ML20149F097
| ML20149F097 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/18/1997 |
| From: | AFFILIATION NOT ASSIGNED |
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| NUDOCS 9707220037 | |
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{{#Wiki_filter:._ . . - .- - .-. .. l. PARSONS POWER 4 4 Critical Design Characteristics Overcooling Events l Millstone 2 1 e Reviewed By: [L , - , Ace t 1 Date: 7.8,!77
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2 .. . . 1 4 k TABLE OF CONTENTS-4 i , 1.0 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM.. .. .. .. . .. 2 4 2.0 CRITICAL DESIGN CHARACTERISTICS . . . ... .. .... .. .. .. .. 5 2.1 DECREASE IN FEEDWATER TEMPERATURE . . . . . . . ... ... . . . . ,5 2.2 INCREASE IN FEEDWATER FLOW.- . .. . . . . . . . . .. ... ........6 2.3 INCREASE IN STEAM FLOW.. . .. . . . . . . .. .. . . ... . 7 2.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE . . . . . . .. . . 10 4 2.5 MAIN STEAM LINE BREAK (MSLB) ACCIDENT, RCS ANALYSIS., . . . . .. . . . . . . . . .11 4 9 d l 4 4 d i I l i 1 l l l
ATTACHMENT 1 i OVERCOOLING EVENTS CRITICAL DESIGN CHARACTERISTICS 1.0 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
- 1.1 Five design basis events (DBEv's) are considered in section 14.1 of the FSAR. The applicability of each accident for each plant operating mode, and whether an analysis was performed by NNECo, is presented in Figure 1.
Figure 1: Tier-2 AMSR - Overcooling Events 14.1.1 14.1.2 _ 14.1.3 14.1.4 14.1.5 Decrease in increase in increase in Steam Main Steam Plant Feedwater Feedwater Steam flow Generator Line Break Oper. Temperature Flow Relief Valve (MSLB) Mode Ooening 1 14.1.3 14.1.3 Analyze 14.1.3 Analyze 2 14.1.3 Mode 3 Mode 1 14.1.3 Analyze 3 14.1.3 14.1.3 Analyze 14.1.3 Mode 2 4 N/A 14.1.3 Mode 3 14.1.3 Mode 2 5 N/A 14.1.3 Mode 3 N/A Mode 2 6 N/A 14.1.3 Mode 3 N/A Mode 2 Analyze - The DBEv was analyzed, by NNECo, for the listed operating mode and the results summarized in FSAR section 14.1. Mode "n"- The DBEv is bounded by the mode "n" case ("n" = 1 - 6). 14.1.3 - The DBEv is bounded by another event, in this case by event 14.1.3. N/A - The DBEv is not applicable for the listed operating mode. 1.2 The AMSR Program will " review" the DBEv's listed as " Analyze"in Figure 1. These are: DBEv 14.1.3 Mode 1 - 10% increase in steam flow at full power (limiting case). DBEv 14.1.3 Mode 3 - 41% increase in steam flow at Hot Zero Power, equivalent to rapid opening of the steam dump and turbine bypass valves. DBEv 14.1.5 Mode 1 - MSLB @ Hot Full Power (HFP) offsite power available, DBEv 14.1.5 Mode 1 - MSLB @ Hot Full Power (HFP) coincident with a loss of offsite power, DBEv 14.1.5 Mode 2 - MSLB @ Hot Zero Power (HZP) offsite power available, DBEv 14.1.5 Mode 2 - MSLB @ Hot Zero Power (HZP) coincident with a loss of offsite power. 2
1.3 Figure 2 shows the systems invohed in the mitigation of each analyzed overcooling event. l Figure 2: Tier-2 AMSR - Systems Involved in Overcooling Events l 14.1.2 14.1.2 14.1.5 14.1.5 14.1.5 14.1.5 Increase in Increase MSLB MSLB MSLB MSLB l Steam in Steam Mode I w/ LOOP Mode 2 w/ LOOP SYSTEM Flow Flow Mode 1 Mode 2 1 Mode 1 Mode 3 RPS X X X X X X CEDM X X X X X X RCS X X X X l TURB X X X X EHC X X X X l MS X X X X X X MFW X X AFW X X X X I CVCS X X X X HPSI X X X X ELECT X X X X X X DIST _ EDO , X X 1.4 Figure 3 shows the Which Critical Safety Functions have corresponding Critical Design Characteristics for each of the analyzed accidents. Figure 3: Tier-2 AMSR - Critical Safety Functions wi th CDCs in Overcooling Events 14.1.2 14.1.2 14.1.5 Increase in Increase MSLB Steam Flow in Steam Mode 1 Critical Safety Function Mode 1 Flow ' (4 Cases) Mode 3 Reactivity X X X Fuel Integrity X X X RCS Heat Removal X X X RCS Pressure & Inventorv * *
- Containment Integrity &
Radiation Control Electrical Power X X X Essential Support Systems Environmental Control
- To be determined from detailed analyses 3
i 1.5 - The analyzed event duration, Minimum Departure from Nucleate Boiling Ratio (MDNBR) and Maximum Linear Heat Generation Rate (Max LHGR), as documented in FSAR chapter 14 for each
- analyzed event, are presented in Figure 4.
. Figure 4: Tier-2 AMSR - Overcooling Events Analysis Results i FSAR SECTION 14.1 ANALYZED DESIGN BASIS Analyzed MDNBR Max : EVENT Duration (sec) LHGR DBEv 14.1.3 Mode 1 - 10% increase in rteam flow at HFP 43 1.21 19.3 DBEv 14.1.3 Mode 3 - 41% increase in steam flow at HZP * *
- DBEv 14.1.5 Mode 1 - MSLB @ HFP offsite power available 600 3.00 17.1 l DBEv 14.1.5 Mode 1 - MSLB @ HFP & loss of offsite power 600 4.6 5.7 4 DBEv 14.1.5 Mode 2 - MSLB @ HZP offsite power available 600 2.4 20.99
- DBEv 14.1.5 Mode 2 - MSLB @ HZP & loss of offsite power 600 1.18 16.5 i-
- Results not available in FSAR section 14.1.3 .
