ML20198P975

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Pages Revising Loss of Normal Feedwater (Lonf) Analyses to TS 2.2.1,TS Bases Change to Floor Value for Thermal Margin Low Pressure Reactor Trip & Proposed FSAR Changes
ML20198P975
Person / Time
Site: Millstone Dominion icon.png
Issue date: 12/28/1998
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20198P949 List:
References
NUDOCS 9901070173
Download: ML20198P975 (51)


Text

{{#Wiki_filter:. ~.. ~... w a w TABLE 2.2-1 REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LIMITS o W

  • s FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES og 1.

Manual Reactor Trip Not Appilcable Not Applicable. 2. Power Level-High E8 Four Reactor Coolant Pumps s 9.6% above THERMAL POWER, f 9.7% Above THERMAL POWER, m e' Operating with a minimum setpoint of with a minimum of s 14.7%, f 14.6% of RATED THERMAL 'of RATED THERMAL POWER, and a POWER. maximum of f 106.7% of RATED THERMAL POWER. 3. Reactor Coolant Flow - 2 91.7% of reactor coolant 2 90.9% of reactor coolant flow Low (1) flow with 4 pumps operating *. with 4 pumps operating. 4. Reactor Coolant Pump 2 830 rpm 2 823 rps ( Speed - Low "a 5. Pressurizer Pressure - High 1 2400 psia 1 2408 psia 6. Containment Pressure - High 5 4.75 psig 5 5.24 psig ~> 680 psia 1 672 psia 7. Steam Generator Pressure - wal Low (2) (5) W.5% 99,5 4, ( Ok 8. Steam Generator Water 2-3fn95 Water Level - each 2-35d5 Water Level - each t y3 Level - Low (5) steam generator steam generator OF 9. Local Power Density - Trip setpoint adjusted to not Trip setpoint adjusted to ?* High (3) exceed the limit lines of not exceed the limit lines g Figures 2.2-1 and 2.2-2 (4). of Figures 2.2-1 and 2.'2-2 (4). c g q-w n 3

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either:
a. The reactor coolant flow rate measured per Specification 4.g.6.1, or e

y 9 ?

b. The minimum value specified in the CORE OPERATING LIMITS REPORT.
  • k

1 - - :e, 'non LIMITING SAFETY SYSTEM SETTINGS I I BASES-r Stoa.v Cenerator Watee-Level -_tpw . 18-1 The Steam Generator Water Level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design j-pressure of the reactor coolant system will not be exceeded. The p;;ified tetreint previd;; ellewence th;t th:re wiH be :;ffisient water inventory in - 3 -the ste:: g;r,e..tvr si t^ e ti c' '. rip te previde =?rgia ef "cr: then 10 n ) ! nut;; bef r; ;;;ili;ry feed.eter.3 required. j Local Power Density-Hiah The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX' monitoring, is provided to ensure that the peak local power density in the fdel which corresponds to fuel centerline melting will not occur as a conse-quence of axial power ma1 distributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level. The trip is automatically bypassed below 15 percent power. The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In -) - addition, CEA gr.oup sequencing in accordance with the Specifications 3.1.3.5 ] and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-i High trip is assumed. Thermal Marcin/ Low Pressure The. Thermal Margin / Low Pressure trip is provided to prevent operation j-when the DNBR is less than 1.17. r e. i j i l 1 W '). l 28,/I,/2,[I,/f MILLSTONE - UNIT 2-B 2-6 Amendment No.

4 ,,m.m i... i n.- i LlHITING SAFETY SYSTEM SETTINGS DASES i Thermal Marcin/ Low Pressure (Continued) ] The trip is initiat henever the reactor coolant system pressure signal drops below either-% 6& la or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3._6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed. Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is,provided which includes: an allowance of 5% of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2*F to compensate for potential temperature measurement uncertainty; and a further allowance of 74 psi to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective termino- ,)) tion of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 74 psi allowance is made up of a 5 psi bias, a 19 psi pressure measurement allowance and a 50 psi time delay allowance. Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuring transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these values. No credit was taken in the accident analyses for operation of this trip; Its functional capability at the specified trip setting is required to e'nhance the overall reliability of the Reactor Protec-tion System. Milt.SIONE - UNIT 2 B 2-7 Amendment No. JE, JJ, JJJ 0048

Docket No. 50-33Q B17S19 l Millstone Nuclear Power Staticn, Unit No. 2 Proposed Revision to Technical Specifications Loss of Normal Feedwater Flow Retyped Pages l L l l l l J t December 1998

en- -..... TABLE 2.2-1 3 $:h - REACTOR PROTECTIVE INSTRUNENTATION TRIP SETP0 INT LIMITS y -l' FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES 1. Manual Reactor Trip Not Applicable Not Applicable. 1 2. Power Level-High ,.4 Four Reactor Coolant Pumps s 9.6% above THERMAL POWER, s 9.7% Above THERMAL POWER, Operating with a minimum setpoint of with a minimum of s 14.7% s 14.6% of RATED THERMAL of RATED THERMAL POWER, and a POWER. maximum of s 106.7% of-l RATED THERMAL POWER.. 3. Reactor Coolant Flow - 191.7% of reactor coolant 190.9% of reactor coolant flow Low (1) flow with 4 pumps operating *. with 4 pumps operating. 4. Reactor Coolant Pump 1 830 rpm 1 823 rpm { Speed - Low 5. Pressurizer Pressure - High s 2400 psia s 2408' psia = 6. Containment Pressure - High 1 4.75 psig s S.24 psig wD 7. Steam Generator Pressure - 1 680 psia 1 672 psia -k Low (2) (5) ws 4{ 8. Steam Generator Water 1 48.5% Water Level - each 1 47.5% Water level - each l Level - Low (5) steam generator steam generator s w e+ Em 9. Local Power Density - Trip setpoint adjusted to not Trip setpoint adjusted to High (3) exceed the limit lines of not exceed the limit lines Dw Figures 2.2-1 and 2.2-2 (4). of Figures 2.2-1 and ?? 2.2-2 (4). M

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either-
a. The reactor coolant flow rate measured per Specification 4.2.6.1, or

+

b. The minimum value specified in the CORE OPERATING LIMITS REPORT.

6 ,P 4 ) -e-mr. - - - + - -. ~

~.-- L LIMITING SAFETY SYSTEM SETTINGS i macre Steam Gs.ierator Water Level - Low i .The Steam Generator Water Level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded. i Lgcal Power Density-Hiah The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a conse-quence of axial power ma1 distributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL 4 . SHAPE INDEX is calculated from the upper and lower ex-core neutron detector l channels. The calculated setpoints are generated as a function of THERMAL POWER level. The trip is automatically bypassed below 15 percent power. 4 i The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted 1 i for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 l and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed. Thermal Marain/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation whea the DNBR is less than 1.17. k 4 i 5 I MILLSTONE - UNIT 2 B 2-6 Aeendment No. 77, (J. 57, JJ,~ JJ), oax

LIMITING SAFETY SYSTEM SETTINGS matre Thermal Narain/ Low Pressure (Continued) .The trip is initiated whenever the reactor coolant system pressure signal drops below either 1865 psia or a computed value as described below, whichever l is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.'.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any-anticipated operational occurrence prior to a Power Level-High trip is assumed. Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes: an allowance of 5% of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2*F to compensate for potential temperature measurement uncertainty; and a further allowance of 74 psi to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 74 psi allowance is made up of a 5 psi bias, a 19 psi pressure measurement allowance and a 50 psi time delay allowance. Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above It of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuring transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these values. 60 credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. MILLSTONE . UNIT 2 B 2-7 Amendment No. 77 J7, J7), JJ7, 0434

i Docket No. 50-336 B17519 4 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications FSAR Change Loss of Normal Feedwater Flow i December 1998 I

1 pD C /M a'6{ MNPS-2 FSAR Cjog 7A,g g 77% OMy 4.3.8.1.7 Relief and Safety Valve Discharge Temperature Temperatures in the pressurizer safety valve discharge lines are measured and indicated in the main control room. A h:gh temperature in one of these lines is an indication that the associated valve may be leaking. High temperature alarms are provided to alert the operator to this condition. 4.3.8.1.8 Quench Tank Temperatures The temperature of the water in the quench tank is indicated in the main control room. A high temperature alarm is also provided. A high quench tank temperature alerts the operator to the requirement for coolicg of the tank contents. 4.3.8.1.9 Reactor Vessel Flange Seal Leakage Temperature This RTD is located in the reactor vessel flange leak-off line. The channel is displayed in the main control room and actuates a high temperature alarm. A high temperature is indicative of excessive leakage past the first reactor vessel flange seal. 4.3.8.1.10 RCS High Point Vents Leakage Temperature Thermocouples installed on the downstream side of each solenoid valve train are utilized to monitor leakage past the system solenoid valves. Under normal operating conditions, the thermocouples will measure the ambient temperature in the piping downstream of each solenoid valve train. The output of each thermocouple is continuously recorded by the gi plant computer. Any leakage through the system valves will cause an increased temperature in the downstream piping which will be detected by the thermocouple. At a predetermined setpoint, an alarm will be actuated, identitying a high temperature reading on the appropri-ate thermocouole. Once a high temperature alarm is received, further actions will be governed by the Technical Specifications for reactor coolant system leakage. 4.3.8.2 Pressure 4.3.8.2.1 Pressurizer Pressure Four inoependent narrow-range pressure channels are provided for initiation of protective systems action. The pressure transmitters are connected to the upper, ortion of the pressurizer via the upper level measurement nozzles and measure pressurizer vapor pressure. All four channels are indicated in the main control room and actuate separate high, low, or low low pressure alarms in the control room. The protection actions these pressure signals initiate are: 1. Reactor trip on high primary system pressure. The reactor trip signals are also used to open the PORVs;

2..

