ML20129B541
| ML20129B541 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/30/1989 |
| From: | Bass J, Dudiak J, Marmo C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20129B484 | List: |
| References | |
| WCAP-12207, NUDOCS 9610230041 | |
| Download: ML20129B541 (39) | |
Text
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WESTINGH0USE CLASS
- WCAP-12207 REDUCTION IN THE MININUN RHR$ FLOWRATE DURING NID-LOOP OPERATION FOR BYRON AND BRAIDWOOD POWER PLANT UNITS I AND 2 June 1989 Prepared by: J. C. Bass J. G. Dudiak C. A. Narmo R V. Chavez R. B. Schreiber A. J. Abels
' WESTINGH0USE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230 9610230041 961019 DR ADOCK0500g6
9 TABLE OF CONTENTS Section 1111g P_gst
1.0 INTRODUCTION
1 2.0 RHRS SYSTEM DESCRIPTION 1 l
3.0 VORTEX FORMATION AND AIR ENTRAINMENT 2 I I
4.0 RHR FLOW REQUIREMENT BASES 2 l
5.0 RHR PUMP THRUST BEARING 9
. l 6.0 TECHNICAL SPECIFICATION CHANGES 12 7.0 FSAR CHANGES 13
8.0 CONCLUSION
S 14 APPENDICES A TECHNICAL SPECIFICATIONS A-1 B FSAR B-1 1
LIST OF FIGURES Fioure litJ3 Egg 1 REQUIRED RCS WATER LEVEL 3 l
2 DECAY HEAT LIMIT CURVES 5 l 3 CORE DELTA T VERSUS DECAY HEAT LIMIT 7 1
LIST OF TABLES TABLE Titfe Egg 1 PUMP OPERATING MODES 11 4
l 11 l
i
l . , -
REDUCTION IN THE MINIMUM RHR FLOWRATE
- DURING MID-LOOP OPERATION j FOR
- BYRON AND BRAIDWOOD P0WER PLANT UNITS 1 AND 2 i
i i 1.0 Introduction
- The purpose of this report is to document the evaluations and analyses performed by Westinghouse to support a reduction in the minimum Residual Heat Removal (RHR) flowrate during mid-loop operation for Byron and
~Braidwood Units 1 and 2.
Mid-loop operation occurs when the plant is operating with the Reactor
- Ccolant System (RCS) partially drained in Modes 5 and 6. Currently, for i Byron and Braidwood Units 1 and 2, there are no Technical Specification RHR flowrate requirements for operation in Mode 5. However, for Mode 6 cperations, a minimus RHR flowrate of 2800 gallon per minute (gps) is specified.
The Connonwealth Edison Company has requested that Westinghouse evaluate a reduction 'In the minimum RHR flowrate belcw the current Technical Specification requirement of 2800 gps, for Byron and Braidwood Units 1 and
- 2. A lower RHR flow during mid-loop operation could reduce the potential for air binding of the RHR pumps.
The evaluations performed to address a reduction in the RHR specified flow requirement during mid-loop operation are consistent with the following !
concerns: (1) decay heat removal. (2) thermal considerations (3). boron I mixing and stratification. (4) control valve cavitation and (5) l inadvertent boron dilution. In addition, a revised RHR pump thrust bearing expected life has been calculated based on reducing the minimum RHR flowrate below 2800 gps.
Proposed revisions to the Byron /Braidwood Technical Specifications and Updated Final Safety Analysis Report (UFSAR) are provided to reflect a reduction in the minimum RHR flow requirements to below 2800 gpm.
2.0 RHRS System Description The primary function of the Residual Heat Removal System (RHRS) is to remove residual heat from the core and reduce the temperature of the RCS during the second phase of plant cooldown. As a secondary function, the .
RHR$ is used to transfer refueling water between the Refueling Water )
Storage Tank (RWST) and the refueling cavity before and after refueling - '
operations. The RHR$ also serves as part of the Emergency Core cooling System (ECCS) during the injection and recirculation phases of a Loss-of-Coolant Accident (LOCA).
1 l
Th3 RHR$ censists of two parallel RHR trains. Th3 inlet line to each train of the RHRS is connected to a reactor coolant loop hot leg, while the return lines are connected to the cold legs of each of the reactor coolant loops. Each train includes one centrifugal RHR pump, one residual heat exchanger (shell and U-tube type), associated piping, valves, and instrumentation.
During RHR$ operation, reactor coolant flows from the RCS to the RHR pumps, through the tube side of the residual heat exchanger, and back to the RCS cold legs. Heat is transferred from the reactor coolant to the Component Cooling Water (CCW) circulating through the shell side of the residual heat exchangers.
3.0 Vortex Formation and Air Entrainment The Reactor Coolant System (RCS) water level is lowered in Hodes 5 and 6 to facilitate removal and reinstallation of the reactor head during refueling outages. Operation with the RCS partially drained may also be necessary for the inspection and maintenance of RCS components such'as reactor coolant pumps and steam generators. However, when the reactor coolant level in the RCS loop piping is lowered, there is a potential for air to be drawn into the RHRS suction line (air entrainment) due to the development of a vortex. Air entrainment into the RHRS m:1d cause air binding of the RHR pumps and thus, result in the inadvertent loss of decay h:at removal capability. The tendency for vortex formation at the RHR suction line, and subsequent air entrainment into the RHRS, is a function of the water level above the RHR$ suction nozzle and the RHR flowrate. j The lower the level, or the higher the RHR flowrate, the greater the '
pstential for a vortex to develop and air to be drawn into the RHRS.
Therefore, the likelihood of vortex formation due to partial draining of the RCS can be offset by reducing the RHR flowrate. Figure 1 shows the RCS hot leg water level (inches above the centerline) as a function of RHR intake flow.
4.0 RHR Flow Requirement Bases The required minimum RHR flowrate during mid-loop operations is based on the following concerns:
- The ability of the RHRS to remove dacay heat such that RCS temperature can be controlled.
- Sufficient flow is provided to ensure that the reactor coolant temperature rise through the core does not exceed reactor vessel internals delta T limits.
Sufficient flow is provided to ensure that the reactor coolant is mixed such that significant boron stratification does not occur.
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- Sufficiont flow is provided to ensure that the pressure drop across the RHR bypass flow control valve does not result in cavitation of the reactor coolant.
- Sufficient flow is provided to ensure that inadvertent boron dilution events can be identified and terminated by operator action prior to the reactor returning critical.
Decay Heat Removal As stated previously, the primary function of the RHRS is to remove decay heat during the second phase of plant cooldown. However, at reduced RHR flowrates, the decay heat removal capacity of the RHRS will be decreased.
Therefore, lower flowrates require that the reactor be shutdown for a longer period of time before the RHRS can remove all of the decay heat being generated. Figure 2 shows the minimum flowrate required to maintain a constant reactor coolant temperature, as a function of time after shutdown. The three curves presented in Figure 2, correspond to the following cases.
A reactor coolant temperature of 140 F, 0 maintained by one RHR train operating at the indicated flowrate.
