ML20141E805

Rev 2 to Analysis to Extend Operator Action Time for Alternate Shutdown Panels in Support of Fitzpatrick Compliance to App R

ML20141E805
Person / Time
Site:	FitzPatrick
Issue date:	11/30/1985
From:	Cornwell K, Hoffman E, Sozzi G GENERAL ELECTRIC CO.
To:
Shared Package
ML20141E797	List: ML20141E792 ML20141E805
References
DRF-C61-00045, DRF-C61-45, MDE-137-0585, MDE-137-0585-R02, MDE-137-585, MDE-137-585-R2, NUDOCS 8601080333
Download: ML20141E805 (24)

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MDE-137-0585 Revision 2 DRF C61-00045

, November 1985 ANALYSIS TO EXTEND OPERATOR ACTION TIME FOR ALTERNATE SHUTDOWN PANELS IN SUPPORT OF FITZPATRICK COMPLIANCE TO APPENDIX R

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Prepared by: 67Mi T K.F. Cornwell, Engineer Application Analysis Services Verified by: '. .3 ^

E. H, Hoffmann, Engineer Application Analysis Services

._ Approved by:

G.L'. Sozzi,fMinaf6f '

Application Analysis Services Approved by: i A.E. Rogers, Manager Application Engineering Services

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n I B601080333 851217 PDR ADOCK 05000333 F Ppg GENERAL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTDC COMPANY e 175 CURTNER AVENUE

• SAN JOSE, CAUFORNIA 95125

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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT i

i i Please Read Carefully i-The only undertakings of General Electric Company respecting informa-tion in this document are contained in the contract between the custe=cr nnd General Electric Ccmpany, as identified in Line putcimme order for this report and nothing contained in this document shall be l construed as changing the contract. The use of this -information by

anyone other than the customer or for any purpose other than that for ,

which it is intended, is no.t authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to.the completeness, accuracy, or usefulness of the information contained in this document.

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CONTENTS

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1. INTRODUCTION 1-1

2.

SUMMARY

AND CONCLUSIONS 2-1

3. EVENT DESCRIPTION 3-1

4. ANALYSIS 4-1 4.1 Overview of Analysis Method 4-1 4.2 Event Sinulation 4-2 4.3 Fuel Integrity Evaluation 4-3 4.4 Suppression Pool Integrity Evaluation 4-4 4.5 SRV Operability Evaluation 4-5

5. RESULTS 5-1 5.1 Fuel Integrity 5-1 5.2 Suppression Pool Integrity 5-1 5.3 SRV Operability 5-2

6. REFERENCES 6-1 4

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1. INTRODUCTION In 1982, General Electric (GE) performed an analysis (Reference 1) to

assist New York Power Authority (NYPA) in demonstrating adequate shutdown capability in compliance with the requirements of 10CFR50.48 and 10CFR50 Appendix R at the James A. FitzPatrick Nuclear Power

Plant. This analysis determined that the maximum time for . the operator to regain control of the reactor shutdown functions at the alternate shutdown ~ panels for a control room fire event was ten minutes. The basis for the analysis performed in 1982 was to ensure that the core would remain covered with a two phase mixture during the entire event. Since that time, the Nuclear Regulatory Commission (NRC) has issued further clarification (Reference 2) that a "no fuel damage" criterion is an acceptable alternative basis for evaluating the performance of the alternative shutdown systems for BWRs which .

utilize the Automatic Depressurization Systems (ADS) and Low Pressure

, Coolant Injection (LPCI).

The analysis documented herein was undertaken to revise the previous analysis with the objective of determining the operator action time based upon this new criterion to' assure no fuel damage. A goal of a thirty minute operator action time was mutually agreed upon by GE and NYPA during the initial portion of the analysis. In addition, two other acceptance criteria ' were considered in the evaluation of the plant response. First, the operator action time should be such that the suppression pool temperature is low -enough to achieve cold shutdown following the event. Second, the fluid conditions at the

. inlet of the. safety relief valves (SRVs) should not impair- SRV operability. The methodology, analyses, and results that satisfy these criteria are documented in this report.