Safety injection systorn actuation on low low primary system pressure; assw2 4.3 24 August 1998

IVINPS-2 FSAR 3. Reactor trip on a low primary system pressure. The set point is a function of the coolant temperatures in the hot and cold legs. The variable set point has high and low limits alarmed in the control room and is not allowed to decrease below l 4.3.8.2.2 Pressurizer Pressure Two independent pressure channels provide narrow-range pressure signals for controlling t4e pressurizer heaters and spray valves. The output of one of these channels is manually selected to perform the control function. During normal operation, a small group of heaters are proportionally controlled to offset heat losses. If the pressure falls below a low-pressure set point, all of the heaters are energized. If the pressure increased above the high-pressure set point, the spray valves are proportionally opened to increase the spray flow rate as pressure rises. An interlock will prevent operation of the back-up heaters in the event of a high level error signal concurrent with a high-pressure condition. These two channels are also used to provide pressurizer pressure signals to the RRS. The two channels are continuously recorded in the main control room and are provided with high-and low-pressure alarms. 4.3.8.2.3 Pressurizer Pressure Two low-range pressure measurement channels provide a control room indication of RCS pressure during plant startup and shutdown in the main control room. They also provide independent pressure signals to the shutdown cooling suction isolation valves (refer to Section 9.3.4.1) which prevent these valves from opening above a selected set point. If the shutdown cooling suction va!ves are open when the pressure exceeds a selected set point (280 psia), an annunciator receives a signal to alarm from these pressure channels. The channels also provide signals to actuate the PORVs for LTOP. These two instrument channels are independent, redundant, and diverse in that the transmitters are purchased from different vendors. 4.3.8.2.4 Quench Tank Pressure This measurement channel provides a quench tank pressure indication in the main control room and actuates a high-pressure abrm. High quench tank pressure indicates that the tank has received a discharge from the safety or relief valves. The operator will then take action to restore the tank to normal operating conditions. 4.3.8.3 Level 4.3.8.3.1 Pressurizer Level Two pressurizer level channels are used to provide two independent level signals for control of the pressurizer liquid level. These signals are used to doenergize the pressurizer heaters on low low pressurizer level to prevent heater burn out, provide input to one pen in the two-pen recorder in the control roorn, and actuate high-and low-pressure pressurizer level alarms in the main control room. The second pen on the level recorder records the programined pressurizer level setpoint computed by the RRS as a function of the average [(y[ reactor coolant temperature. The level transmitters are compensated for the steam and water densities existing in the pressurizer during normal operation. 4s w p2 4.3-25 August 1998

- ~ - ~ - MNPS-2 FSAR protrip alarm provides audible and visual annunciation in addition to CEA withdrawal prohibit signals. O and Os are processed and buffered for remote display on the main control board. O is also taken to the Control Element Drive System (CEDS) and the Plant Computer for use in the power dependent insertion limit calculation. 7.2.3.3.3 Low Reactor Coolant Flow-This reactor trip is provided to protect the core against DNB in the event of a coolant i flow decrease. The flow measurement signals are provided by measuring the differential pressure across each of the two steam generators. Each steam generator differential pressure signal is proportional to the square of the steam generator mass flow rate. These signals are summed to provide a signal that is proportional to the square of the reactor ' vessel mass flow rate. The measured signal is compared to a pre-determined trip setpoint in the reactor protection system trip bistable. This configuration is shown in Figure 7.2-4, and is repeated in each of four redundant channels. A reactor trip is (MM initiated when the measured value falls below the bistable trip setting in two-out-of-four [" coincident channels. j i Pre-trip alarms are similarly provided to warn of decreased coolant flow conditions. The Zero Power Mode Bypass switch allows this trip to be bypassed for subcritical testing of CEDMs. The trip bypass is automatically removed above 10 percent power. 4 Provisions are made in each channel which establish different low flow trip and pre-trip bistable setpoints depending on the operating status of the four reactor coolant pumps. The switching arrangement of Figure 7.2-4 shows this feature. The flow dependent setpoint selector switch is the same used by the thermal margin / low pressure trip i function described in section 7.2.3.3.7. Reactor coolant pump operating requirements, and flow dependent setpoint selector switch position and surveillance requirements are defined by Technical Specifications. 7.2.3.3.4 Low Steam Generator Water Level An abnormally low steam generator water level indicates a loss of steam generator secondary water inventory. If not corrected, this wou AreIt in a loss of capability for removal of heat from the RCS, Y4c4 &A v % ynrm F) } dryc.a t does & acc.w j The low steam generator water level reactor trip protects against the los of feedwater flow accident (see Section 14.2.7) and assures that the design pressure of the RCS will $,N - not be exceeded. The trip set point specified in Table 7.21 assures that sufficient . water inventory _will be in the steam generator at the time of trip to gcis :: !:::t (aged " "~- before the auxiliary feedwater (AF) k mM d fci the n.a.cVel vi decey host. cletws wha.,/ Sk f nmese cleay he / an.) yewv ricam cyn ca fu wak level. A reactor trip signalis initiated by two-out-of-four logic from four independent chan-nels. Each channel actuates on the lower of two signals from two downcomer level - differential pressure transmitters, one on each steam generator. Audible and visual - ).; pretrip alarms are actuated to provide for annunciation of the approach to reactor trip conditions. - isawa - 1 7.2 8 June 1998

- - - ~ - I'. f l [, MNPS-2 FSAR , initiated by two-out-of-four coincidence logic from the four independent measuring channels if the pressurizer exceeds 2400 psia. This signal simultaneously opens the ' power-operated. relief _ valves (PORV). 1 The trip signals are provided by four independent narrow range pressure transducers measuring the' pressurizer pressure. Pretrip alarms are initiated if the pressurizer pressure exceeds 2350 psia as indication of the approach to reactor trip conditions. J 7.2.3.3.7: Thermal Margin / Low-Pressure Trip The TM/LP trip is provided for two purposes. The thermal margin portion of the trip, in conjunction with the low reactor coolant flow trip, is desigr.ed to prevent the reactor [ core safety limit on DNB from being violated during anticipated operational occurrences. The low pressurizer pressure portion of the trip functions to trip the reactor in the event i of a LOCA. /$66 A reactor trip is initiated whenever the RCS pressure signal drops below either hMo) -+069 psia or a computed value as described below, whichever is higher. The comput-ed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of RCPs operating and the axi_al offset. Consi; tent with the . Technical Specifications, the minimum value of reactor coolant flow rate, the maximum azimuthal tilt, the maximum CEA deviation permitted for continuous operation are e assumed in the generation of this trip function. In addition, CEA group sequencing !.i 'accordance with the Technical Specifications is assumed. Finally, the maximum insertion of CEA banks which can ' occur during any anticipated operational occurrence j prior to a High Power Level trip is assumed. i Figure 7.2-6 and 7.2-11 describe the operation of this trip system. The higher of the 1-two inlet temperatures is added to a correction term proportional to thermal power. This feature compensates for temperature stratification errors in the coolant pipe. Figure 7.2-12 shows a block diagram of the thermal power calculation. The calculation begins with the generation, by tempenture transmitters, of currents representing the cold and hot leg temperatures in each loop. By forcing these currents t: through precision resistors and utilizing the resulting voltage drops, voltages repres-ent-ing cold leg temperatures (T,i and T ) and hot leg temperature (T,) are sent to the e2 calculator. The latter signal is the average T for the two loops. qggqi 3 in the calculator, the higher cold leg temperature signal is selected and subtracted from 4 the hot leg temperature signal to determine the temperature rise. The calculator j generates terms proportional to the first and second powers of the temperature rise and to the product of temperature rise and cold leg temperature. These three terms L reprosent thermal power for tuutpump operation and steady state conditions, account-ing for coolant density, specific heat, and flow rate variations with temperature and power. The sum of these terms represents the core power for four-pump operation under _ steady' state or mild transient conditions. This sum is multiplied by a factor F, which is j [ c unity for four-pump operation and less than unity for other configurations. This factor compensates for the fact thati for a given power, f'ie temperature rise is greater for reduced flovJ. The multiplying factor is selected by the flow dependent setpoint selector. switch (S3), which also selects the Lc,w Flow trip setpoints. - 752.MP2 7.2-10 June 1998