A fonctor coolant temperature of 140 F, 0 maintained by two RHR trains, each operating at the indicated flowrate. !
A reactor coolant temperature of 200 F, 0 maintained by one RHR i
train operating at the indicated flowrate.
The curves indicate that decay heat decreases as a function of time after initial reactor shutdown. Thus, as the time after plant shutdown increases, the decay heat removal requirements for RHR flow are ' reduced.
Thermal Considerations Another potential concern with low RHR flowrates is thermal effects in-the reactor vessel and reactor coolant loops. Thermal stratification (non-isothermal conditions) may be a concern for the following two reasons:
Core reactivity varies as a function of coolant density and, thus, temperature. 1 Coolant temperatures in the core are inferred from temperature measurements taken in the RCS/RHRS loops.
\ l l
Regardingthefirstconcern,corereactivitygndt.Miateshutdownmargin are evaluated for a range of temperatures (68 F to 2000F) during Mode 5 operation. Therefore, thermal stratification will not create temperature conditions in the core more adverse than already considered.
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Flow enters the reactor vessel from each cold leg nozzle. The flow is forced to the bottom of the vessel and up through the reactor core.
Thus, adequate mixing of the reactor vessel volume in order to minimize thermal stratification is expected at reduced RHR flows.
In addition to potential thermal stratification, a reduction in RHR flowrate will increase the reactor coolant temperature rise through the core. The decay heat load is removed by increasing the temperature of the coolant as it passes through the core. As the mass flowrate is decreased, the temperature rise must increasc to maintain constant heat removal.
Certain structural considerations of the reactor vessel internals, limit acceptable core temperature rise. In particular, the most limiting components in terms of core temperature rise are the baffle-former bolts and baffle-barrel bolts.
Based on the fatigue usage factor of the baffle-former and baffle-barrel bolts, the maximum allowable steady state difference between reactor vessel inlet and outlet temperatures during mid-loop operation is 720F.
The minimum allowable RHR flowrate as a function of time, and the decay heat limit curve for Mode 5 (temperature less than 2000F), are shown in Figure 3. The curves intersect at approximately 47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />, after which the core delta T becomes the limiting factor. In Mode 6 (temperature less than 1400 P), decay heat will always be limiting.
Boron Mixing and Stratification Sufficient RHR flow must be provided to maintain a uniform boron concentration throughout the RCS. " Boron stratification" refers to the localized variations in boron concentration. Boron stratification is most likely to occur in the RCS when a contro116d boration (or dilution) operation is first initiated. During this operation, the RHR flow ensures mixing within the RCS volume. Thus, as RHR flow is reduced, the mixing rate decreases, and the time required to obtain a uniform RCS boron concentration increases. Typically, however, the RCS boron concentration is stabilized at the required shutdown margin prior i.c reducing RHR flowrate, ensuring a uniform boron concentration, Provided that the reactor coolant is not intentionally diluted during mid-loop operations, precipitation and local evaporation would be the most likely mechanisms for inducing a boron gradient in the reactor vessel.
However, concentration would be in the range of 2000 ppe (1%
concentration). Since the saturation temperature of a 1% solution is less .
than 320 F, boric acid precipitation would not occur. Even if mass evaporation would occur, the local boron concentrations (without mixing) would actually increase, which is in the conservative direction.
Centrol Valve Cavitation RHR flow is reduced during mid-loop operation by fully closing the RHR bypass flow control valve (FCV-618 or 61g), and then slowly closing the associated hand control valve (HCV-606 or 607). The pressure drop across 6
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j V the control valve increases as flow is reduced. Eventually, cavitation of the reactor coolant could result. Cavitation that occurs in control valves under high pressure drop conditions, is due to a portion of the liquid transforming into the vapor phase during rapid acceleration of the a fluid inside the valve, and the subsequent collapse of these vapor bubbles
! downstream of the valve. Severe cavitatien could cause excessive wear and j vibration in the piping downstream of the valve.
l i
An analysis was performed t! sing standard equations published by valve manufacturers to predict the onset of cavitation. The results indicate that cavitation will not occur at flowrates greater than 2000 gpm.
1 However, Westinghouse does not recossend establishing an RHR flow limit of 2000 gpm at this time. This limit may be over1r conservative for the j following reasons.
- The formulae used to predict the onset of cavitation are riot l
! exact. Thus, cavitation may not occur until a flowrate lower than !
2000 gpm is reached. '
4 When cavitation occurs, it will initially be at a low level, which may not be detrimental to the valve or piping.
l Therefore, it may be possible to reduce the flowrate to significantly below 2000 gpm without harm. Any cavitation that is severe enough to cause damfge would be evident from excessive noise and vibration in the piping downstream of the valve. At this time. Westinghouse reconnends that the limit on RHR flowrate initially be established based on the other
- concerns identified (decay heat removal, thermal considerations, and boron i
nixing and stratification). The first time that RHR flow is reduced to a 4
value less than 2000 spa, the piping immediately downstream of the controi valve should be visually monitored. If excessive vibration or audible i noise is observed, it may be necessary to establish a higher minimum RHR flowrate, based on cavitation concerns.
Inadvertent Boron Dilution The proposed reduction in RHR flow during mid-loop operation will potentially impact those transients explicitly analyzed in Modes 5 and 6.
The only non-LOCA event analyzed in these modes is the Chemical and Volume Control System (CVCS) Malfunction, which results in dilution of the i
primary coolant (presented in Section 15.4.6 of the Byron /Braidwood UFSAR).
A reduction in RHR flow during mid-loop operation has the potential to impact the CVCS Malfunction event in the following two areas.
A reduction in explicit RHR flowrate assumptions.
The vessel mixing assumption during a boron dilution.
A CVCS malfunction event in Mode 6 is prevented by administrative controls
, which isolate the RCS from any potential source of unborated water. The appropriate CVCS valves that are required to be closed and secured are 8
identified in the Byron and Braidwood Technical Specifications. Thus, the proposed reduction in RHR flow during mid-loop operations has no impact en the Mode 6 analysis.
Mode 4 is the only mode analyzed in the CVCS Malfunction event that explicitly accounts for a minimum amount of RHA flow. The Mode 5 and 6 analyses do not assume an explicit RHR flow value. Since mid-loop 1 operation is not permitted in Mode 4 and the RHR flow requirement outside of mid-loop operation will not,be changed, there is no impact on the CVCS Malfunction event in Mode 4.
The Modes 4 through 6 analyses that account for RHR performance (either explicitly or implicitly), assume that the RHR flowrates provide adequate vessel circulation to prevent boron stratification and support the boron dilution transient mixing assumptions. The proposed reduction in the minimum RHR flow requirement during mid-loop operation impacts only the Mode 5 analysis. Therefore, it was necessary to confirm that the current Mode 5 boron dilution analysis remains valid for the reduced RHR flowrate. The proposed reduction in RHR flow to 1000 gpa during mid-loop operation will not invalidate these assumptions, thus, the current Mode 5 analysis is still valid.
