 15.D.2-10 370*F Rated FWT 15 '. D-5? -Feedwater Controller Failure, 15.D.2-12 320*F Rated FWT x

115.D -Feedwater Controller Failure, 15.D.2-14' 250*F Rated FWT

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', PY-CEI/NRR-0174L ATTACHMENT

-15.D PARTIAL FEEDWATER HEATING (PFH) OPERATION 15.D. i . Introduction & Summary This section presents the results of a safety and impact evaluation for the

.' operation of the Perry Nuclear Power Plants (PNPP) with partial feedwater heating at a steady state conditions during the operating cycle and beyond the end of cycle conditions. This evaluation is performed on an equilibrium cycle basis and is applicable-to-its initial core and its suasequent reload cycles.

1.

The results of this evaluation justify PNPP operation at 100% thermal power steady state conditions with rated feedwater temperature ranging from 420*F to 320*F, and also beyond the end of cycle with rated feedwater temperature ranging from 420*F to 250'F.

Operation with partial feedwater-heating (PFH) occurs in the event that (i) certain stage (s) or string (s)-or individual heater becomes inoperable, or (ii)

-intentionally valving out the extraction steam to the feedwater heaters at the-end of~an operating cycle. Chapter 15 has already evaluated the consequence

.of the transient with a sudden feedwater temperature loss of 100*F when initiated from 420*F rated feedwater temperature. This appendix will justify the continued operation of PNPP at the steady state condition ranging from rated feedwater temperature of 420*F to 320*F during the operation cycle and, as low as 250*F beyond the normal operating cycle.

r Evaluations required to justify PFH operation include the abnormal operating transients,. thermal hydraulic stability, the critical feedwater nozzle and

, .sparger fatigue usage conditions and the worst-loss of coolant and containment response-conditions. The results are summarized below:

p a) .

The abnormal operating transients in Chapter 15 were re-evaluated to 1 determine the required operating MCPR limits for PFH operation. Accord-

, g ing to'the worst limit'ing trans'ient, the operating limit.MCPR needs to be-

'ncreased i by.0.01, that is, 1.19 for the initial' core and 1.20 for the reload core during operation when the rated feedwater temperature is between'370*F and 320'F. For operation beyond-the cycle ranging from-320*F to 250*F rated feedwater temperature, the operating limit MCPR 15.D.1-1 '

A

- - PY-CEI/NRR-0174L ATTACHMENT

~

needs to be increased by 0.03, that is, to 1.21 for the initial core and

~1.221for the reload core.

'b) The loss of coolant accident (LOCA) and containment response as described tin _ Chapter 6 were re-evaluated for PFH operating condition. It is found

.. g Thhat'the conditions with normal feedwater temperature at 420*F bound those at PFH conditions.

c)' Fuel-integrity was evaluated with respect to general design Criterion 12.

(1GJFR50, App. A). It-is shown that PFH operation satisfies the stabil-

-ity' criteria'and fuel integrity is not compromised.

d). The effect of acoustic and flow induced loads on the-reactor shroud,

. shroud support and jet pumps were re-investigated to assure that design

limits are not-exceeded. zThe effect of PFH on feedwater nozzle and spargerJfatigue usage factor.was examined. It was found that the in-creased-fatigue usage in 40 years still meets the acceptance criteria.

l There areLalso'other impact evaluations such-as_the feedwater piping, the effect of annulus' pressurization and the' consequences of Anticipated Transient-

~Without Scram (ATVS). These evaluations concluded that.the Perry design is

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~" adequate 'for PFHLoperation

. Operation with feedwater heater (s) out of service during~the operating cycle:and_ operation at end of cycle with_ final feedwater temperature reduction are acceptable for PNPP.

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", A PY-CEI/NRR-0174L ATTACHMENT L 15.D . 2 . Fuel' Integrity - MCPR Operating Limit 15.D.2.1c Abnormal Operating Transients i All abnormal operating transients in Chapter 15 were investigated for PFH

, operation. ;Three-limiting abnormal operating transients are discussed here in

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' detail. They are:

a) ~. Generator Load' Rejections with Bypass Failure (LRNBP)

b)" lFeedwater Flow Controller Failure (FWCF) c). - Loss of 100*F Feedwater Heating (LFWH)

, ThA evaluations were performed at 104.2% power, 100%' core flow with rated 1

J feedw'a'ter tiemperature of 370*F, 320*F and 250*F at end of equilibrium cycle.

d EPlant heat balance, core coolant hydraulic and nuclear transient data consis-

'tatiwIthFSARChapter15inputweredevelopedandusedintheanalyses.