MNPS-2 FSAR . The coefficient of the term proportional to the temperature rise (K,) is set by the potentiometer labe!ed "AT Power Calibrate" on the Reactor Protective System Calibra-tion and Indication Panel (RPSCIP) front panel. This factor is adjusted to make the thermal power calculation agree with the plant calorimetric calculation. The thermal power (B) is subtracted from the nuclear power (p), generated by the NI Channel, and the difference is displayed on a meter with a range of -10 percent to + 10 percent of full power. The meter has adjustable upper and lower setpoints. The contacts energize a local light when the deviation goes outside the range defined by the setpoints. To make the nuclear power signal agree with the thermal power and/or the plant calorimetric calculation, a potentiometer labeled " Nuclear Power Calibrate" is provided on the RPSCIP front panel. This potentiometer adjusts the gain of the NI Channel from 0.8 to 1.33. An auctioneering circuit selects the higher of nuclear power or thermal power for use in the remainder of the system. , The signal Q, the maximum of nuclear or thermal power, is modified by a CEA position function. The resulting signal is then augmented by an axial factor which is generated in the LPD Trip section as shown on Figure 7.2-13 and described in 7.2.3.3.10. The resulting signal is called Om. A pressure setpoint P. is calculated as a linear function Om and of the modified inlet temperature described above. The flow dependent coefficients of the linear function are selected by the S3 switch. An auctioneering circuit selects the maximum of this calculated pressure setpoint and a 3 4 constant pressure P., and sends the resulting signal to the trip unit as a downscale trip setpoint. Trip will occur if the primary pressure drops below the calculated setpoint f L or below+05% psia, whichever is larger. A pretrip setpoint, 75 psi above the trip point, ggo} is also generated. The trip signal is initiated by a two-out-of four coincidence logic from four independent safety channels, and audible and visual pretrip alarms are actuated to provide for annunciation on approach to reactor trip conditions. The pretrip action also initiates a CEA withdrawal prohibit. The zero power mode bypass switch allows this trip to be bypassed for low power testing. The trip bypass is automatically removed above 104 percent power. The Thermal Margin trip setpoint is processed and buffered for remote display on the main control board in four dualindicators which compare the trip setpoint with indicat-ed pressurizer pressure. 7.2.3.3.8 Loss of Turbine The trip for loss of turbine is an equipment protective trip and is not required for reactor orotection. (Refer to Chapter 14.0). ' This trip is initiated above a preset power level, by actuation of 2 of 4 low hydraulic fluid pressure switches associated with the turbine-generator control systems. Its .. purpose is to help avoid the lifting of the steam generator safety valves during the system transient after a turbine trip, thus extending the service life of these valves. 782.0#2-7.2-11 June 199Bl a

MNPS-2 FSAR TABLE 7.2-1 REACTOR TRIP AND PRETRIP SET POINTS Pretrip Alarm Na Reactor Trio Set Point Trio Set Point 1. Reactor Coolant Pump Underspeed*

  • N.A.

1830 rpm 2. High Power Level 2% below trip .s.10% above 97 203 setpoint measured power O 3. Low Reactor Coolant Flow'*

4. Pump Operation, %

N.A. 91.7 97-285 4. Low Steam Generator Water Level, % (Auctioneered low of SV .) ~ yg,5 SG #1, SG #2) - 5. Low Steam Generator Pressure * * *, psia (Auctioneered low of SG #1, SG #2) 780 680 h'I) 6. High Pressurizer Pressure, psia 2350 2400 7. Thermal Margin / Low Pressure 75 psia above Variable trip set trip set point p #oint with minimum of 4650 psia $5I'd 8. Loss of Turbine ' * * *

  1. f05

-(Low Hydraulic Fluid Pressure) psig N.A. 1500

9.

High Containment Pressure, psig N.A. 5

10. Manual Trip (Push Buttons)

N.A. N.A. .1 1. Local Power Density 20 kW/ft 21 kW/ft

    • Manualinhibit permitted below 104 percent power: automatically removed above 1_04 percent power.
  • **Manualinhibit permitted below 780 psia: automatically removed above 780 psia.

/* * *

  • Inhibited below 15% power.

T72-1.MP2 ',' .1 of 1 June 1998

. ~ - MNPS-2 FS.AR [ The SGFPs are located in the cycle between the low pressure heaters and the high pressure heater. i! To maintain the proper feedwater chemistry a secondary chemical feed system is provided 4 for injecting hydrazine as an oxygen scavenger and ammonium hydroxide to control carbon dioxide content and pH. The amount of hydrazine injected into the feedwater by the chemical feed pumps is automatically controlled by a signal from the hydrazine analyzer. ' The amount of ammonium hydroxide injected is autornatically controlled by feedwater j conductivity measurement. The point of chemicalinjection during normal operation is in i the condensate ret' rn header from the Condensate Polishing' Facility (CPF), upstream of u i the bypass valve. An altemate point of injection, when the CPF is bypassed, is in the CPF 1/W qv-s s discharge header downstream of the CPF bypass valve and upstream of the steam jet air / ejectors. i } Feedwater regulating and control is described in Section 7.4 b a b. '10.4.5.3 Auxiliary Feedwater System-y 'M Ec) The AFWS, shown in Figure 10.4-2,is designed to provide feedwater for the removal of sensible and decay heat, and to cool the primary system to'300 Fin case the main ) condensate and SGFP are inoperative due to loss of normal electric power sources or the main steam. A single motor driven pump is capable of adequately removing decay heat. h*M 4 This has been demonstrated d.g a c.....r....,...-......, analysis of the loss of feedwater event (AFW4esign basis). - (Reference /10.4-1 th;m.;h 10.0 0). The suction condition ' ) f yields a net positive' suction head (NPSH)in excess of that required. The AFWS rn (r,'h be used for normal system cooldown to 3OO*F. The AFWS supplies feedwater from the conden' ate storage tank (CST) to the steam generators for evaporation and heat absorp-s tion. Reactor decay heat and sensible heat are transferred to the steam generators by natural circulation of the reactor coolant if power is not available for the reactor coolant dumps (RCP). jf gp in order to perform its safety related function (per Section 14.2.7) assuming single failure, . the auxiliary feedwatar system (AFWS) is comprised of two full capacity wbsystems. One suusystem consists of two motor-driven AFW pumps, that are automatically connected to the diesel generators in the event of a loss of offsite power. The second subsystem consists of one turbine-driven pump that is independent of AC power, The turbine-driven (97-97) pump has a maximum capacity of 600 gpm at 2437 feet tdh and the(two motor-driven pumps have a constant 300 gpm capacity each at 2437 feet tdh. W "$ y Q' *D as The auxiliary feedwater pumps (AFP) are located in two separate pump rooms at elevation 1' 6" in the Turbine Building. The Turbine Buitding is protected from potential flooding by the flood wall system as shown in Figure 2.5-18 of the FSAR. Access to the first room h;~h'h which houses the two motor-driven AFPs is by stairs leading down from the ground floor at elevation 14'-6* The enclosure over the pump room stairwell serves as a protective barrier against direct water streams into the Pump Room due to a possible overhead pipe failure. The second room which houses the turbine driven AFP is a vault physically L separated from the motor-driven AFP room by a reinforced concrete wall. The only access " means to this room is through a water-tight fire door, tos4 w 2 10.4-6 October 1998

j MNPS 2 FSAR 3 REFERENCES ] 10.4-1 Let r from W. G. Council to rector of Nuclear Re tor Regulation, " Millstone clear Power Station, U

o. 2 Automatic initi ion of Auxiliary Feedwater "

dated May 20,1980. (gr.-iG } 1 .4-2 Letter from W. G ouncil to Director of N lear Reactor Regulation, " illstone "M' Nuclear Power tation Unit No. 2 Auto tic initiation of Auxiliary F edwater," dated Octo r 31,1980. 10.4-3 " RET N-02: A Program for Tran ent Thermal Hydraulic nalysis of Comp Flui Flow Systems," EPRI NP-1 00-CCM, dated Octo 1984. N' Y~l. "$slIs hac (dnd Alo.,2 fSAR Chaapfri /0 dos 5 of A/ coma / 5 1 0 "