It has been demonstrated above that a reduction in the minimum RHR flowrate requirement during mid-loop operation to 1000 gpa or greater will have no adverse effects on the non-LOCA accident analyses. Therefore, the results and conclusions presented in the Byron /Braidwood UFSAR remain valid.
5.0 RHR Pump Thrust Bearing Westinghouse evaluated the Byron /8raidwood RHR pump motor thrust bearing
, life for the hydraulic thrust, deadweight and seismic loads which act on the thrust bearing.
The hydraulic thrust loads were based on results recorded during Ingersoll-Rand testing of the same pump model. This testing recorded only hydraulic thrust loads developed by the pump internals and is independent of the pump support conditions.
The seismic evaluation is a calculation of the bearing load due to the rotor degdweight loads, vendor test hydraulic loads and the seismic load !
of 2.1 g . The total bearing load is based only on those loads reacting on the pump / motor rotor assembly and is independent of pump support conditions. The seismic evaluation of the thrust bearing demonstrated that the bearing could withstand the magnitude of the combined seismic, l hydraulic and deadweight loads and operate through the duration of a '
seismic event. Furthermore it demonstrkted that the thrust bearing capacity is adequate to withstand seismic event loadings incurred and that the duration of five OBE and one SSE events is so short that the seismic conditions have an insignificant effect on the overall bearing life.
1 identified in Byron /Braidwood RHR pump equipment specification p
9
- 4 l
- l 5
Bearing life wag predicted based on the normal hydraulic and deadweight loads. The 810 bearing life was calculated for worst case operation and for realistic normal pump operation with both single pump service duty
- and split service duty between the two pumps per plant. The B10 bearing
! life is the minimum expected bearing life as defined by AF8MA Standard 9.
l The 810 thrust bearing life was first calculated assuming that the RHR pump operates continuously under the worst thrust load conditions. The
{; thrust bearing life while operating at the flawrate corresponding to the peak hydraulic thrust (approximately 2000 GPM; is 7g37 hours. This value is very conservative since the RHR pumps do not operate at this flowrste during any of the defined plant modes of operation.
l~
The 810 bearing 11fe was then calculated as a cumulative value based on pump operation at the various operating modes identiff sd in Table 1. This bearing life was converted into a replacement interval which must be 1
followed in order to ensure that the bearings are capable of operating for the entire post-accident requirement. The replacement intervals were
- calculated for both one year and b ' ~ post-accident operation. The intervals were calculated both ass e that one RHR pump performs all service duty and also assuming that the pumps each see 50 percent service duty f modes.gr . The theresulting refueling,thrust shutdown, bearing midloop and plantintervals replacement cooldownare cperating tabulated below.
2 The formula for bearing life 810 is available in engineering handbooks and bearing manufacturers' catalogs, but may appear in
. different forms and use different factors depending upon the source. This is because it is a statistical life based on empirical
- data. Results, however, using the various forms are not expected to be significantly different. This evaluation is based on the TRW Engineer's Handbook (2nd Edition,1982) formula
- )
Life = 1500 (service factor)3 Service Factor = (rated capacity) / (equivalent load)
Rated capacity -
defined by Handbook based on bearing model Equivalent Load = 0.62 (radial load) + (thrust load) 3 The selection of single pump service and 50 percent split duty service is intended to predict a mininAm and a realistic bearing life expectation. The results are presented in terms of a recommended bearing replacement interval as a more practical means
/ for developing a maintenance program than is the predicted bearing life expressed in total hours of operation.
A modification to the pump impeller wear ring can be performed to reduce the effective downthrust on the motor thrust bearing. An evaluation of the specific operating modes for another plant application incorporating the modification showed a significant increase (-4x) in thrust bearing interval replacement. A plant specific evaluation is required to determine the anticipated increase in bearing replacement interval (or if the recommended replacement interval would exceed the life of the plant) with the modification incorporated.
10
l i
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! Post-Accident Period Sinale Pumo Service Duty solit Service Duty I 100 days 6.g years 13.7 years 1 year 5.9 years 11.8 years l
- There is a precaution regarding the use of the thrust bearing replacement j intervals as identified above. The vendor testing of hydraulic thrust
- loads identified that the thrust load drops significantly between flow i rates of 2800 and 3000 GPM. Since the Byron /Braidwood RHR pumps operate
{ for nearly 70 percent of their life at 3000 GPM, the calculated bearing
! replacement intervals are predominantly controlled by the thrust load at
! 3000 GPM. Due to the shape of the hydraulic thrust curve, any slight
! change in flowrate or a minor variation in pump hydraulic characteristics i while operating at a nominal 3000 GPM can result in a significant reduction in the thrust bearing life. For this reason, it is advisable to j increase the flowrate to at least 3300 GPM for all operating modes which j currently have 3000 GPM.
! 6.0 Technical Specification Changes i
! During pid-loop operations, the RHRS operates at reduced flowrates to
! avoid air binding of,the RHR pumps. The acceptability of the reduced RHR i flowrate could be based on accompanying administrative *::tions such as a
- prescribed minimum time after shutdown, elimination of .,ilution sources, j avoidance of boration/ dilution operations, or core temperature rise i monitoring. Since the requirements are dependent on plant conditions, it
- is reconsnended that a single flowrate requirement not be included in the
! Technical Specifications. Instead, acceptable io1R flowrates (minimum and i maximum) that are consistent with the plant conditions would be specified j in the plant procedures.
Since the safety concerns for Modes 5 and 6 are similar, it is also reconunended that the Mode 5 and 6 specifications be consistent in
- addressing the above concerns.
The Technical Specifications place limitations on the RHR$ during mid-loop operation by specifying a minimum flow requirement for the purpose of decay heat removal and the number of RHR trains which must be operable.
The Technical Specifications do not, however, contain restrictions ba. sed on minimizing air entrainment in the RHRS as a result of vortexing which may occur during mid-loop operation under certain conditions.
1 11
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I TABLE 1 - PUNP OPERATING N0 DES NODE FREQUENCY DURATION FLOW SUCTION (gpm) PRESSURE
. (psia)
Surveillance Testing 4 per year 30 minutes 500 59 RHR Initiation 2 per year 5 minutes 500 400 2 per year 30 minutes 575 400 Plant Cooldown 2 per year 30.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 3000 391
- Refueling & Shutdown 2 per year 1 month 3000 38 Mid-Loop Operation 2 per year 2 weeks 1000- 26
, to 1300 Spurious SI 2 per year 30 minutes 500 48 -
Large LOCA l'per 40 yr. 30 minutes 3800 41 Pcst-Accident 1 per 40 yr. 1 year 3950 26 Recirculation
/
1 3 i l
t 1
l l
1 12
Based on the above discussion, it is recossended that Surveillance Requirements 4.9.8.1 and 4.9.8.2 be revised to delete reference to a specified flowrate of greater than or equal to 2800 gps. In addition,
- revision to the Bases of Technical Specifications 3/4.4.1 and 3/4.9.8 are proposed to identify: (1) the concerns that are to be addressed in determining the minimum RNR flow requirements during mid-loop operation, (2) the dependency of the required minimum RHR flowrate on plant conditions, and (3) the potential for vortexing to cause air binding of the RHR pumps and subsequent loss of decay heat removal, due to partial draining of the RCS.