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SUMMARY

AND CONCLUSIONS To assure safe shutdown and to demonstrate compliance with 10CFR50.48 and 10CFR50 Appendix R. an analysis was performed to extend the time allowed for operator actions at the alternate shutdown panels of the James A. FitzPatrick Nuclear Power Plant. The analysis investigated

, the potential consequences of operator action time on fuel integrity, suppression pool integrity and SRV operability. The operator action time is the time when the operator initiates cold shutdown action at the alternate shutdown panels with up to seven SRVs, one Residual Heat Removal (RHR) pump aligned in the LPCI mode, one RER heat exchanger, and.one RHR service water pump. The analysis concludes t-that a .30 minute operator action time is sufficient to prevent fuel damage, maintain suppression pool integrity such that cold shutdown can be ach'ieved and avoid any conditions adverse to SRV operation.

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3. EVENT DESCRIPTION Th,is section describes the worst case event scenario for the highly unlikely event of an uncontrolled fire in the control room, cable spreading room, or the relay room. The fire is postulated to be severe enough to cause personnel evacuation of the control room and loss of control in the control room due to a fire in the relay room or cable spreading room. Under these circumstances, plant shutdown will be performed at a remote station which has the manual control of

.seven SRVs, one RHR pump aligned in the LPCI mode, and one RHR heat exchanger and service water pump which are used to achieve cold shutdown.

At the start of the event the reactor is assumed to be operating at full power, normal water level, and steady state conditions. As soon as the operator decides that the fire cannot be extinguished immediately and controls can not be maintained in the control room, he manually scrams the reactor, trips the main turbine, closes the Main Steam Isolation Valves (MSIVs), verifies scram, and evacuates the control room. This set of manual actions can be performed easily and quickly.

Immediately after scram and isolation, the reactor pressure increase is limited by the SRVs operating in the pressure actuation mode. The function of SRVs in this mode is not affected by the fire, since the SRVs are located in the inerted containment and they function in a mechanical mode which does not rely on external power. There is no potential for reactor overpressurization because the SRVs are sized to accomodate this type of isolation event. There is no potential for fuel damage because this event is similar to the transient event

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of MSIV closure analyzed in the FSAR.

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The reactor isolation also ~ causes a loss of steam supply to the turbine . driven feedwater pumps which results in a loss of feedwater flow to the reactor. Normally, other high pressure makeup systems, e.g., High Pressure Coolant Injection (HPCI), and Reactor Core Isolation Cooling (RCIC), will operate to maintain water inventory when substantial reactor inventory is lost due to the relief valve actuations. However, for the limiting case, these high pressure systems are assumed to be inoperable and the rate of reactor inventory loss is maximized. This " boil off" continues with the reactor maintained at high pressure (~100,0 psig) with inventory loss through the pressure actuation mode of the SRVs until the operator can initiate manual actions at the remote station.

I As soon as the operator arrives at the remote station he establishes communications, verifies controls and low pressure ECC I system availability. He then manually opens several of the SRVs to i reduce vessel pressure. The pressure decrease causes flashing'of the saturated water and results in an increase ~ in reactor water level during the blowdown. Eventually, the water level drops as the depressurization continues. When the reactor pressure drops below the shutoff head of the RHR system, the RHR pump, operating in the LPCI mode, begins to inj ect subcooled water into the reactor. The subcooled. water injection may collapse the steam- voids 'in the two-phase mixture which may result in a momentary drop in water level. However, the high rate of coolant injection will rapidly replenish the inventory . and reflood the core. (It should be noted that the above scenario conservatively assumes that the automatic-initiation of the core spray system and other RHR pumps would also be j inoperative and hence not be available to maintain core inventory.)