  • b f /v w Tr e sien / asi tt (s h fuc,J A % ilo w y
    • N" N% " E/hf O y 9, fe g,, o, T, em en s &we<

(c'/fWa f/w fle<9us /- /99 g, ) }. 1054.MP 10.4-13 June 1998

I AM CR44GQ MNPS-2 FSAR pgg y c,n g r z w QMV core was, however, redeveloped based on event specific XTGPWR (Reference 14.2-3) calculations. Core radial power distributions from full power EOC XTGPWR cases with differences between the hot-region inlet temperature and the cold-region inlet temperature are used to determine the power split between the halves of the core as a function of the difference in inlet temperatures. Since the temperature differences used in the XTGPWR cases meets or exceeds the inlet temperature difference calculated by ANF-RELAP during the transier.t calculation, the power splits used in ANF-RELAP are bounding. The XTGPWR 1 [ 13 2 calculations for the single MSIV closure event differ from those used in the SLB event. The SLB analysis requires ' power distribution' data for all-rods inserted minus the most. - reactive stuck rod, whereas in the single MSIV closure event it is assumed that an all rods out power distribution is appropriate. The limiting results were obtained from the case with the lower stearn flow rates. The results of the limiting EOC analysis are given in the event summary, Table 14.2.4-3, and in Figures 14.2.4-1 through 14.2.4-5. As indicated in the event summary table the secondary safety valves open early in the transient limiting the temperature rise on the hot (she') side of the core associated with the closed MSIV. The reactor trips on low steam genera-tot pressure which terminates the power rise. i The peak LHR and Minin:um Departure From Nucleate Boiling Ratio (MDNBR) are predicted iT'IM. to occur on the cold side of the core at the time of reactor trip. The peak LHR is 18.7 kW/ft and the deterministic MDNBR was found to be 1.40. Thus it is concluded that theh[h') DNB limits will not be violated and that fuel failures are precluded during the single MSIV ) closure event. ) j The secondary side safety valve setpoints were modeled with a.t.3% drift allowance, and the flow characteristics were modeled with a 3% allowance for accumulation. The maximum secondary side pressure is 1092 psia, which is less than 110% (1100 psia) of g,,332 design pressure. 14.2.4.7 Conclusion The calculated minimum DNBR for the single MSIV closure event is above the critical heat i i flux correlation safoty limit, so the DNB SAFDL is not exceeded in this event. The peak 1r.112. LHR is less than the 21 kW/ft limit to centerline melt. The maximum secondary side pressure is below 110% of design pressure. Thus, the single MSIV closure event has been demonstrated to meet all required acceptance criteria, 14.2.5 Steam Pressure Regulator Failure I Millstone Unit 2 does not have any steam line pressure regulators, so this event is not h credible for this plant. No analysis needs to be considered for this event. ) j 14 2.6 Loss of Nonemergency AC Power to the Station Auxiliaries 1 18-in This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed. 1-4.2.7 Loss of Normal Feedwater Flow 14S2.MP2 34.2-7 September 1998

+ i MNPS-2 FSAR L 1'4.2.7.1 ' Event initiator The Loss of Normal Feedwater Flow transient is initiated by a trip of the main feedwater pumps or a malfunction in the feedwater control valves. 14.2.7.2 Event Description LThe loss o'f main feedwater flow willincrease tha secondary-side temperature and reduce the steam generator heat removal capability because the main feedwater system is supplying subcooled water to the steam generators. The rise in the secondary-side tamperature leads to a rise in the primary system coolant temperature. As the primary c system temperatures increase, the coolant ext, ands in the pressurizer which increanas the pressure by compressing the steam volume, maw sen,a34/ay uses o,, jam < stra, duy n /ve. The temperatures of the secondary sides and primary loops are controlled by the opening and closing of the e fety ve!ver er. 26 ase.T. No. The long-term cooling of the primary L system is assured by the secondary-side water inventory supplied by the Auxiliary ~ i Feedwater System (AFWS). Two motor-driven auxiliary feedwater (AF) pumps are automatically started upon a steam generator low liquid level signal, if a loss of offsite 4 3 power occurs, the motor-driven AF pumps are powered by the emergency diesels. In [ addition, a turbine-driven AF. pump can be manually actuated. (yg i 14.2.7.3 Reactor Protection j- ) System overpressure protection is provided by the primary and secondary system safety t valves.' A reactor trip occurs on low steam generator level with additional reactor protec- ) tion provided by the high pressurizer pressure trip, variable overpower trip, and the TM/LP trip. Reactor protection for the loss of Normal Feedwater Flow event is summarized in

Table 14.2.7-1.

f '14.2.7.4 Disposition and Justification t This event is only credible for rated power.and power operating conditions because the main feedwater system is not required to provide feedwater to the steam generttors fo'r 10ther reactor operating conditions. The consequences of this event for rat d e power i . operation bound the consequences for other conditions because of the higher initial stored i. energy in the primary system, the minimum steam generator inventory, and the greater t impact of the loss of feedwater flow on the secondary system. mE/'r d 4 L The ne term pressurizat' n and DNB as cts of this av nt are bounde y those of ent 14. This is due to et that in the alytical meth ology for this vent (Refer-t fg.p e e 14.2 5),it is i cated that rea or trip occurs t time zero coi cident with. rbine trip big i l n a low staam o aratar watar la al minnal[In Event 14.2.1, reactor trip is delayed until a~ Ad* high prwurizer pressure signal is received.l This results in a higher power level at trip, greater pressurization'and greater challenge to the SAFDLs than in Event 14.2.7. Long

term pressurization, if it occurs, is very gradual and is arrested by opening of the pressuriz-

- er code safety valves. fSER T ]j 7 --- (. 3 3 8 C)~~ The Loss of Normal Feedwater event is analyzed to assess the maximum expected . ressurizer level swell and the long term adequacy of the AFWS to restore and maintain p 'r p 4szwr j .14.2-8 September 1998 { u v ,_s. w-a

. _ -. _. _ _ _ _ -. _.. =. _ - MNPS-2 FSAR f \\ S S R t-C . steam generator inventory and prevent steam generator dryout. The ma -l level swell is examined to assure that the pressurizer does not become water solid. Th; :nt!;:!: )

d

- ;ne .. piain opmadiiv s,ui,d,i;vi... preT8ntative of tne EGC. uuonide.in as da '. bad fer S;;n; 1 1 2.1. 4CasStr D) The(ig!: fd!eise evi..;desed in :h;e er.e:pe e ;;e fe;;ere ei on AF pun.p t; Ots;t, ef cff9 p;;;;;nuMn;! rrM:;;n of th. seee;;r ::in; pum The disposition of events for the Loss'of Normal Feedwater Flow event is summarized in Table 14.2.7-2. t i 14.2.7.5 Definition of Events Analyzed Jddl6 kwo cases were anahd for a Loss of Norma odwater Flow transi t initiated from rated power ope ion:- g .(1 Initial conditions and s points biased to maxi e pressurizer liquid level. (2) Initial condition nd setpoints biased to nimize steam generator uid inventory. J . i - M -MEL4P . The analysis is performed with the SLOT".AX L"L, code (Reference 14.2-6). The k0 SLOT^X 'f.L code includes relevant aspects of the mass and energy balance of the primary and secondary systems. The following events are assumed to occur at th use M '--d-"-- initiation. ') (1) Reactor. trips on steam generator low level with specified time delay. 1 (2) Turbine conservatively trips with drnu::eneetss closure of turbine stop valve m, mm-, .r-m e gece.d a 64-nu f-f f m 4 5 [ '(3) Main feedwater valves are assumed to instantaneously close at event I initiation. l I(4). Back-up eaters in the pre urizer are assume o be fully operabt brough Out t transient. i (5).. tart sequence f emergency diesel nerators is initiate with 10 minute delay for deliv of 600 gpm of AF This conservative bounds the fai re (Wu,\\ - of one' AF p p (electric driven o urbine driven) to art. kl% / t 1-Additional conservative conditions are applied for analysis of each case to present the Greatest challenge to the event acceptance criteria. 9 -"nctric tube p Jguing is uunsiumed / it: S"O': :ncthode;egy hee ehessa that eyinineine iube vlugg:ng predeeee th; :nsi,t UM . w... .....m...v msun o.-- MluI ,L,A>ssf r p 14.2.7.6 : Analysis Results NMb l The av is initiated when h steam generato aff at the low leve13r[setpoint. The j

g. ( -

ste generators are co ervatively assumejao isolate at eventprtiation. ' For an i ):. I s of offsite power umec e orimarv c_oolantgumps are also trico(d at event initi n. i 14P2 MP2 14.2-9. September 1998 l ] 4 m