I The Byron /Braidwood Technical Specifications have been marked-up to reflect these proposed changes. The affected pages are provided in Appendix A of this report.
7.0 FSAR Changes The impact of a reduction in the minimum RHR flow requirements on the Byron /Braidwood Updated Final Safety Analysis Report (UFSAR) has been reviewed. It is reconmended that Section 6.4.7.2, Residual Heat Removal System Design, be revised. The following additional information is provided with regard to mid-loop operation: (1) the concerns addressed in determining the reqilired minimum RHR flowrate. (2) the specification of the RHR flow requirements consistent with plant conditions, and (3) the potential for vortex formation and air binding of the RHR pumps.
The Byron /Braidwood UFSAR has been marked-up to reflect the above l changes. The mark-up is providad in Appendix A of this report. ;
I 13
l 8.0 CCnclusions
\
Based on the evaluations and analyses performed by Westinghouse to support '
a reduction in the minimum RHR flowrate during mid-loop operation for Byron and Braidwood Units 1 and 2, the following is concluded:
The minimum RHR flowrate dictated by decay heat and core delta T concerns is a function of time after shutdown, as is shown in Figure 2 and Figure 3.
Cavitation of the reactor coolant may occur when RHR flow is reduced below approximately 2000 gps, due to the increase in pressure drop across the flow control valve.
A reduction in the minimum RHR flowrate requirement to 1000 gpm or greater during mid-loop operation, will not adversely affect the non-LOCA accident analyses, In addition, it is recommended that a single flowrate requirement not be included in the Technical Specifications. Instead, acceptable RHR flowrates (minimum and maximum) that are consistent with the plant conditions would be specified in the plant procedures for Byron and-Braidwood Units 1 and 2.
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s 1 - - * - - = _ sn i _
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APPENDIX A TECHNICAL SPECIFICATION CHANGES t
A-1
e-
! REFUELING OPERATION $
3/4.9.8 RESIDUAL MEAT REMOVAL AND COOLANT CIRCULATION NIGN WATER LEVEL J
LIMITING CONDITION FOR OPERATION i
3.9.8.1 At least one residual heat removal (RNR) loop shall be OPERABLE anc
- in operation."
APPLICA8FLITY: ICDE.6, when the water level above the top of the reactor vessel flange is greater than er equal to 13 feet.
ACTION:
With.no RHR loop OPERABLE and in operation, suspend all operations tavety m; an increase in the reactor decay heat lead or a reductisa in 6cron concent?s-tien of the Reacter Coolant Systas and immediately initiets etreettive action to return the required RNA loop to OPERA 8LE and operating status as soon as pessible. Close all containment penetrations providing direct access from the containment atmosphere to the evtside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in speration and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. -
"The Iput leep any he removed fees operation for up to I hour per 8-hour period during the perfomance of CORE ALTERATIONS in the vicinity of the reacter vessel het legs.
SYRON = UNIT 5 1 & 2 3/4 9-9
~.
. REFUELING OPERATIONS LOW WATER LEVEL -
LINTTING CONDITION FOR OPERATION 3.9.8.2 Two residual heat removal 'R) Isops shall be OPERABLE; a;W at least one W R 1eap shall.be in operation.'
APPLICABFLTTY: SCOE 5, when the matar level e%ve the top of the reacter vessel f ange is less than 23 feet.
MT10N:
- a. With less than the required RNA loops OPERA 8LE, immediately initiate cerfective action to return the required RHR loops to OPERA 8LE status, or establish greater than er equal to 23 feet of water above the reactor vessel flange, as soon as possible.
- b. With no RNR loop in operation, suspend all operations involving a reduction in beren concentration of the Reacter Coolant Systes and immediately initiate corrective action to return the required RHR loop to speration. Close all containment penetrations providing J direct access free the containment atmosphere to the outside i ateesphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIRENENTS 4.9.8.2 At least one RNR loop shall be ' verified in operation and circulating feector coolant ' ~ - -" ' - -- ^
at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
" Prior to initial criticality, the RNR leep may be removed free operation for o to 1 heve per 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period during the performance of CORE ALTERATIONS in the vicinity of the reacter vessel het legs.
l 1
3/4.4 REACTOR C00LMT SYsTu .
BASES ,
3 /4. 4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops i:: operation and maintain DNSR above the applicable design bases DNSR during all normal operations and anticipated transients. In 20E51 and I with one reacter cool-ant loop met in operation this specification requires that the plant be in at least WT STA25Y within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
In ICOE 3, two reacter coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal .
accident however, a single reacter coolant leep provides sufficient heat removal If a bank withdrawal accident can be prevented,14.. by opening the Reactor Trip system breakers. Single failure considerations require that two loops be OPERA 8LE at all times.
In MODE 4, and in MDDE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least.two loops (either RHR er RCS) be OPERABLE.
In MODE 5 with reactor coolant Isops not filled, a single RMR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steas generators as a heat removing component, require that at least two RHR loops be OPERABLE. .
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flew to ensure sizing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with Soron reduction will, therefore, be within the capability of operstar recognition and control. l
" The restrictions on starting a reactor coolant pump with one er more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by ener1r additions free the Secondary Coolant System, which could exceed the lin' ts of Appendix 8 to 10 CFR Part 50. The RCS will be protected against everpressure transients and will not ascoed the limits of l Appendia 8 by restricting starting of the RCPs to when the secondary water '
tamperature of each steam generator is less than 50*F above each of the RCS cold leg tamperatures. .
The requirement to maintain the boren concentration of an (selsted loop greater than er equal to the beron concentration of the operating loops ensures that no reactivity additten to the core could occur during startup of an isolated leep. Verification of the baron concentration in an idle loop prior te openin0 the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated leep. .
Startup of an idle loop will inject cool water free the leep into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated leap startup until its temperature is within 20'F of the operating loops.
BYRON - UNITS 1 & 2 8 3/4 4 1
REFUELING OPERATIONS BASES e
3/4.9.5 RFUfLING MACHINE The OPERABILITY requirements for the refueling ent6ne and auxiliary hoist enson 1Ast: (1) refueling machines will be used for a:aament of drive rods and fuel assemblies, (2) each refueling machine has sufficient lead capacity to lift a drive red er fuel assembly, and (3) the core internals and reactor vessel are protected from escessive liftin engaged during lifting operations. g force in the event they are inadvertently 3/4.9.7 CRANE TRAVEL - SPENT FUfL STORACE FACILITY ,
The restriction en movement of leads in excess of the nominal weight of a fuel and control red assembly and associated handling tool over other fuel assemblies in the storage pool areas ~ ensures that in the event this lead is .
dropped: (1) the activity release will be limited ta that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the detivity release assumed in the safety analyses.