The operator will continue to allow the reactor to depressurize and I

allow the LPCI flow to refill the system. Eventually the water level will rise to the steam liras and spill out the SRV lines back into-the ~ suppression pool. This mode is referred to as the alternate shutdown cooling path. -

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4. ANALYSIS 4 . ,1 OVERVIEW OF ANALYSIS METHOD Three separate evaluations were performed to determine the potential consequences of the operator action time on fuel integrity, suppression pool integrity and SRV operability. The impact on fuel integrity was evaluated because of the potential for' uncovering the core and subsequent fuel heatup. The operability of the SRVs was investigated because following core reflood the vessel is allowed to continue to refill until the water level reaches the steamlines and liquid is discharged to the suppression pool through the SRVa.

Finally, this event may lead to significant suppression pool heatup because the RHR heat exchanger is not assummed to be aligned in the alternate shutdown path until af ter the reactor vessel is refilled and liquid is discharged through the SRVs.

The fuel integrity evaluation was performed by determining the duration of the care uncovery and the resulting peak cladding temperature (PCT). These calculations were performed utilizing the NRC approved GE evaluation model (SAFE) and the CE core heatup analysis code (CHASTE). The details of these calculations are provided in subsequent sections.

The integrity of the suppression pool was evaluated by determining that the peak temperature and pressure are below design conditions, and that these conditions provide adequate Net Positive Suction Head (NPSH) for the RHR pumps such that cold shutdown can be achieved.

These calculations were performed by first determining the quantity of energy released to the suppression pool through SRV actuations util,izing the GE evaluation model (SAFE). Then an energy balance was performed to determine the peak pool temperature due to the energy added by SRV actuations and the energy removed by the RHR heat exchanger during the shutdown process. From the peak suppression -

pool conditions the available NPSH can be determined and compared to the required value to verify that cold shutdown can be achieved.

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The SRV operability was demonstrated by reviewing the ability of the SRVs to discharge reactor water at low pressure and the acceptability of, the loads imposed on the SRV piping and . supports during this discharge. Because there is a potential for the SRVs to open with liquid in the steamlines, the acceptability of the loads was assessed by conservatively assuming that the SRVs open instantaneously.

Further a range of upstream conditions was considered that included saturated vater and subcooled liquid upstream of the SRVs. The operability of the SRVs during low pressure subcooled liquid discharge has been demonstrated experimentally and found to be acceptable as documented in Reference 3. The forces on the discharge piping and supports were evaluated for the liquid diacharga and compared to the design loads from normal steam discharge. The fluid conditions during SRV discharge were obtained from the GE evaluation model (SAFE).

4.2 EVENT SIMULATION The NRC approved GE licensing evaluation model (SAFE)* was used to obtain the system response for each of the analyses. The basic analysis assumptions are consistent with 10CFR50.48, 10CFR50 Appendix R and the emergency procedure guidelines (EPG). The key model or input assumptions which are applicable to all phases of the analyses are summarized below:

1) The reactor is assumed to be operating at 100% power, and at normal water level at the time of event initiation.

2) The reactor is assumed to be manually scrammed at event initiation.

3) . The MSIVs begin to close at event . initiation due -to manual

-closure.

8 SAFE is the code normally used for 10CFR50.46 LOCA/ECCS conformance evaluations.

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4) Feedwater flow is assumed to ramp to zero in five seconds, two seconds after the MSIVs start to close. (This represents a

, . nominal feedwater flow characteristic typically used in ,

. licensing analyses for plants with turbine driven feedwater pumps.)

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5) The 1979 ANS decay heat correlation is used to realistically model decay heat (Reference 4).

6) The SRVs are assumed to actuate ' at their nominal pressure setpoints, with discharge flowrates at their nominal nameplate rating.

, 7) There are seven SRVs available at the remote station. However, only six were assumed available.

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8) There is no spurious operation of any component.