-.. -.. ~ r . MNPS-2 FSAR UN lAn event summar9 is presented in Table 1 .7-3. The transient resp ses are presented in Figures 14.71 through 14.2.7 4 for he minimum steam genera r inventory case and Figures 14 .7-5 through 14.2.7-8 for o maximum pressurizer i el case, in both cases, the tran nt execution time was 5 0 seconds. Th maximum pressurizer lig level is calculated to be 01 ft' at 180 seconds. i fficient steam volume r ains to preclude the expu son of liquid from the pressuri r s6fety valves. The minimum ste generator level is calculat to be approximately 6.5 ft occurring at about 1000 sec ds after the event initiatio. The steam generator leve teadily recovers (gg from this min' um level, thus ensuring co inued heat removal. i b '14.2.7.7 Conclusions A loss of normal feedwater event does not result in the violation of SAFDLs, peak pressur-izer pressure does not exceed 110% of the design rating and primary liquid is not expelled i - through the pressurizer safety valves; Adequate cooling water is supplied by the AFWS td allow a safe and orderly plant shutdown and tu prevent steam generator dryout. Thus, the loss of normal feedwater event has been demonstrated to me' et all required acceptance criteria. i 14.2.8 - Feedwater System Pipe Breaks inside and Outside Containment (%t.) is/% ) This event is not in the current licensing basis for Millstone Unit 2 and, therefore, is not analyzed. I L tas2. w a 14.2 10 September 1998

MNPS-2 FSAR REFERENCES 14.2-1 " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981. 14.2-2 "RELAPS/ MOD 2 Code Manual, Volum.1: Code Structure, Systems Models and - 1r-131 Solution Methods; Volume 2, Users Guide and input Requirements," W NUREG/CR-4312, EGG-2396, Revision 1, EG&G Idaho, Inc., Idaho Falls, ID 83415, . March 1987. 14.2-3 "XTG-A Two-Group Three Dimensienal Reactor Simulator Utilizing Coarse Mesh Spacing," XN-CC-28(A), Revision 3, Exxon Nuclear Company, Richland, WA (ij#[2:)a -l ' ^ 99352, October 1978. 14.2-4 Technical Specifications for Millstone Unit 2, Docket No. 50-336. 14.2-5 " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors - Analysis if*l31 of Chapter 15 Events," ANF-84-73(P)(A), Rev. 5, Advanced Nuclear Fuels Corp., October 1990. 14.2-6 TRAX-ML: A puter Code for ysis of Slow Tr sients in PWRs " XN-NF-85-24(, xxon Nuclear C any, Richland, 99352, Sept er / 1986. 14.2- "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," ANF-89-151(P)(A), Aiceracd Nucleei Fueis Corp., AprH-it 831 4s w $/ cons 0we< Cys h v, ( *[ a SiehluJ, w 993S2,irr igy y /y,2 % "/noIlsMc t.L,+,ts, :) dosy af Nomai fer./wate, F/uw Tras,,<~ p wi+f fraluac.J AniI,a,y,c,, /_ /,, f,(, ii $/11 f ~ 9 2 - 0/5, fc v, . &mrvi (), u dopo +i,; k /99 V, Dec m I<< -) 34s2.up: 14.2-11 September 1998 i

i-H i e U.S. Nucl:ar Regulatory Commission B17519\\ Attachment 5\\Page~1 FSAR Markups for Loss of Normal Feedwater Event (FSAR 14.2.7) p inserts to Text: A. The'near term. pressurization and DNB aspects of this event are bounded by those of Events 14.2.1 and 14.3.1 respectively. L B in Event 14.3.1, the RCPs are tripped as the initiating event. Reactor trip occurs on low coolant flow, and the core flow rate at the time of trip is significantly lower i than in the Loss of Normal Feedwater Flow event where the RCPs are tripped coincident with the reactor trip. The core power to flow ratio is much higher for Event 14.3.1, thereby producing a more limiting minimum DNBR. lC Each case was analyzed using 102% power, maximum allowed positive reactivity feedback, and maximum permitted pressurizer level. The full power initial condition maximizes the core decay heat that must be removed in the post-scram period. A primary concern in simulating this event is to demonstrate -adequate long-term cooling capability. The single active failure assumptions reduce heat removal capacity by significantly limiting the amount of AFW flow i-supplied to the steam generators. j D Two single failures considered in the Loss of Normal Feedwater Flow event are the failure of a motor driven AFW pump to start and the failure of the steam driven AFW pump to start. Also considered is a loss of offsite power coincident with reactor trip (Reference 14.2-7). .E' The Millstone Unit-2 AFW system consists of two independent motor-driven pumps which are assumed to start automatically within 240 seconds of AFW system actuation on low-low SG level. There is also a steam driven AFW pump which may be started by operator action. (The operator action is credited 10 minutes following reactor trip in the safety analysis). The piping configuration allows each pump to supply both steam generators simultaneously. There are two potential single active failures in this configuration: One is the failure of the steam-driven AFW pump to start, and the other is the failure of one of the two motor-driven AFW pumps to start. Because of the differences in pump capacity and actuation times, it is not immediately obvious which single active failure is the most limiting.

l U.S. Nuclur R guintory Commission B17519\\ Attachment 5\\Page 2 i l Another unceitainty when analyzing the Loss of Normal Feedwater Flow event is the effect of the RCP trip. If the RCPs remain on, the pump heat imposes a .significant ' heat load on the system. If the RCPs are tripped, primary to secondary heat removal capability is degraded due to sole reliance on natural j circulation. The loss of offsite [ ver option (RCP trip), combined with the two single active l failure possibilities produces a total of four base cases. The four cases collectively demonstrate compliance with both the pressurizer overfill criterion j and the steam generator secondary water inventory criterion when the MSSVs i are the sole secondary steam release path. The biases and initial conditions for the cases are identical and are selected to maximize pressurizer level increase and to minimize steam generator level recovery. A fifth case considers the i effects of steam dump system operation and determines whether the SG l inventory boiloff required to cool the RCS to no-load temperature is offset by increased AFW flow at lower SG pressures. { The initiating event for each case is an instantaneous loss of main feedwater. 4 F In accordance with Siemens Power Corporation's (SPC) methodology, symmetric tube plugging is modeled for the four base cases utilizing the MSSVs as the sole secondary steam release path. No SG tube plugging is applied for the fifth case. This conservatively minimizes post-trip SG liquid inventories by producing slightly higher SG pressures and consequently lower AFW flows when SG pressure is controlled by the steam dumps instead of by the MSSVs. G The Loss of Normal Feedwater Flow event is initiated from 102% power with each steam generator at nominal liquid levels. A total instantaneous loss of all Main Feedwater flow initiates the event. When loss of offsite power assumptions are applied, the loss of offsite power and reactor coolant pump trip is assumed to occur at scram. The reactor trips on the steam generator low water level signal. The turbine trips one second after the reactor trips. Cases one through four were evaluated assuming an analytical low SG level reactor trip setpoint of 34%. Case five was evaluated with an analytical low level trip setpoint of 43%. H The cases where offsite power is available and the RCPs maintain forced coolant flow through the primary system produce the lowest steam generator inventories. The loss of one of the two motor-driven AFW pumps combined with post-trip RCS temperature control using the steam dumps resulted in a minimum steam generator liquid mass inventory of 5,540 lbm per SG at 654 seconds. The steam generator level steadily recovers from this minimum level, thus ensuring continued heat removal. An event summary is presented in Table 14.2.7-3. The transient responses are presented in Figures 14.2.7-1 through 14.2.7-5. The transient execution time was 1800 seconds.