- 3/4.9.8 RESIDUAL NEAT REMOVAL AND COOLANT CIRCULAT!0N The requirement that at least one residual heat removal (RNR) loop be in operation ensures that
- (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is saintained through the core to sinimite the effect of a beren dilution incident and prevent boron stratification.
w m sg.t r The requirement to have two RNR loops OPERA 5LE when there is less than 23 feet of water abate the reactor vessel flange ensures that a single failure )
of the operating RNr. leep will not result in a templete less of RHR capability. I With the reactor vezsel head removed and at least 33 feet of water above the reacter vessel flange. a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RNR loop, adequate time is provided ta initiate emergency procedures to see1 the core.
3/4.9.9 CONTAletENT PURGE ISOLATION iVairii l The OPERABILITY of this system ensures that the containment purge penetrations will be automatically feelated detection of high radiation levels within the containment. The OPERABI of this system is required to restrict the release of radioactive asterial from the containment staosphere to the environment.
1 SYRON - UNIT 5 1 & 2 8 3/4 9-2
. ur Byron /Braidwood Technical toecification Bases Insert The surveillance req &suent verifies that the RHR loop is operating and circulating reactor coolant to ensure the capability of the RHR system to maintain compliance with plant design limits. The required RHR loop reactor coolant flowrate is determined by the flowrate necessary to:
(1) provide sufficient decay heat removal capability, (2) saintain the reactor coolant temperature rise through the core within design limits, for compliance with fleurates assumed in the boron dilution analysis.
(3) prevent thermal and boron stratification in the core, (4) preclude cavitation of the reactor coolant downstream of the RHR flow control valve, and (5) ensure that inadvertent boron dilution events can be identified and terminated by operator action prior to the reactor returning critical. .
In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR loop flowrate determination must also consider the RHR pump suction requirements. At thit water level, the RHR pump can experiencc wortexing or cavitation conditions which would cause the loss of RHR pump operation, if the flowrate demand is too high. Operation with reactor coolant water at this level is often called mid-loop operation. Care must be taken in determining the RHR loop flowrate, when operating with water level in this region, to prevent loss of the RHR pump and subsequent loss of the RHR loop for decay heat removal.
f 4
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1 I
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i 1
4
e APPENDIX 8 FSAR CHANGES 1
3-1 l l
l
- a l
. S/3-UFSAR m
)
l i
If one RNRS pump is out of service and the alternate train becomes unavailable due to a passive failure during cooldown, the auxiliary feedwater system, along with the steam generator safety valves and steam generator power-operated relief valves, provides a completely separate, independent, and diverse means i of performing the safety function of removing residual heat, l which is normally performed by the RHRS when the RCS temper- l sture is less than 350*F. The auxiliary feedwater system is l capable of performing this function for an extended period of l time following plant shutdown until the RNRS is made available. l l
When the steam generators are down for maintenance, the RCS is depressurised and the RNRS operates at steady-state pressure and temperature conditions significantly below the RHRS design values. Passive failures of magnitude that could affect RNRS operation are not expected to develop at these conditions.
However, if one RHRS pump is out of service and the steam generators are down for maintenance, the development of a passive failure in the remaining RNRS train would not make that train unavailable for residual heat removal since in-service inspections are conducted periodically and ASME " code-allow-able" defects are not expected to grow appreciably during the life of the plant. A passive failure of the RHRS piping is not expected to produce a rapidly propagating crack that could result in a major rupture of a system pipe. Therefore, a detectable leakage crack is not expected to produce the effect of rendering an RNRS train inoperable. The operator would continue to use the RHRS train in conjunction with the chemical and volume control system. The centrifugal charging pump (s) will provide the makeup supply to compensate for the system inventory leakage.
Loss of RHRS cooling during maintenance activities and espe-cially due to air entrainment in the system has been con-sidered. If it ie -e;uired th:t th: ;;;e:0516::1 1 ::1 5:h;;t 1;;; ;;r sed7te1 fiv-INSECT.,_J0tain th; ;teea. ge..esete; tuh::, th:
r_e=te 1: th::ttled t: ri :t 1500 ;;r th:: ;5 e=ah a* the e ci 03:1 h::t I; seel 1;;;:. Draining is to the point where the indicated level is r, table and at the elevation of the center of the reactor vessel nossles. At this point, the reactor coolant level is monitoreS continuously to assure that the RHRS inlet lines do not becomo uncovered. Inventory makeup, if required, can be accomplished via the chemical and volume control system (CVCS)/ centrifugal charging pumps (s).
Should a RNRS inlet line become uncovered, air may be drawn into the suction piping and entrained in the fluid. Factors which minimise the effects of air entrainment on pump perfor-mance are as follows:
1
- 1. the location of the pumps provides positive head on the pump inlet, and 5.4-45
I Byron /Braidwood FSAR Insert During mid-loop operations, the RHRS operates at reduced flowrates to prevent air binding of the RHR pumps. Since the minimus RHR flow requirement is dependent on plant conditions, a sigle flowrate requirenent is not specified in the Technical Specifications. Instead, acceptable RHR ficarates (minimum and maximum) that are consistent with the plant condi' sons are specified in the plant procedures. The acceptability of the reduced minimum RHR flow requirement is based on the following concerns. ,
- The ability of the RHRS to remove decay beat such that RCS temperature can be controlled.
- Sufficient flow is provided to ensure that reactor coolant temperature rise through the core does not exceed reactor vessel internals delta T limits.
- Sufficient . flow is provided to ensure that the reactor coolant is mixed such that significant boron stratification does not occur.
l
- Sufficient flow is provided to ensure that the pressure drop '
across the RHR bypass flow control valve does not result in cavitation of the reactor coolant.
- Sufficient flow is provided to ensure that inadvertent boron dilution events can be identified and terminated by operator I actionpriortothereactorgraturningcritical.
5
Attachment C B. S. Ilumphries (Westinghouse) letter (CCE-96-207) to J. R. Meister (Comed), " Commonwealth Edison Company, Braidwood Units 1 &
2, Safety Evaluation SECL-96-193, RIIR Operation During Reduced Inventory Conditions," dated October 18,1996 I
l
--.- g g;g gg gg - - - -
l sa 3ss Westinghouse Energy Systems Ptnskrgh Pennsyivaras 15230 035:
Electric Corporation CCE-96-207 October 18,1996 Mr. J. R. Meister Commonwealth Edison Company l Braidwood Nuclear Station j Rural Route #1, Box 84 l Braceville, IL 60407 l
Commonwealth Edison Company l
l Braidwood Units 1 & 2 Safety Evaluation SECL-96-193 RHR Operation During Reduced Inventory Conditions
Dear Mr. Meister:
Enclosed for formal transmittal to Commonwealth Edison, Braidwood Station is Safety Evaluation SECL-96-193, "RHR Operation During Reduced Inventory Conditions," for Braidwood Units I and 2. -
Please contact us if you have any questions.
Very truly yours, WEST!NGHOUSE ELECTRIC CORPORATION C ,
B. S. Humphries, Manager l
' Commonwealth Edison Project Operating Plant Programs l
, Enclosure
~
ec: R. Krbec Braidwood ~
J.Sanchez Braidwood T. Tulon Braidwood a v.