The postulated. event was analyzed utilizing an operator action time of. 30 minutes. The additional' assumptions and . inputs necessary to obtain the appropriate system response .for each of the three i

evaluations are listed in each of the following respective sections.

] 4.3 FUEL INTEGRITY EVALUATION To evaluate the impact of the-30 minute operator action time on both.

4 fuel integrity and suppression pool integrity the system response was obtained from the SAFE code. Based upon this response.a fuel heatup analysis was performed utilizing the GE ~ core heatup analysis code

-(CHASTE).

In this analysis the cladding temperature was conser a-tively computed without credit for any heat transfer during the time 4

the fuel is uncovered. Realistically, steam cooling due to boil off and flashing would prevent any significant cladding heatup'during

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this time period. If the maximum cladding temperature predicted by this analysis is less than the lowest temperature at which cladding perforations are expected - (approximately 1500*F), then the no fuel damage criterion will be met. The results of this evaluation are contained in Section 5.1.

4.4 SUPPRESSION POOL INTEGRITY EVALUATION-The ' initial conditions and key parameters utilized in addition to

those listed in Section.4.2 are summarized below

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1) The initial temperature of the suppression pool is 95'F and the pressure is 14.7 psia, per FitzPatrick technical specifications.

2) The RHR heat exchanger is initiated at the time the SRVs begin discharging liquid.

3) The RHR heat exchanger effectiveness is 133.8 Btu /sec *F I

corresponding to one RHR pump and one RHR service water pump.

(Reference 5) 1

4) The service water temperature is 77*F (" maximum normal" value i per Reference 5).

The SAFE code was used to predict the system response to the I

postulated f' ire event._ From this response the energy added to the suppression pool due to the SRV discharge was obtained. An energy balance was then performed to determine the temperature response of the suppression pool. Additionally, the pressure in the' suppression

] pool corresponding to the peak temperature was estimated by assuming

{ that,the liquid and air were in thermodynamic equilibrium _ and that j-the mass of air in the containment remained constant.

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Two criteria can be used to judge the acceptability of the suppres-sion pool response. First, the temperature and pressure in the pool must remain below the design conditions of 220*F and 56 ' psig.

Second, to achieve cold shutdown the fluid conditions in the suppres-sion pool must provide adequate NPSH to the RHR pump in order to prevent the pump from cavitating. The NPSH available to the pump is a function of both the pressure losses in the piping and the fluid conditions in the suppression pool and can be determined from the following relation:

NPSH (Available) =

p A +HVEL + "Z -

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P = pressure in suppression pool air space HVEL = Vel city Head (negligible)

H7 = Elevation Head H

= Head Loss

.P VAP

= Vapor Pressure The elevation head and head loss terms were obtained from References 6, 7, and 8. The other terms were determined from the fluid conditions in the suppression pool. The results of this evaluation

.are presented in Section 5.2.

4.5 SRV OPERABILITY EVALUATION The alternate shutdown cooling mode leads to the reactor pressure vessel being filled ' with liquid to the elevation of the steamlines.

Once the steamline elevation is reached the fluid is returned to the -

suppression pool through the SRVs and SRV discharge lines. The SRV operability concerns arise due to the magnitude of the potential 4-5

loads which could occur if the SRVs are suddenly opened with subcooled water upstream. To conservatively bound all combinations of ypstream conditions it has been assumed that the conditions of the fluid entering the steamline are also imposed at the SRV inlet.

Actually, the steamlines will initially be filled with stcam when two phase and subsequently, single phase liquid spills into the steamlines. Both two phase fluid and saturated liquid upstream of the SRVs would tend to produce lower reaction forces than those produced by subcooled liquid, given the same upstream pressure (Reference 9). However, to conservatively bound all potential i.

upstream conditions a range of fluid inlet conditions varying from saturated to subcooled liquid was evaluated. The maximum subcooling was obtained by assuming that once LPCI is initiated it remains on, at full capacity, until the water level in the vessel rises to the steamline elevation and subcooled liquid spills into the steamlines.