1 U.S. Nucl:ar R:gul tory Commission B17519%ttachment 5\\Page 3 l The cases where offsite power is assumed to be lost coincident with the reactor i trip, and primary to secondary heat transfer is achieved via natural circulation, generated the highest pressurizer levels. The loss of one of the two motor-driven AFW pumps produced the maximum pressurizer level of 76.3% at 43 seconds. Sufficient steam volume remains to preclude the expulsion of liquid 4 from the pressurizer safety valves. An event summary is presented in Table 14.2.7-4. The transient responses are presented in Figures 14.2.7-6 through 4 14.2.7-10. The transient execution time was 2400 seconds. 4 k i j

1 l I 4 E 1 s 20. c a ma 18. ------,.------,.------- r.------ ~ C 3 M2 4, m-e D 18 ' ------J.-------8-------L------J.-------8-------.' 1 .i > i4 ,3 o 13 - ------J..-----J.----...L.-----J.----J.-------:- g p 30 - - - - - 4. - - - _ - - -; - - - -. - 8,. - - - _ _ - - ; -.... - - - -;. - - - 3 i 18 - ,1 J.------ J.------- L------ J.----- e 8 ' ~ l U O i i e O 4 ' ------2.-..,T_,--..--L------J.------J.-------- 2- . --.-.-3...---.5.-.----.p...---9..----q---.----. e I ..1 0 0 - 0 400 800 1200 1600 2000 2400 Time (sec) i a 1 i Figure 14.2.7-10 SG Collapsed Liquid Level for Maximum Pressurizer Level Case: Loss of Offsite Power, One Motor-Driven AFW Pump Falls to Start i 1 i I

MNPS-2 FSAR 14.6,5.2.5 Definition of Events Analyzed The purpose of the SBLOCA analysis is to demonstrate that the criteria stated in 10 CFR 50.46(b) are met. The criteria are: (1) The calculated peak fuel element cladding temperature does not exceed the 2200 Flimit. (2) The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the core. (3) The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching. (4) The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the Core. 14.6.5.2.5.1 Description of Small Break Loss of Coolant Accident Transient The SBLOCA is generally defined as a break in the PWR pressure boundary which has an area of 0.5 f t or less (approximately 10% of cold leg pipe area). This range of break areas encompasses smalllines which penetrate the primary pressure boundary. Small breaks could involve pressurizer relief and safety valves, charging and letdown lines, drain lines, and instrumentation lines. The limiting break size is generally in the neighborhood of 2% of the cold leg pipe area. The most limiting break location is in the cold leg pipe at the discharge side of the pumps, particularly with primary pumps tripped on the SIAS. This break location results in the largest amount of inventory loss and the largest fraction of ECCS fluid ejected out the break. This produces the greatest degree of core uncovery and th opgest fuel rod heatup time. /8M The SBLOCA transient is charactenze by a slow depressurization of the primary system with a reactor trip occurring at a low primary pressure of 1750 psia (conserva-tively bounding the actual value of psia)in the Millstone Unit 2 plant. The SIAS occurs when the system has depressurized to 1600 psia. The capacity and shutoff head of the HPSI pumps are important parameters in the SBLOCA transient. The single failure criterion is satisfied by the loss of one diesel generator. In the Millstone Unit 2 SBLOCA analysis, one additional HPSI pump is assumed to be out of service for maintenance, so that only one HPSI pump is available. HPSIinjection is delayed for 30 seconds after reactor scram to model a possible loss-of offsite power at reactor scram. This 30 second delay in starting the diesel generator and HPSI pump does not affect the limiting 0.1 fta break or smaller break size cases because the delay time is satisfied before the primary system pressure has d the shutoff head of the HPSI pumps. For break sizes larger than 0.1 ftz, HPSl flow is delayed only a few seconds due to the startup delay time. Also, the HPSI flow rate l ch. -R becomes less significant as the break size increases above the limiting break size. Therefore, the assumption of loss-of-offsite power has a very srnali effect on the SBLOCA analysis. 14 S6MP2 14.g-l O June 1996

= MNPS-2 FSAR TABLE 14.0.7-1 TRIP SETPOINTS i i Parameter Setooint Uncertainty Low steam generator pressure 680 psia 22 psi Low steam generator water level qq' 5 /0,0 -96% -O-7 in. ' Variable high power 9.6% of rated 5% above current power (< 107% of rated) 5/90 Low reactor coolant flow 91.7 % 2% 4 High pressurizer pressure 2400 psia 22 psi J i \\ l 1 l ) I'so7-1.Hr2 1 of 1 October 1994

MNPS-2 FSAR O b/f#9#6# FC& %AtfdtGn frZon TABLE 14.2.7-1 ()/1/ C 7 AVAILABLE REACTOR PROTECTION FOR THE LOSS OF NORMAL FEEDWAT Reactor Operatina Conditions __ s/1o Reactor Protection 1 Low Stearn Generator Water Level Trip High Pressurizer Pressure Trip Thermal Margin / Low Pressure Trip Variable Overpower Trip 2 High Pressurizer Pressure Trip Variable Overpower Trip Low Steam Generator Water Level Trip 3 Variable Overpower Trip ) 4-6 No Analysis Required; Not a Credible Event h 14S371.MP2 1 Of 1 October 1994l

MNPS-2 FSAR 80 @fN6( ho A Zbfo/N1ryn r TABLE 14.2,7-2 04Lp DISPOSITION OF EVENTS FOR THE LOSS OF NORMAL FEEDWATER FLOW EV l Slof j Reactor Operatina Conditions Disposition 2 1 Analyze to assess maximum pressurizer level swell and long-term adequacy of AFW. 3 i Pressurization and DNB aspects bounded by Event 14.2.1. 3 2, 3 . Bounded by the above, no analysis required. 4-6 No Analysis Required; Not a Credible Event i i ) \\ 1 i i 14327 2.MP2' 1 Of 1 October 1994 }

1 ~-,. ~ - MNPS-2 FSAR lucr TABLE 14.2.7-3 .I EVENT

SUMMARY

FOR THE LOSS OF NORMAL FEEDWATE EVENT 1 i Maxi m Minimum 4 Pre urizer Steam Generator Event velCase Level Case Reactor trips on low steam generator level AFW 0.0 sec. 0.0 sec. C sequence initiated Primary coolant pump trip 0.0 sec. N/A Turbine stop valve closed 0.2 sec. 0.2 sec. Cessation of main feedwater flow 1.0 sec. 1.0 sec. Pressurizer heaters actuate 2.0 sec. 10.0 sec. Steam generator safety va es open ' 12.0 sec. 10.0 sec. Maximum pressurizer f uid level reached at 1301 ft* 180 sec. AFW available steam generators 600 sec. 600 sec. Minimum st m generator reached at N/A 6.5 ft. ..) 1000 sec. J j l 4 i y N Tosen /% w lib /j /9', 3, ) c/ i J ) f 4 h ' ll 1 - 14sn.s.ue2 1 of 1' October 1994 \\ t

A Table 14.2.7-3 1 Sequence of Events for Minimum Steam Generator inventory Case: One Motor-Driven AFW Pump Falls to Start, with Offsite Power and Steam Dumps 1 Time (sec.) Event O Totalloss of Main Feedwater 20 Pressurizer Spray actuates 27.9 Reactor trip signal on low SG water level 28.8 Control rods beDin to drop 4 29.9 Main Turbine trip 32 Maximum pressurizer level,73% 38 Peak Steam Generator Pres _sure (993 pala) 48 AFW Actuation signal on low-low SG water level 53 Charging flow initiated in response to pressurizer level program 58 SG Blowdown isolated 288 Train "A" motor-driven AFW pump starts 4 628 Steam-driven AFW pump starts 654 Minimum SG liquid inventory occurs 1800 End of calculation t j i I ^

( Table 14.2,7-4 Sequence of Events for Maximum Pressurizer Level Case: Loss of Offsite Power, One Motor-Driven AFW Pump Fails to Start Time (secd Event O Totalloss of Main Feedwater 34.1 Reactor trip signal on low SG water level 35.1 Control rodc begin to drop; RCPs tripped 36.1 Main Turbine trip 39 Pressurizer PORV cycles open/clossd 41-AFW actuation signal on low-low SG water level signal 43 Maximum pressurizer level,73% 46 Peak Steam Generator pressure (1055 psia) 51 SG Blowdown isolated 73 Charging flow on (pressurizer level below program) 170 Charging flow off (pressurizer level at program setpoint) 281 Train "A" motor-driven AFW pump starts 635 - Steam-driven AFW pump starts 730 Minimum SG liquid inventory occurs 757 Maximum post-trip RCS average temperature (571 *F) l 2400 End of calculation u 4 f

c /OO (NdN6f

  1. MM"

MNPS 2 FSAR l TABLE 14.6.5.2-3 MILLSTONE UNIT 2 SMALL BREAK LOSS OF COOLANT ACCIDENT SYSTEM ANALYSIS PARAMETERS 4 i j Primary Heat Output, MWt 2,700' l . Primary Coolant Flow Rate, Ibm /hr 1.36 x 10' (360,000 gpm) Primary Coolant System Volume, ft* 10,506'* l Operating Pressure, psia 2,250 inir.t Coolant Temperature, 'F 549 i ] Re:ctor Vessel Volume, ft* 4,534 Pressurizer Total Volume, it' 1,500 a- .l Pressurizer Liquid Total,~ft* - 922 . SIT Total Volume, ft* (one of four) 2,019 i. i [ SlT Liquid Volume, ft* 1,150.5 SIT Pressure, psia 215.0 { SIT Fluid Temperature, 'F - 106.8 Total Number of Tubes per Steam Generator 8,523 i 4 ~ Number of Tubes Plugged per Steam Generator 500(5.87 %) i. l p Staam Generator Secondary Flow Rate, Ibm /hr 6.02 x 108 Stsam Generator. Secondary Pressure, psia 828.8 Steam Generator Feedwater Temperature, 'F 435 R1 actor Coolant Purnp Rated Head, ft 271.8 y 1 ..LRiactor Coolant Pump Rated Torque ft lbf 31,560 l R: actor Coolant Pump Rated Speed, rpm 892-o 4 'Primari/ West output used in SPC-RELAP model - 1.02 x 2700 - 2754 MWt. g,,g y