~
DCT 18 '96 19:35 FR LJEC 418 374 6337 TO 818194583803 P. 0241 SECL No. % 193 Customer Reference No(s).
Westinghouse Reference No(s).
WESTINGHOUSE SAFETY EVALUATION CHECK LIST 1.) NUCLEAR PLANT (S): Braidwood Units la 2 2.) SUBJECT (TITLE): RRR Operation During Reduced Inventory Conditions 3.) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59(b) has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.
Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.
CHECK LIST PART A - 10CFR50.59(a)(1) 3.1) Yes ___ No _X. A change to the plant as described in the FSAR?
l 3.2) Yes _ No .X. A change to procedures as described it the FSAR?
3.3) Yes _ No .X. A test or experiment not described in the FSAR?
3.4) Yes __ No .X. A change to the plant technical specifications?
(See Note on Page 2.)
4.) CHECK LIST PART B - 10CFR50.59(a)(2) (Justification for Part B answers must be include on page 2.)
4.1) Yes ,_ No .X., Will the probability of an accident previously evaluated in the FSAR be increased?
4.2) Yes .,._ No .X. Will the consequences of an accident previously evaluated in the FSAR be increased?
4,3) Yes _ No .X., May the possibility of an accident which is different than any already evaluated in the FSAR be created?
4.4)
Yes __ No 1 Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
4.5) Yes .,_ No .X,, Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
4.6) Yes _ No X. May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?
4.7) Yes _ No _X. Will the margin of safety as described in the bases to any technical specification be reduced?
Page 1 of 9
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OCT 18 '% 19:36 FR bEC 41237U6337TOB18154583603 P.03/a1 SECL-96193 Page 2 of 10 NOTES:
If the answer to any of the above questions is unknown, indicate under 5.) REMARKS and
, explain below.
. If the answer to any of the above questions in Part A (3.4) or Part B cannot be answered in the 1 negative, based on written safety evaluation, the change review would require an application for license amendment as required by 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90.
I 4
5.) REMARKS:
The answers given in Section 3, Part A, and Section 4, Part B, of the Safety Evaluation Checklist, are based on the attached Safety Evaluation.
i 1
I FOR FSAR UPDATE '
Section: N/A Pages: N/A Tables: N/A Figures: N/A l
i SAFETY EVALUATION APPROVAL LADDER:
Nuclear Safety Preparer: Date: 10!!!!N l B. F. Maurer I Nuclear Safety Verifier: - - Date: O 0 ti. W. Whiteman Licensing Engineer Review: W Date: /# / ib iC J'. (iorrison
,' OCT 18 '96 19:36 FP WEC 412 374 6337 TO 818154583803 P.04/11 SECL-96-193 Page 3 of 10 SAFETY EVALUATION RHR OPERATION DURING REDUCED INVENTORY CONDITIONS BRAIDWOOD UNITS 1 & 2 1 l
l
1.0 BACKGROUND
l In response to Generic Letter 8712 (Reference 1), Commonwealth Edison (Comed) made the commitment that RHR flow would be reduced to approximately 1000 gpm through each RHR train prior to RCS draindown and while the RCS is in a drained down condition (Reference 2). At reduced inventory conditions, normally one RHR train is in operation, while the other RHR train is operable ;
but not operating, if a condition should occur that would affect the operating RHR train, the other train would be available for continued decay heat removal. Comed maintains the necessary .
emergency procedures for recovery from such an event.
The Reactor Coolant System (RCS) water level is lowered in Modes 5 and 6 during outages to -
facilitate miscellaneous maintenance activities. Operation with the RCS partially drained may be necessary for the inspection and maintenance of RCS components such as reactor coolant pumps and steam generators. However, when the reactor coolant level in the RCS toop piping is lowered, there is a potential for air to be drawn into the RHRS suction line (air entrainment) due to RCS loop level fluctuations and/or the development of a vonex. Air entrainment into the RHRS could cause air binding of the RHR pumps and thus, result in the inadvertent loss of decay heat removal capability.
The tendency for vortex formation at the RHR suction line, and subsequent air entrainment into the RHRS, is a function of the water level above the RHRS suction nozzle and the RHR flowrate. He lower the level, and/cr the higher the RHR flowrate, the greater the potential for a vortex to develop and air to be drawn into the RHRS.
In order to facilitate RCS cooldown while maintaining adequate margin to prevent air entrainment in the RHR system, resulting in cavitation of the operating RHR pump, the effect of implementing an increased RHR flow rate of 3300 gpm at reduced inventory conditions has been evaluated. This evaluation demonstrates that extending the approved range of operation of an RHR train from 1000 gpm to 3300 gpm at or above prescribed mmimum RCS water levels, as shown in Figure I to this safety evaluation, does not represent an unreviewed safety question per the criteria of 10CFR50.59 and will not require a change to the Technical Specifications.
2.0 LICENSING BASIS Title 10 of the Code of Federal Regulations. Part 50, Section 59 (10 CFR 50.59) allows the holder of a license, authorizing operation of a nuclear power facility, the capacity to investigate and disposition a change to the normal plant configuration. The increase in RHR flow rate at reduced inventory
T-I OCT 28 '% 19b7 FR WEC 412 374 6337 TO 818154583803 P.05/11 t
I SECL-96193 Page 4 of 10 conditions represents a change to the on' rmal plant configuration. Prior Nuclear Regulatory Commission (NRC) approval is not required to implement a change provided that the proposed change does not involve an unreviewed safety question or result in a change to the plant technical specifications. However, it is the obligation of the licensee to maintain a record of the changes or modifications to the facility, to the extent that such changes impact the Updated Final Safety Analysis Report (UFSAR). The code further stipulates that these records shall include a written safety l
evaluation that provides the basis for the determination that the change does not involve an unreviewed safety question. It is the purpose of this document to support the requirement for a written safety evaluation.
The RHR system is described in UFSAR Section 5.4.7, " Residual Heat Removal System," and refueling operations and operations during Mode 6 are discussed in UFSAR section 9.1 (Reference 3).
RHR operational criteria are further discussed in Reference 2. Technical Specification 3/4.9.8.2,
" Refueling Operations, Low Water Level" (Reference 4) presents the minimum requirements for RHR cperation during Mode 6'with a water level less than 23 feet above the vessel flange.
l 3.0 EVALUATION ,.
The primary function of the RHR system is to remove residual heat from the core and reduce the temperature of the RCS during the second phase of plant cooldown. As a secondary function, the l RHRS is used to transfer refueling water between the refueling water storage tank and the , fueling cavity before and after the refueling operations. The RHR system also serves as part of we Emergency Core Cooling System (ECCS) di ring the injection and recirculation phases of a Loss of-Coolant Accident (LOCA).