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- , arc1 (FA) turbine control valve closure characteristics were assumed in the 4

_ analyses.

A, _

-The end;of! equilibrium cycle exposure point with all the control rods fully.

. - withdrawn is the most limiting point <in the cycle with the worst scram re-

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S activity worth' characteristics. A middle.of the cycle point (2000 MWD /T ibefore end'of-equilibrium cycle) was:also analyzedifor 370*F and 320*F rated

,w +M i ..feedwater temperatures;to demonstrate operation during.the operating cycle at-('

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- {thesefeedwater-. temperatures. This' point is-chosen because'it:is close enough'

' 9to end.of'. cycle such that the" scram'charactesistics have not been.aignificant -

'?lyfimproved"relativeto,earlierpoints1nthecyclebutthevoidreactivity.:

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1 characteristics are!different .tihan end oficycle. Scram characteristics are isignificantly improved at: exposure-lower than.this. point and the transient responses will'be bounded by the two point! analyzed. .It is shown that the'end

'of equilibriumicycle ~ condition . bounds th'e middle?of cycle. conditions.

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The; computer 'model' ~ described lis' Reference15.D.11-1:was used to simulate the

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Stransient fa):~and b); events. ' The"results : for.the bounding cases are summarized

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.TtblesL 15.D'1!andf15[D-2. :As"shown in Table 15.D-2, the operating MCPR-m, t

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^' PY-CEI/NRR-0174L ATTACIRIENT p

. limit'shall be l.19-(1.20 for reload core) for operation between rated feed-water temperature of 370*F and 320*F.

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Operation between 320*F and 250*F rated feedwater temperature requires a rated operating limit of 1.21 (1.22 for reload core)~ .

Lower initial operating pressure and 7 team flow rate (due to lower feedwater temperature) provide better overpressure protection for the limiting MSIV closure' flux scram event. Hence, it is concluded that the pressure barrier integrity.is maintained under partial feedwater-heating (PFH) conditions.

s The' transient responses for transients a) and b) are presented in Fig. 15.D-1 othrough 15.D-6.

The 100*F loss of feedwater heating transient was evaluated at 104.2% power, 100%' core flow with rated feedwater temperatures of 250*F-and 420*F at the end cofLequilibrium_ cycle using the computer model described in Ref. 15.D.11-2 and Lmethodology_ described in-Ref. 15.D.11-3. Results show that the 100*F loss of

'feedwater heating has less effect on colder feedwater than on the normal f feedwater temperature of ' 420*F. Thus, the ACPR results.for the case with l 250*F in'itial rated feedwater : temperature are bounded by the ~420*F rated

. normal case. Moreover, it'is'less likely to have's sudden 100*F loss at'an initial-feedwater temperature of 250*F.

1 115.D.2.2'~ Rod Withdrawal Error'

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. A rod l withdrawal. error' analysis case consistent with those documented in Appendix 15B (BWR 6 generic rod withdrawal' error analysis) was performed at-

, initial"feedwater temperature of 250*F to bound all rated ' feedwater tempera-J ture-conditions.' The' analysis indicated that the^ initial steady' state feed-

,- . water te'mperature-has negligible effect with regard to ACPR in a-random rod

< withdrawal error condition. 'Thus, the.ACPR values initiating from 250*F-

-feedwater tempersture condition fall within the statistical data base used to ,

, establish-the' Rod Withdrawal Limiter System setpoints. Therefore, the generic-

' ' ERod. Withdrawal Error. Analysis 1 adequately boun'ds PFH operation conditions.,

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, , PY-CEI/NRR-0174L ATTACHMENT Table 15.D-1 cc 4 Summary of Transient Peak Value Results 104.2% Power, 100% Core Flow

, Maximum

Expo ~ Rated Maximum Maximum Maximum Steam-sure Fdwtr Neutron Dome Vessel line

_ Point Temp. Flux Pressure Pressure Pressure iTransientL (MWD /T) (*F) '% (NBR) (psig) (psig) (psig)

E0EC* 250 235 1193 1221 1189 Load Rejec-tion.With E0EC ~320 246 1198 1224 1201

-Bypass Failure IE0EC '370- 245 1202 1230 1209 E0ECl 250 174 1128 '1150 1127 Feedwater

' Controller E0EC .320 139 1145 .1167 11451

' Failure E0EC: -370 144 1160 1187 1158

' *End of equilibrium' cycle

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PY-CEI/NRR-0174L ATTACHMENT Table 15.D-2 I L- Summary of Critical Power Ratio Results*

104.2% Power, 100% Core Flow

~ Feed End of Expo - water Req'd Tran sure' Temp. Initial sient

Transient 1 - Point (*F) MCPR ACPR MCPR

'E0EC+ 250 1.18 0.11 1.07 Load Rdjec-

tion Withe E0EC 320 1.18 0.11 1.07

-Bypass Failure.