  • Includes pressurizer total volume and 5.87% average SGTP.

p [\\f.seg2,.n4P2

l Of 2

- June 1996

MNPS-2 FSAR TABLE 14.6.5.2-3 MILLSTONE UNIT 2 SMALL BREAK LOSS OF ' COOLANT ACCIDENT SYSTEM ANALYSIS PARAMETERS Initial Reactor Coolant Pump Speed, rpm 8 49.9 ' ' ' RIsctor Coolant Pump Moment of inertia, Ibm-ft' 100,000 SIS Fluid Temperature, 'F 100.0 Rractor Scram Low Pressure Setpoint, psia + 1,750 SIAS Activation Setpoint Pressure, psia 1,600 Srcondary Safety Valve Setpoint, psia (Nominal) (Up to 3% Un' certainty in Lift i Pressure was Considered in the Analysis) 1,000 HPSI Delay Time, sec. 30.0 9k -lT Gl% HPSI Flow Rate vs. RCS Backpressure HPSI Flow RCS Pressure (One Pump) (osia) (aom) 1225 0 1200 79 1100 198 800 371 500 475 209 564 200 567 169 576 l %-li 160 579 4 120 592 80 005 40 619 14.7 628 %-11 4 !$(c5 +

  • * *Value used in SPC-RELAP for initialization.

96-19 .+The assumed value of 1750 psia conservatively bounds the actual value of tB50 psia. . 1?O652 3.MP2 ' ' 2 Of 2 - June 1996

V v MNPS-2 FSAR MAY, 1990 t l sys- = l 0,0-4 .s - N. : lhe x 1. s I. l N. u l I l \\ i 500-- --l-r- - - ,I o 8 9 g i .f - ggg-q. 3: c% V i 9 l D l sao l l l 2 l 0 500 1000 1500 2006 0500 3000 3500 4000 4500 5000 11tseibec I F!GURE 14.2.7-1 CORE AVERAGE TEMPERATURE FOR LOSS OF NORMAL FEEDWATER FLOW (STEAM GENERATOR liWENTORY) l l t

Y v v g g C D RC3 Hot Leg Temp l l l C 3 RCS Cold Img Temp 62s - - - - - - - j - - - - - - - w,- - - - - - - p - M RCS Avercge Temp b. j-------g--------i.--------e.-------,. g 600 = f ses 1. - - - - - -.; - - - - - - - -;- - - - - - - - ;. - - - - - - ; - - - - - -.4.- - - - - - - _. 3 g o q eso /

- - - - - - - -l- - - - - - - - ;. - - - - - - - 4. - - - - - - - -l- - - - - - - -

m o t-m_. - -{- - - - - - i - - - - - - - -l- - - - - - - - r - - - - - - - i - - - - - - - -l- - - - - - - - y +- w ses i. i i 1 .i i 300 i u-0 400 800 1200 1600 2000 2400 Time (sec) Figure 14.2.7-1 RCS Loop Temperatures for Minimum Steam Generator inventory case: Offsite Power Avnllable, "B" Motor-Driven AFW Pump Falls to Start

i .t 4 k t{

i!

$-(( l' ~ D d S $ m4 ( 099 1 Y 5 A 0 M ,2 00

6 4

L AM AIIl .i 0 R I I !n 0 O I. ,0 N 4 F )Y O R O 0 S T 0 SN = ol i e -\\. 3 LV

6 OE N

RI O 0 FR 8,' R m0 O 0 A 3 ET S o RA UR F e SE 0n SN v EE 6., RG 2 8 P 2m M S = n RE YA P 0' AT N g, c. 0 DS M jj .i Il ,0 N( 2 O W C EO u SL 1

i 0

F i7.' N ll;0 5 R

Np T

1 E 2A W 0 i.D ue+ N i : 0 E '{ !I' !I; 0

2. E N

1 4F 1 N-= E N-bl 0 R ,0 U 5 G s IF N = I 0 e o g 0 o e o 0 a s g k 9 . s $m e dd ,y . ~ g t t .ili l, fll !lill

m r-e - e l l l l 1 0 4 4 4 1160 C C SG 1


3,-------i,-------r------1,---

C 3 SG,, 1 0 t 4 s 1100 -------J-------8------L------J-------8-------' s e e s 4 4 0 0 t t t 9 I i Q 1060 -------i-------i-------r------1-------i-------- 4 I e s 4 J 0 0 I e 6 3000 .'.......a.......s........L.......a........s.......J 0 0 0 0 1 0 4 I e I e 4 4-5 950 3-------i------- r------ t-------S------- e e e e e t i I 4 8 g goo 4........s........L.......a...............J I e 4 8 9 1 I 4 _M I e 1 850 i e r 3-------i-------- e e e e i i e a e s e gm.......4.......s........L.......a.......s.......J l 0 0 4 0 8. 1 l 750


1.-------i.-------r.------1.-------i.--------

700 O 400 000 1200 1600 2000 2400 Time (sec) Figure 14.2.7-2 Steam Generator Dome Pressure for Minimum Steam Generator 4 inventory Case: Offsite Power Available, "B" Motor-Driven AFW Pump Falls to Start 1 I

U. w I MAY,1990 ~ MNPS-2 FSAR 1 11 YWPR 3,,, =. I i .g i I r-I t i.- .=. . h, e g-f N i. l l t W aa I +- g_ - I i I i t y r . f;

f. -

A- -.m e. ,. _.=- \\ I i i .8 nO l i T.N j i I [ .t N 3 l 400-l 0 600 1000 1600 2008 2000 8000 3500 4000 4600 D / 1 '!!me, seo / [ / s\\ f t FIGURE 14.2.7-3 PRESSURIZER LIQUD VOLUME FOR LOSS OF NORMAL FEEDWATER FLOW (STEAM GENERATOR INVENTORY) i 4

f 200 i. ,i i 'g g g l l na ________>_______{.___-_-_',______-- m{ 75 -____-__4._______ r g g g i. .5 s0 ... _ _ _ _ _ _ ; _ _ _ _ _ _ _ _l_ _ _ _ _ _ _ _ ;. _ _ _ _ _ _ _ j. _ _ _ _ _ _; _ _ _ _ _ _ _ y r -c ; ' s---o l u c l g23 _______{________;________;.______________q._______ u A l l l l l -lC 3 CNTRLVAR_620l 0 0 400 800 1200 1600 2000 2400 Time (sec) I Figure 14.2.7-3 Pressurizer Level for Minimum Steam Generator inventory Case: Offsite Power Available, "B" Motor Driven AFW Pump Fails to Start r 1 x

t l i

f t'

f ! i - -l >b! qse3" 9 \\O 0 9 1 \\ [> Y N 0 A ,0 M 84 F O 0 ' /- ,0 S) 0 SY 4 OR LO N-0 OE T RN ,->8 ';0 FV 5 N -l[ 3 I L ER V O N ET 00 LA JF \\

0 R

R 3 OE e UN A e OE G S 0s L = 0 w EM 0 o / 2h OA E S 2 T Y (S 0 R S 0 AW P L-3 D w O N 1 N L M OF C ER 0 SE u !,. 0 T -N' 6 A 1 W D N.. 4 E 0 - E

7. F N-

,,t-0 I ,I1* .[ 0

2. A L

\\ 1 4 M ~ 1 R EO 0 RN i-j1 ,l 0 U K I 0 G F l 0 M s. 0 s. 5 u e u 8 s v 1 s a i $ I j s.e 8 $ E o $ O o s G d ,:i flifl1 lijl l lilll l! l l i

. ~. a

  • y 160000 e

e e e e i e e C 3 SG 1 135000 ------+ i------- ------- r------ T--- C 3 SG 2 e e e 4 I 1 6 a 120000 ------- J------ J------. L--.--- J------ J------ J 4 4 4 0 ^ 0 4 0 8 0 0 105000 ,1------..e e I r------t-------i-------- p 0 4 0 0 4 V 0 I e 1 0 30000........J.......J........L.......J.......J.......J 4 I i e i I I I l i 2 76000 - ------ 3------ ,4 e 0 0 0 r------ 1-------i-------- t 8 I e t T I e 0 0 0 3 00000 - ------ d------ J------- L--~~~~ J-------8------ J y e a e e a t i 43000........,......._,.i.........I,........,. ..4 .3 e 8 4 0 0 4 4 0 00000 ' ....J.........s........L.. .4.........s........J 15000....... i. 0 O 400 000 1800 1000 2000 2400 Time (sec) Figure 14.2.7-4 SG Liquid Mass inventory for Minimum Steam Generator Inventory Case: Offsite Power Available, "B" Motor-Driven AFW Pump Falls to Start m.