The RHRS consists of two parallel RHR trains. The inlet line to each train of the RHR system is )
connected to a reactor coolant loop hot leg, while the return lines are connected to the cold legs of i each of the reactor coolant loops. Each train includes one centrifugal RHR pump, one residual heat exchanger (shell and U-tube type), and associated piping, valves, and instrumentation.
l During RHR system operation, reactor coolant flows from the RCS to the RHR pumps, through the l tube side of the RHR heat exchangers, and back to the RCS cold legs. Heat is transferred from the l reactor coolant to the component cooling water circulating through the shell side of the residual heat l exchangers.
l l The required minimum RHR flow rate during reduced inventory operations is based on the following functional considerations: 1) provide sufficient decay heat removal capability,2) maintain the reactor coolant temperature rise through the core within design limits for compliance with flow rates assumed in the boron dilution analysis, 3) prevent thermal and boron stratification in the core, 4) preclude cavitation of the reactor coolant downstream of the RHR flow control valve, and 5) ensure that inadvertent boron dilution events can be identified and terminated by operator action prior to the reactor returning critical. These issues have been evaluated and presented in Safety Evaluation SECL-89-867 (Reference 5) and WCAP-12207 (Reference 6). However, since the desired change is I
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~T OCT 18 A9619:37FRN5b 4123746337Tb810154583803 P.06/11 l
SECL-96193 Page 5 of 10 to increase the RHR flow rate at reduced inventory conditions, the possibility of vortexing and air entrainment becomes the dominant concern, s
De sensitivity of air entralnment on RHR pumps to RCS level and pump flow rate is of increased importance at reduced inventory conditions. Based on operating experience and various test
, programs, guidelines have been developed regarding required water level for various RHR flow rates.
Correlations between RCS hot leg water level and RHR intake flow rate have been developed for l
RHR operation with a partially filled system. Correlations applicable to the Braidwood units are presented in Ref-rences 5 and 6. The relationship of maximum allowable RHR flow' rate versus minimum reduced inventory water level (relative to the centerline of the hot leg pipe) is provided for the range of RHR flows from 1000 gpm to 3000 gpm.
De supporting calculations that provided the basis for the flow versus water level correlations in References 5 ad 6, and WCAP 11916 (Reference 7), have been reviewed. The calculations have been determined to remain applicable for the reduced inventory conditions at Braidwood. As a result l of this review, it is concluded that the range of RHR flows of 1000 gpm to 3000 gpm can be reasonably extended to a flow rate of 3300 gpm in one RHR train. The minimum allowable water level in the RCS with an RHR flow rate of 3300 gpm has been calculated to be 2.25 inches above the, ,
center!!ne of the hot leg pipe. ne calculational methodology used in this calculation is supported by the test data for calculated Froude numbers between I and 3. The 3300 gpm flow rate in this ,
application results in a calculated Froude number of 2.3, which is well within the bounds or applicability of the methodology. Mainta!ning the water level at or above this level empirically limits the air entrainment to acceptable levels for RHR pump operation.
The concerns previously addressed in References 5 and 6 are not exacerbated by the increase in RHR flow to 3300 gpm. Decay heat removal capacity will be increased, with a subsequent decrease in the time required to remove the decay heat and maintain the RCS at the desired temperature. The increased flow rate will also result in improved thermal mixing in the reactor vessel and will minimize the temperature rise across the core. It will also result in increased boron mixing throughout the RCS. With increased RHR flow, the likelihood of cavitation of reactor coolant across the RHR control valves is further reduced.
De effects on non-LOCA transients in the previous evaluations (References 5 and 6) were related to minimum RHR flow rate assumptions and the ability of the RHR to provide adequate circulation to prevent boron stratification in the RCS. The increase in RHR flow rate to 3300 gpm will be beneficial to both of these concerns, and thus, there will be no adverse effects on the non LOCA transients.
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OCT 10 '% 19:38 FR LJEC 412 374 6337 TO 818154583803 P.07/11 SECL-%193 i
Page 6 of 10 l
4.0 DETERMINATION OF UNREVIEWED SAFETY QUESTION Increasing the RHR flow rate to a maximum of 3300 gpm during reduced inventory conditions at Braidwood Units I and 2 with the attendant limitation on water level has been evaluated using the i guidance of NSAC-125 and does not involve an unreviewed safety question, per the criteria of l 10CFR50.59, on the basis of the following justification.
4.1 Will the probability of an accident previously evaluated in the UFSAR be increased?
No. A change in RHR flow rate during reduced inventory operation will potentially impact
! transients for which a minimum RHR flow rate assumptions are specified. Since the change
! represents an merease in the RHR flow rate. no transients will be adversely affected. Also, it is necessary to assure adequate circulation of coolant in the RCS to prevent boron stratification and support the boren dilution transient mixing assumptions. The increase in flow rate is beneficial to maintaining the appropriate mixing assumptions. The evaluation further provides the acceptable limiting relationship between the maximum allowable RHR flow rate with the coincident allowable minimum RCS water level to preclude vortexing and subsequent air entrainment in the operating RHR train, thus maintaining RHR operability under reduced ..
inventory conditions.
4.2 Will the consequences of an accident previously evaluated in the UFSAR be increased?
No. A change in RHR flow rate during reduced inventory operation will potentially impact transients for which a minimum RHR flow rate assumptions are specified, as well as boron mixing assumptions. As discussed in the response to Question 4.1, the increase in RHR flow rate to 3300 gpm will not invalidate the Braidwood accident assumptions, and thus, the consequences of the accidents evaluated in the UFSAR would not be increased. The acceptable limiting relationship for maimum allowable RHR flow rate versus allowable minimum RCS water level is provided to preclude vortexing and subsequent air entrainment in the operating RHR train. Thus, RHR operability under teduced inventory conditions will be maintained.
4.3 May the possibility of an accident which is different than any already evaluated in the UFSAR be created?
No. Acceptable RHR flow rates versus specific reduced inventory water level conditions have l been provided. This relationship assures that vortexing and air entrainment of the RHR system l will be avoided. Also, normally only one RHR train will be in operation at any given time.
l Thus, should RCS conditions result in the impairment of the operating RHR train, the other train which is maintained in an operable condition per the Technical Specifications can be used to maintain the decay heat removal function. The other considerations which can be impacted by a change in RHR flow rate (decay heat removal. thermal and boron mixing in the vessel and RCS, and control valve cavitation) have s.lso been evaluated. The increase in RHR flow rate does not adversely affect these considerations. No new accident is created and no new single failures have been identified. Safety related systems and equipment required to mitigate the consequences of
m _ _ -. . - _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ . . . _ . _ _ _ _ _ _ _. _
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OCT 10 '96 19:39 FR WEC 412 374 6337 TO 818154583803 P.08/11 i SECL-%-193 ,
Page 7 of 10 )
I postulated accidents are unaffected and will operate as required. For those cases where two trains are in service, no new accident will be creeted.