This scenario produces the maximum subcooling in the liquid because 2

it results in the shortest elapsed time to fill the vessel, and hence a smaller amount of decay heat is transferred to the liquid.

The system response from SAFE provided the inlet conditions at the steamlines, and hence SRVs, at the time of liquid discharge. These conditions are shown in Table 5-1. From this table, the SRV inlet conditions range from saturated liquid to a maximum of 100*F of subcooling, at a pressure of approx; .tely 115 psia. This range of conditions is outside the range of inlet subcooling tested as a part of the BWR Owners' Group (BWROG) program (Reference 3), even though the pressure is well within the range tes ted .~ (Tested conditions include liquid inlet subcooling from 15 to 50*F with pressures up to 250 psig.) In order to evaluate the SRV operability, the model of F.J. Moody, Piping Forces Caused By Suddanly Opened SRVs (Reference

10) was used. Application of this model to the BWROG tests indicated that the predicted SRV loads were above those measured. Therefore, this model offers a conservative and convenient method of evaluating

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This model conservatively assumes that the SRVs open instanteously with various upstream conditions. The reaction force across the valve and discharge piping result from two different but i related. types of loads. First, the sudden introduction of a large pressure difference across the valve generates a compression or shock wave which advances through the stagnent fluid (either air or steam) in the pipe. Soon thereafter, the fluid expands (flashes) filling the cross-section of the pipe and begins to travel down the pipe.

The interface between the stagnent fluid (air or steam) and the moving fluid (liquid) imparts an additional force which is referred to as the interface force. Generally, the shock wave and consequental shock force move down the pipe at a much faster speed j than the fluid interface and its associated force. However,' for purposes of this analysis, it was conservatively assumed that both forces occurred concurrently and thus can be directly added together to produce a conservative upper bound of the reaction force.

To conservatively bound the loads a range of potential fluid conditions was assumed, including the maximum subcooled upstream conditions. For comparison, the reaction force due to the sudden opening of the SRV with the reactor at high pressure.and'with steam upstream conditions (conditions under normal relief mode of 1

operation) was .also determined. This calculation was performed utilizing the SRV inlet conditions during the initial portion of the I

event when th'e . SRVs are actuating in their normal pressure relief mode. If the forces due to the saturated or. subcooled liquid discharge under the conditions postulated are less than those caused

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by the normal steam discharge forces, then the operability of the SRVs will not be impacted. The results of these' calculations are discussed in Section 5.3.

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5. RESULTS

~_ 5.1 . FUEL INTEGRITY m'~ The system pressure and reactor water level responses for the postulated fire event with an assummed operator action time of 30

, minutes are presented in Figures 1 and 2 respectively. Figure 1 shows that-the reactor pressure-cycles between the setpoints of the SRV, approximately 1100 psig, until manual depressurization occurs at 30 minutes. Figure 2 shows that the extreme top portion of the core uncovers shortly before vessel depressurization and recovers acmentarily '=lien Line vessel is depressurized. This briet period of uncovery produces negligible cladding heatup because the power generation from the. top portion of the core or fuel bundle is very low. The portion of the fuel bundle which experiences the highest heatup for the event remains covered during this period. As the vessel depressurization continues . the core uncovers again. This

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time a high power generation region of the core is uncovered and the fuel cladding begins to heatup. Figure 3 shows the fuel cladding heatup following uncovery based on- the bounding assumption of adiabatic heatup. This heatup is terminated when this region of the core is recovered. The resulting maximum cladding. temperature is only 1013*F. .It can-be seen from Figure 3 that the duration of the fuel cladding heatup is less than 3 minutes. Since the' PCT is less than the lowest temperature at which fuel cladding perforations are expected (approximately 1500*F) and the duration of the fuel' cladding

. heatup period is short, no fuel cladding damage would occur.