I m i r 1 MNPS-2 FSAR MAY, 1990 t L i (,- l ses i i. I j i l 1 .... j.. l 4. TAVEC _ } t j i } i z y _p. - t { i 4 ........p... . ;l -. _4 i i .i \\: 3 L gyg., +:. c. ....g.....*..)- u .q. r=, 8 I i i 1 I ^ -- l - - y- --- - ----!


-!--- -- - ---+$ --- ----- ? -- -- ---- ! --

0 37s.s- = - - >4 t [ e E* I t. UFO- --t.---*-----*e.******--*:-----****---I' t---------*t-------t-i 689.8 -


4-**-**-*------------3-----**-*~'-**---

' - - - ~ ^ ti i i i t SSS- '--~~~t--~~~~-~t"~~~~~~t"~~~~--~~-~~~t-~~~~- -t -~~~t-- i i i i N - \\. ggy. 9.. ......p ..........p. 8-- j .p..: _ :p - - j j 4 i i g i s00-x a i i 0 50 100 150 200 250 300 350 M )' %n Thne, see t i 6 FIGURE 14.2.7-5 CORE AVERAGE TEMPERATURE FOR LOSS OF NORMAL FEEDWATER FLOW (PRESSURIZER INVENTORY)

e e i 1 1 1 1 1 i i 1 C 3 SG 1 18 ---------v------9-------r------T--- C 3 SG 2 I i 1 0 m 1 I I 1 a

  • J O 16 '-------2-------8-------L------2-------8------J 1

1 1 I i ) I I I I i mU l i I i 1 > 14 ------- 7------ 9-~~---- r------ T------ 9- ] a i i l i 1 1 I i i t 12


2-------8-------L------2---

8------ 2 8 I I I I l .-g i 0 i I 1 1 I I I I .7 10 3-------i------- r-------


i--------

M I i 1* 6 0 1 I I I i J m 8 -. s.......J.------.u. ..s.----- J.--...-- U i 1 a l 1 M I I e i 1 b 1 I i 1 0 c 6 1------- v--------W------- UU 1 I l i _Q l i I i U 4 ' ......J-. L....--.s-...--.J-.....- l \\ l P 1 1 O 0 1 1 1 J M e...-- ..,t 6,.....- ,1------ ,1.......- i I i 1 i 0 1 4 1 I. 0 o 400 800 1200 1600 2000 2400 Time (sec) 1 Figure 14.2.7-5 FG Collapsed Uquid Level for Minimum Steam Generator inventory Case: Offsite Power Available, "B" Motor-Driven AFW Pump Falls to Start l l J 9 5 9

V v MNPS-2 FSAR my, sgo t i i too ~ .I PSCSA1 \\- x, ..i., :. ....... j............. j........... i..........i..... p..._ i 3ese. l l + i 8 I i 4..............:....... goes. d g g i i m r K g t i 1..:! i e m. U a i i l n g e j:. ....[. g g. I O. j I- - -- ------ b -------- *!- =-- ( M 860-i i l l 1 s- -- - ' 8 g t

N i

i 800- +-----------4---------*-------:---+-------*:-N --e- ,y i w,' 1 N0 N s P 4 s s a a 0 0~ 60 100 150 200 250 SC3 360 00 / m Time, see / d FIGURE 14.2.7-6 SECONDARY PRESSURE FOR LOSS OF NORMAL FEEDWATER FLOW (PRESSURIZER INVENTORY)

es0 -..............a................a........s._...__.- ses e i ..........,.......4........q......._- g 800 S i i ---*-------*-- h ,, 575 r .c l

N.

e b-l d... ;........ h I l 330 n r M i-525 - - - - - - - - j - - - - - - - -l- - - - - - - - l - C 3 Pm Hot Leg Temp l l l C 3 RCS Cold Leg Temp A----a RCS Average Temp. I s 500 0 400 800 1200 1600 2000 2400 Time (sec) Figure 14.2.7-6 RCS Loop Temperatures for Maximum Pressurizer Level Case: j,, Loss of Offsite Power. One Motor-Driven AFW Pump Falls to Start 1 i k I l 1

~ O v MNPS-2 FSAR s i i j .i i. ~ i. i VTPR I goes. -. 8.-..s I: i j g i c3 laso- ------*:------


3

- --- - 4 ~ - -- i ---- -- ---- l-i 3 .o j .i ~d 3 oo., 4.... ..j. y_ O' I i l 5, 9 g T* liso-i i i 1 --~e = - N ---~-~4~---~+------*--------e- ~~4-~= 1 l g I a "w 1100- ---------------~--------~i----------i----------4 - p ---- i p, .s 1950- ----+~~----+~~---:----.* -- --~ 4 ~~- ~ t- ~ -- a~~~-~.- I 1000 1 3 O s0 100 150 200 250 300 350 N N 'R me,800 R j n i FIGURE 14.2.7-7 PRESSURIZER LIQUD VOLUME FOR LOSS OF NORMAL FEEDWATER FLOW (PRESSURIZER LWENTORY)

f 1200 i e e s 1860 i i e C 3 SG 1


3-------i-------r-------1---

C 3 SG 2 i e a 0 4 4 a

00

4-------.,0


c-----2------s-------:

I I e e t i e t i I 8 0 6 6 6 i g 1050 - ------i-------g-------r------1-------i-------- v 1000


8--------

J-------8-------- 0 0 0 0 8 g 4 4 I i i L 8 0 0 8 1 950 3-------i------- r------ 1-------i-------- 8 8 4 a e I i 0 6 0 g, 900 J-------8------- L------ J-------8--------' 8 I I I I g I 4 I 1 B I e a 850 ~------ 3------ ,e------- r------ 1-------i-------- f .I e I e i 800


J-------J.------L------J-------8-------'

s e. e. e I ,s0 -------,.-------,.------,-------,. 700 i i e i i O 400 830 1200 1800 2000 2400 Time (sec) Figure 14.2.7-7 Steam Generator Dome Pressure for Maximum Pressurizer Level Case: Loss of Offsite Power, One Motor-Driven AFW Pt.mp Falls to Start d

D. MAY,1990 MNPS-2 FSAR i i i i I I I DCWA1 N i i i i i i {.. j... 4 +[ r s. S-I i l N i i x u,g M. 2 g< 1 y an_ ..;!..... ;...,N.......;..... q. 3 l 4 I N.:: 2 I ~4 si 4i e 1.N i a ni_ .s i.- .i e t@ 8 l l l N i j T{ l i L .T O j l n.' 30 .g t w u i 8 i l o 0 i i i. i t 1 l ,i... N .i.- 10-.- m- - L J l i i i 3 t i l' ] 0 6O ido ide the 260 She 350 MS Time, see \\ q \\ ,b .N / FIGURE 14.2.7-8 SECONDARY SDE LIQUD LEVEL FOR LQSS OF NORMAL FEEDWATER FLOW (PRESSURIZER NVENTORY)

a e 100 l l a i n i i I e i e 1 I I i 1 8 ^ ~ a C { 75 p------*g --------i,-------p-------j--------l-------- g g s v e i e i i i T 1 I e I l e i e 50 3 ---~~~~-r-------n-------r-------{-------'-------- l 8 g g I I I I L 1 I I I O i 4 c ,i N e ,e l l 2 i 25 -------~}--------l--------l---.----*--------*-------- y i i m l l l \\ i. e i i i i s C

3 CNTRLVAR 620 0

4 I. t 0 400 800 1200 1600 2000 2400 Time (sec) Figure 14.2.7-8 Pressurizer Level for Maximum Pressurizer Level Case: Loss of Offsite Power, One Motor-Driven AFW Pump Falls to Start 4 L m m.

~.- - . r C ,150000 C 3 SG 1 I l l l l C 3 M2 125000 --------.4.-------d.-------.6-----_-4.-------.-------- s

0.,

100000


*--------i----~~--.--------*--------i--------

= n000.._ _ ___

4.._______l...-_-__;._______.;___ _____l__ __ _ _ _ _- g y 50000 - - - - - - - j - - - - - - - -l- - - - - - - - l- - - - - - - - j - - - - - - - -l- - - - - - - - 8 2 _ _ _ _ _ _ _ _l_ - _ - _ _ _ _ t - - l - - _ _ _ _l_ _ _ _ _ _ _ _ 3000 i i i i 0 0 400 600 1200 1600 2000 2400 Time (sec) Figure 14.2.7-9 SG Liquid Mass Inventory for Maximum Pressurizer Level Case: Loss of Offsite Power, One Motor-Driven AFW Pump Falls to Start 4 I a L e 9 - - - -}}