4.4 Will the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR be increased? l No. When the reactor coolant level in the RCS loop piping is lowered, there is a potential for air to be drawn into the RHR suction line due to RCS loop level fluctuations and/or the development of a vortex. Air entrainment into the RHR system could cause air binding of the RHR pumps and thus, result in the inadvertent loss of decay heat removal capability. %e
, tendency for vortcx formation at the RHR suction line and subsequent air entrainment is a function of the RCS water level and the RHR flow rate. An acceptable limiting relationship for maximum allowable RHR flow rate versus allowable minimum RCS water level is provided to preclude vortexing and subsequent air entrainment in the operating RHR train, assuring the operability of the RHR system under reduced inventory conditions.
4.5 Will the consequences of a malfunction of equipment important to safety previously evaluated in l the UFSAR be increased? ..
No. The maximum allowable RHR flow rate as a function of RCS water level is sufficient to preclude vortex formation which could lead to pump cavitation and loss of the operating RHR train. Maintaining the prescribed flow rate / water level will preclude air entrainment due to
! vortexing and subsequent loss of RHR operation. Plant systems assumed operable to mitigate the consequences of an accident are will remain operable. There is no change to any analysis assumptions due to the malfunction of safety related equipment resulting from the increase in i
RHR flow rate.
l 4.6 May the possibility of a malfunction of equipment important to safety different than any already evaluated in the UFSAR be created?
No. Acceptable RHR flow rates versus specific reduced inventory water level conditions have l been provided. This relationship assures that vortexing and air entrainment of the RHR system will be avoided. Also, normally only one RHR train will be in operation at any given time.
Thus, should RCS conditions result in the impairment of the operating RHR train, the other train which is ma!ntained in an operable condition per the Technical Specifications can be used to maintain the decay heat removal function. The increase in RHR flow rate does not create any I new failure modes that could adversely impact safety related equipment. No new equipment malfunctions have been introduced. For those cases where two trains are in service, no new equipment malfunction will be created.
4.7 Will the margin of safety as defined in the bases to any technical specifications be reduced?
i No. The increase in RHR flow rate does not violate the RHR flow rate requirements for Mode 6 l-in the Byron /Braldwood Technical Specifications. The Technical Specifications place limitations on the RHR system by specifying a minimum flow requirement for the purposes of decay heat
i OCT 18 '% 19:39 FR LEC 412 374 6337 TO B18154583803 P.09/11 e
SECL 96-193 Page 8 of 10 removal, maintain the reactor coolant temperature rise through the core within design limits, prevent thermal and boron stratification in the core, and ensure that inadvertent boron diktion events can be identified and terminated by operator action prior to the reactor returning writical.
They do not, however, contain restrictions based on minimizing air entralnment in rho RHRS as a result of vortexing which may occur during reduced inventory operation under certain conditions. As stated in the Technical Specification Bases, operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles can result in vortexing or cavitation conditions which could cause the loss of RHR pump operation. Care must be taken in ;
determining the RHR flow rate to prevent the loss of the RHR pump and subsequent loss of the l
RHR loop for decay heat removal. The analysis presented in this safety evaluation supports the l basis for Techn! cal Specification 3/4.9.8 to assure the operability of the RHR system. Thus, the margin of safety provided by the Technical Specification' shutdown margin limits is not reduced.
5.0 CONCLUSION
S An increase in RHR flow rate during Modes 5 and 6, specifically when the RCS is in a reduced i inventory condition has been evaluated for Braidwood Units 1 and 2. The effect on potential ,,
. vortexing at the RHR suction has been evaluated and acceptable RHR flow versus RCS water level limits have been provided to preclude loss of the RHR heat removal capability. It has demonstrated ;
the acceptability of extending the approved range of operation of an RHR train to a range of 1000 gpm to 3300 gpm. Also, previous considerations for RHR operation have been reviewed and I
determined to be unaffected by the increase in RHR flow rate. Therefore, i,t is concluded that the increase in RHR flow rate in accordance with the recommended limits on RCS water level does not represent an unreviewed safety question as defined in 10CFR50.59, and does not require a change to the plant technical specifications.
OCT 18 '% 19:40 FR LEC 412 374 6337 TO 818154583803 P.10/11 l
l SECL 96193 Page 9 of 10
6.0 REFERENCES
t
- 1. U.S. NRC Generic Letter 87-12, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled."
- 2. Commonwealth Edison letter to the NRC, Morgan to Miraglia, dated September 25,1987.
1
- 3. Updated Final Safety Evaluation Report (UFSAR) for Byron /Braidwood Stations, Commonwealth Edison Company.
1
! 4. NUREG-1276, Technical Specifications for Braldwood Station Units 1 and 2, Docket Nos. 50456 and 50-457
- 5. Westinghouse Safety Evaluation SECL 89 867, ' Reduction in the Minimum RHR Flowrate l During Mid Loop Operation," Byron & Braidwood Units 1 & 2, July 1989.
l 6. WCAP-12207, " Reduction in the Minimum RHRS Flowrate During Mid-Loop Operation for Byron and Braidwood Power Plant Units 1 & 2,' June 1989. ..
- 7. WCAP-11916, " Loss of RHRS Cooling While the RCS is Partially Filled," Revision 0,
! July 1989, prepared for use by the members of the Westinghouse Owners Group.
I
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SECL-%I93 Page 10 of 10
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BwAP 1205-6T9 Revision 1 10CFR50.59 SAFETY EVALUATION VALIDATION FORM
- 1. LIST the documents implementing the proposed activity (eg Modification Number, etc.) and include revision data (where appropriate).
Westinchouse SECL No. 96-1931 RHR Flow Durino Reduced Inventory conditionst BwOP RC-4. Revision 6 //cce - Me Cr% few Aetest%ces Ch y) x I 2. IDENTIFY the activity addressed by the Safety Evaluation being validated.
l l 5 Same activity as listed above in Step 1. Proceed to Step 3.
O Different activity:
- a. List the documents implementing that activity (eg: Modification Number) and include revision date (where appropriate).
l
- b. Describe and justify any differences (ogs different train, unit, station, etc.) between the tuo activities.
l l
- 3. PZVIEW the Safety Evaluation being validatid tnd ensure the following:
- a. The Safety Evaluation was performed by a procedure implementing NOD-TS.11; such as:
- BwAP 1205-6T1, Revision 1 or later
- BAP 1210-5T1, Revision 1 or later
- ENC-QE-06.1, Revision 3 or later
- b. The proposed activity does not extend beyor.d the plant mode bounds assumed in the Safety Evaluation.
- c. There are no know facility changes (eg Modifications, etc.) in
. place since the safety Evaluation was writton which would invalidate-it.
- d. The Safety Evaluation conclusions are reasonable, well supported and documented, and have determined that prior NRC review / approval is NOT required.
- 4. PROCESS this Safety Evaluation Validation as follows:
- a. Attach a copy of the Safety Evaluation.
- b. Complete the Preparer & Reviewer verifications below and sign.
- c. Initiate an OSR to address the proposed activity and this validation.
PREPARER of this Validation (must be qualified per BwAP 1205-6T7):
(Print) J. Tolar (Sign) v MA (Date) (8 / I6 / D REVIEWER of this Validation (must qualified per BwAP 1205-6T7):
(Print) R. Krbec (Sign)
(Date)/C // [/b (Final) l 1
919(062493)