'E0EC 370- 1.18 0.10 1.08 c-i:

E0EC '250 1.21 0.15 1.06 Fr edwater. *

Controller E0ECT 1320 ~1.19 0.13 1.06-

Failure 1 u- lE0EC 370J '1.18 10.11 :1.07 c*This table-is: applicable to initial: core with a safety limit MCPR of 1.06.

For app 1'ication1to reload core,= a 0.01: needs to be added.'

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PY-CEI/NRR-0174L A'ITACHMENT j ,.1

-15.D.3- Fuel Integrity - Stability

General' Design' Criterion 12 (10CFR50, Appendix A) states that power oscil-lat' ions which result in exceeding specified acceptable fuel design limits be either not possible or be readily and reliably detected and suppressed.

LHist'orich11y,compliancetoGDC-12wasdemonstratedbyassuringthatneutron 1 flux oscillations would not occur. This eliminated the need to perform fuel

~1ntegrity calculations under limit cycle conditions. As a result of stability

'. tests;at operating BWRs and extensive development and qualification of GE

. : analytical models,! stability criteria have been developed which also demon-

< _ strate compliance to GDC1 12. Reference'15.D.11-4 provides these stability compliance 1 criteria for GE. fueled BWRs operating in the vicinity of limit cycles. <

cw Operation in the partial feedwater heating (PFH) mode is bounded by the fuel '

integrity analyses in Reference 15.D.11-4s In general, the effect of reduced

-fesdwater temperature resultsiin a higher initial CPR which yields even' larger.

4 _ ;marginsLthan those reported in reference 15.D.11-4. The analyses are indepen-

dent of the stability margin since the reactor is already assumed in limit "E cycle oscillations.

Reference 15.D.11.4 also 'emonstrates d that for neutron-

~ flux-limit. cycle oscillations just below th's 120% neutron flux scram setpoint,

, 'fue11 design ~11mits are not. exceeded ~for those GE BWR fuel designs contained in s

GenerallElectric Standard Application for Reactor Fuel;(GESTAR, Reference- ,

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15.D.11-5). 'These; evaluations ~ demonstrate that substantial thermal / mechanical L' '

' margin _ is available for the GE BWR fuel' designs even-in the unlikely, event of.

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/ Jvery;1arge oscillablons.

, To provide assurance that acceptable' plant performance'is achieved'during ioperation in the -least stable region of the power / flow map, as well as.during-

all= plant maneuvering and operating states,- a generic set of operator.recommen-dations has_been developed and. communicated to all GE BWRs. These recommen-4

dations; instruct the operator.on'how to reliably detect and suppressL11miti

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l . ecycle-neutron flux oscillations should they occur._LThe recommendations were'

' developed to conservatively bound the expected-performance of all current.

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'E PY-CEI/NRR-0174L ATTACHMENT When operating in the partial feedwater heating mode during a cycle, the colder feedwater flow increases the core inlet subcooling and will also result in power distribution-changes. These changes result in reduced stability margin when operating in the high power / low flow region of the operating

' domain. ' Tests performed at an overseas BWR/6 in October, 1984 evaluated the effects of reduced feedwater temperature on stability margins. The result

shows that.the reduction in stability margin is within the conservative basis of the operator recommendations and therefore the recommendations are applic-able for partial feedwater heating during the cycle.

For operation at the end of the cycle with partial feedwater heating to extend the operating cycle, the power distribution approximates the target power

_shapei(typically a Haling power distribution) with all control rods fully

-withdrawn. Reducing the feedwater temperature at this point will result in an increased peak but at a higher elevation in the core. The change in power shape partia11y' offsets the reduced inlet enthalpy effect on stability and the result is a small change in stability margin. The change in stability margin is well within the conservative basis of the operator recommendations and therefore the recommendations are applicable to operation with PFH down to rated feedwater temperature of 250*F at the end' of cycle conditions.