5.2 SUPPRESSION POOL' INTEGRITY =

The temperature response of the suppression pool to the postulated event ' is shown' in Figure 4. The RHR heat exchanger is. assumed to initiate pool cooling at 43 minutes, corresponding to the time when- -

the SRVs begin discharging liquid.. The suppression pool temperature 4

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e at this time is below 150*F which is sufficiently low to assure that the RHR pumps can operate without cavitation. The peak suppression pool temperature of 193*F is reached in another 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. The corresponding pressure in the suppression pool at this time is estimated to be approximately 27 psia. Both the suppression pool temperature and pressure are below the design conditions of 220*F and 56 psig. The peak temperature of 193*F and corresponding pressure of 27 psia results in an available NPSH of 47.5 ft. Since the required NPSH is only 13 ft, plenty of margin remains to assure that the pumps can continuously operate and bring the reactor to cold shudown.

5.3 SRV OPERABILITY The SRV fluid conditions are ~ shown in Figure 5 and summarized in Table 5-1 (Figures 1 and 2 depict the system response). The saturated and subcooled liquid reaction forces are bounded by the 6100 lb reaction force determined from the subcooled inlet conditions. For comparison purposes, the reaction forces due to the sudden opening of the SRV in a high pressure steam environment during the initial portion of the event was computed as approximately 22,000 lbf. Thus, the discharge force with liquid upstream conditions is approximately one-third of the normal opening loads for SRV actuation in a steam environment. Therefore, SRV operability will not be affected as a result of the postulated SRV liquid discharge.

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TABLE 5-1 FLUID CONDITIONS FOR SRV LOAD CONSIDERATIONS LIQUID STEAM DISCHARGE DISCHARGE Vessel Pressure (psia) 115 1100 Fluid Enthalpy (Btu /lbm) 207 to 309 1200 Amount of Subcooling (*F) O to 100 N/A Maximum Force (1bf ) 6100 22000 e

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6. REFERENCES

1) ,

" Analysis In Support of FitzPatrick Compliance to Appendix R",

. General Electric Company, NSEO-105-1282, December 1982.

2) Memorandum for Roger J. Mattson (NRC), From L.S. Rubenstein (NRC), " Use of the Automatic Depressurization System (ADS) and Low Pressure Coolant Injection (LPCI) to meet Appendix R, Alternate Shutdown Goals", Decmeber 1982.

3) " Analysis of Generic BWR Safety / Relief Valve Operability Test Results," General Electric Company, NEDO-24988, October 1981.

4) " Decay Heat Power in Light Water Reactor," ANSI /ANS 5.1-1979, Approved by American National Standards Institute, August 29, 1979.

5) "Containnent Data Sheet for James A. FitzPatrick", CE Document Number 22A5747.

6) Stone & Webster Engineering Calculation for FitzPatrick, "RHR Pump Suction Losses for Containment Spray Mode, Single and Parallel Operation (RPSLCSMSPPO)," Job Order #12966.08.00024, Calculation #10-26, Rev. O, dated 12/12/78.

7) Stone & Webster Engineering Calculation for FitzPatrick "RHR Pressure Drop Calculation for Modes C1 & Cl Train 'A'", Jeb Order #11825.10, Calculation #10-23, dated 7/22/77.

8) Stone & Webster Engineering Calculation for FitzPatrick, "RHR

, Pressure Drop Calculation for Codes Cl & C2 Train 'B'", Job Order #11825.10, Calculation #10-24, dated 5/12/77.

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9) "The Thermal Hydraulics of a Boiling Water Nuclear Reactor".

R. T. Lahey, Jr., F. J. Moody, American Nuclear Society, 1977.

10 -" Unsteady Piping Forces Caused by Hot Water Discharge from Suddenly Opened Safety / Relief Valves," F.J. Moody (GE), Nuclear Engineering and Design, Volume 72, pages 213-224, 1982.

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