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3 PY-CEI/NRR-0174L ATTACHMENT L15.D.4 Loss' of Coolant Accident Analysis c

.A Loss of Coolant. Analysis (LOCA) was performed for PNPP with PFH operation condition at-250*F rat'ed feedwater temperature. Reduction of feedwater temperature results in increased subcooling in the vessel thus increasing the mass flow. rate out of a LOCA break. However, an increase in initial total

-system mass and a delay-in lower plenum flashing also occur. They act to-

-:gether.to decrease the impact of increased flow out of the recirculation line break. As a result of this offsetting effect, the peak cladding temperature (PCT) was shown to be lower than the 2115'F value reported for PNPP and below the 2200*F'10CFR50.46 cladding temperature limit.

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. PY-CEI/NRR-0174L ATTACHMENT 15.D.5 :Lentaiament Response Analysis 1The impact of partial feedwater heating (PFH) on the containment LOCA response was evaluated. -Both' Main Steam Line (MSL) break and recirculation line break

- werel analyzed over.the entire. power / flow region. Reduced feedwater tempera-ture increases the subcooling of the coolant, and the mass flow rate from the postulated recirculation pipe break also increases, but is limited to the critica1Lflow of.the break. The final outcome is that the peak drywell and

' containment pressures under the partial feedwater heating conditions are boundedby-thedesignvalueshnFSARChapter6.

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ti PY-CEI/NRR-0174L ATTACHMENT J15.D.6' Acoustic Load and Flow Induced Loads Impact on Internals

' Acoustic loads are loads on vessel internals created by a sudden LOCA.

Acoustic. loading'Is proportional to total pressure wave amplitude to the

-vessel due'to LOCA.

N Loads are created on the shroud, shroud support and jet pumps due to high velocity flow in the downcomer in a postulated recirculation line break.

These flow-induced loads are affected by the critical mass flux rate out of the break. Partial feedwater. heating operation increases subcooling in the downcomer thus increases critical flow. However, PFH also increases density.

The reactor internals most impacted by acoustic and flow induced loads arc the shroud,-shroud support and jet pump. The impacts on these components were

evaluated over the operating power flow region. The analyses concluded that these components have'been designed to handle the loading during reduced

feedwater temperature conditions.

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(15.D.7, Feedwater Nozzle Fatigue Usage

.;An evaluation was performed on the PNPP feedwater nozzle with partial feed-

water heating'at rated feedwater. temperature of 250*F for conservatism. An

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18 month operating cycle with partial.feedwater heating based on an 80%

capacity factor is: equivalent to 438 full power days per cycle. This results a ' in anfadditional 0.'0214 fatigue usage factor over 40 years of continuous operation'at 250*F. .Furthermore, if we assume. additional end of cycle opera-tion.with feedwater temperature between 420*F and 250*F for 41 full power days

- perl cycle for 40. years the resultant fatigue usage factor would increase by

, 0.001. The total fatigue usage factor will still be less than 0.8, which is below the limit of 1.0.

The ' abo' v e. assumption ofl40 years of continuous partial feedwater heating operation is extremely conservative. The nozzle fatigue is expected to be much less.than the resultsl presented above. Hence, PFH operation is an

. acceptable mode even'for the most " fatigue-critical" vessel nozzle.

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.. - PY-CEI/NRR-0174L ATTACHMENT 15.D.8 Feedwater Sparger Impact Evaluation An evaluation was performed to examine the impact of partial feedwater heating operation on the feedwater sparger for PNPP. Six cases were analyzed to determine the' number of days allowable per year (for 40 years) for partial feedwater heating operation without exceeding the feedwater sparger fatigue usage < factor limit of 1.0. Results of this study are presented in Table 15.D-3.

This table' indicates the annual average number of days allowable for partial feedwater heating, reducing from normal 420*F to 370*F or to 320*F rated feedwater temperature with an additional 41 end of cycle days at 250*F. For

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, example, the feedwater sparger is designed to operate with 21 days of partial feedweter heating at rated 320*F during a fuel cycle and 41 days of partial Lfeedwater heating at rated 250*F beyond the end of the fuel cycle for every fuel cycle for 40 years. The feedwater sparger is acceptable for partial feedwater heating operation within these limits.

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,, PY-CEI/NRR-0174L ATTACHMENT Table 15.D-3 i

Summary of Feedwater Sparger Fatigue Analysis Feedwater Temperature reduction Allowable Number of Days per Year to-250*F for 41' days at for 40 Years at Feedwater End of each 18-Month Cycle for Temperature of 40 Years 370*F 320*F 3 Step- 127 21 7 Step 144 24 No'end of cycle reduction 256- 61 3 Step means ~3 average steps-of feedwater temperature reduction from 420'F to 370*F or 320*F.

7~ Step means ~7 average' steps of feedwater temperature reduction from 420*F l' -to.370*F-or 320*F.

This evaluation-assumes 70% capacity factor. Allowable number.cf days which

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results in a feedwater sparger fatigue usage factor of'1.0.

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,15.D.9 Reactor Protection System Setpoints At reac*'. power'leve7.s where significant amounts of steem are being generated, the fast closure of turbine stop or control valves will result in rapid reactor vessel pressurization._ When pressure-increases, power increases, especially if the bypass valves fail to open. For this reason, scram occurs

' on' turbine stop valve position and control valve fast closure to provide

-margin:to the core thermal-hydraulic safety limit. At low power levels high

, ' neutron flux scram, vessel pressure scram, and other normal scram functions provide sufficient protection. Therefore, below 40% rated power, turbine stop valve'and control valve scram functions are bypassed. The 40% NB rated power Lis-sensed through the direct measurement of the turbine first stage pressure.

nAs=feedwater temperature is reduced steam flow decreases. If the core thermal powerismaintai$edwithpartialfeedwaterheating,thesteamflowchange

'means that the turbine first-stage pressure versus power relatisnship is

altered. Thus, it is necessary to readjust turbine stop'and contro1' valve

scram bypass setpoints.(sensed from turbine first stage pressure)-for partial feedwater~ heating operation. A new setpoint is established for the trip units Jprior to commencement of each partial feedwater heating operation at each 1;; Loperating cycle.

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• ATTACHMENT F

, 15.D.10' Miscellaneous Impact Evaluation

, ,?15.D.10.1 Feedwater System Piping

- 1The impact'of partial feedwater' heating operation on the feedwater system

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piping up_to the first feedwater guide lug outside the containment has been

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evaluate'd for feedwater temperature at 250*F. Results of the study show that

'with the additional partial feedwaste heating operations, the feedwater piping fatigue usage factor still meets the allowable limit of 1.0.

.15.D.10.2' Impact on Anticipated Transient Without Scram (ATWS)

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,; ;An? impact. evaluation performed for PNPP shows that reducing feedwater tempera-cture helps to re' duce the consequences of an ATWS event. As a result of

? reduced:feedwater temperature, steam flow and core average void fraction are

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reduced. 'This.results in lower void coefficient and greater CPR margin which

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corresponds to milder trans1ents.

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15.D.10.3 Annulus! Pressurization Load-(APL) Impact

A boundary; analysis was performed lto' determine the impact of: partial feedwater -

37 heating. operation'on annulus pressurization ~ loads-(APL). -It is found that partial feedwater heating has-a.small-impaction annulu's pressurization loads

, Land.is}boundedbkthenormaloperationAPLlimits.

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?15.D.10.4 Fuel Mechanical Performance 4

Evaluations lwere' performed.to determine.the acceptability of PNPP partial' '

l ifeedwater heating operation on_GE-o; fuel rod and; assembly thermal / mechanical' g (performance._ Componenti pressureLdifferential .i fuel' rod overpower values

< sere determ'ined for anticipated operational? 2rrences with. partial feedwateri

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, " heating! conditions.. These values' were found t 'nunded by'those appliedlin tthe" fuel fod and'. assembly design bases and there PNPP with partial;

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ATTACHMENT s

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15.D.11= . References-

- 15.D.11-1 " Qualification of the One-Dimensional Core Transient Model for [

. Boiling Water Reactors" NEDO-24154 October 1978.

L15.D.11-2 "Three Dimensional BWR Core Simulator" NEDO-20953-A, January 1977.

15.D.11-3 Letter,.J. S. Charnley (GE) to F. J. Miraglia (NRC), " Loss of

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- Feedwater Heating Analysis",' July 5, 1983 (MFN-125-83).

L15.D.11-4 " Compliance.of.the General Electric Boiling Water Reac' tor Fuel Designs $toStabilityLicensingCriteria"NEDE-22277-P, December 1982.

15.D.11-5." General Electric Stan'ard d Application for Reactor Fuel" NEDE-24011-P-A,

. . January 1982.

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