1.6 Critical Design Characteristics Development Method The following method was used to develop the overcooling event CDCs. Five (5) overcooling events are
- described in section 14.1 of the FSAR. Each event was reviewed and design inputs extracted. Each l ' design input was assigned to one or more Critical Safety Functions (CSFs). The CSF diagrams were then used to develop functional / system level CDCs for each event.
1 [; 1.7 System Boundary Diagrams System Boundary Diagrams (SBDs) are developed for each accident mitigation system. Using the SBDs, the AMSR Program Team will identify the system configuration and component actions required . to meet the system / functional CDCs. 'Ihis information will be loaded into the Tier-2 Data Base and will . constitute the Chapter 14 requirement at the component level. I 1.8 CDC Validation - The CDCs will be validated "as present" in the installed plant configuration. The validation method will be determined following review of the detailed analyses supporting the FSAR Charpfer 14 events.
, Millstone-2 system and component test data will be used to the maximum extent possible to perform this 1 j validation. When CDCs cannot be validated by test, then analysis or alternate means will be used to perform the validation.
4
1 2.0 CRITICAL DESIGN CHARACTERISTICS i ) \ A discussion of each FSAR section 14.1 DBEv, and the functional / system CDC listing, are presented in sections 2.1 through 2.5 of this attachment. These CDCs will be augmented with information derived ; from FSAR Chapter 6,7,8 and 9, and the detailed analyses and calculations that support FSAR Chapter 14. 2.1 DECREASE IN FEEDWATER TEMPERATURE 2.1.1 EVENT DESCRIPTION The decrease in feedwater temperature event is discussed in FSAR section 14.1.1. The event can be ! caused by loss of one or more feedwater heaters. This can occur due to a loss of extraction steam or ; due to heater bypass. The worst case loss of feedwater heaters would occur if the low pressure heaters were bypassed. The most limiting case is from rated power. i 2.1.2 DESIGN BASIS l N/A - Bounded by event 14.1.3 2.1.3 SYSTEM INTERFACE N/A - Bounded by event 14.1.3 2.1.4 EVENT DISPOSITION This event is bounded by the Increase in Steam Flow Event discussed in FSAR Section 14.1.3, and thus is not analyzed. 2.1.5 FUM.'.TIONAIJSYSTEM CDCs - EVENT 14.1.1: DECREASE IN FEEDWATER TEMPERATURE This event is bounded by event 14.1.3. Therefore, no functional / system level CDCs are identified. 5
if
- 2.2 INCREASE IN FEEDWATER FLOW d
2.2.1 EVENT DESCRIPTION The increase in feedwater flow event is discussed in FSAR section 14.1.2. The event is initiated by a failure in the feedwater system which causes an increase in feedwater flow to the steam generators. The i limiting consequences of the increase in feedwater flow will occur at rated power conditions and will bound all other power operating conditions due to the initial steam generator inventory and initial margin to DNB. The largest cooldown which can be postulated due to feedwater addition at full power is the inadvertent startup of all three AFW pumps. This cooldown rate is less than that calculated for event 14.1.3, Increase in Steam Flow, and thus is bounded by that event. 2.2.2 DESIGN BASIS N/A - Bounded by event 14.1.3 2.2.3 SYSTEM INTERFACE N/A - Bounded by event 14.1.3 l 2.2.4 EVENT DISPOSITION ~ This event is bounded by the Increase in Steam Flow Event discussed in FSAR Section 14.1.3. 2.2 5 FUNCTIONAUSYSTEM CDCs - EVENT 14.1.1: INCREASE IN FEEDWATER FLOW . This event is bounded by event 14.1.3. Therefore, no functional / system level CDCs are identified. l I
.I I
/ 6
1 2.3 INCREASE IN STEAM FLOW 2.3.1 EVENT DESCRIPTION This event is initiated by a failure or misoperation in the main steam system which results in an increase in steam flow from the steam generators. The simultaneous opening of the condenser dump and turbine bypass valves could result in a steam load increase of 41% above full rated load. The consequences of this event for all other power operating conditions, including Mode 2, are bounded by the rated power operating conditions. The Mode 3 condition bounds modes 4 - 6 for "zero power" initial conditions. This energy removal rate also bounds the rated power operating conditions for events 14.1.1,14.1.2 and 14.1.4. 2.3.2 DESIGN BASIS The MNPS-2 Increase in Steam Flow analysis is based on the following primary assumptions:
- a. Most reactive control rod stuck in its fully withdrawn position.
Reference:
FSAR Section 14.0.6
- b. Single failure criteria for offsite power case is (to be determined)
- c. Single failure criteria for LOOP case does not apply, i 2.3.3 SYSTEM INTERFACE ;
The following systems interface during this event:
- a. Main Steam
- b. Control Element Drive
- c. Reactor Protection System ;
- d. Turbine Generator (Turbine Stop Valves)
- e. Electro-Hydraulic Control System
- f. Electrical Distribution l
2.3.4 EVENT DISPOSITION ! l Two cases are analyzed. For Mode 1, the limiting event is a 10% increase in steam load initiated from full power. This event is mitigated by reactor and turbine trip. For Mode 3, the limiting event is a rapid opening of both the atmospheric dump turbine bypass valves resulting in a steam flow equal to 41% of full rated load. The event sequence is not included in the FSAR. 4 e 7
, . _ .. . = ... . ..
2.3.5 FUNCTIONAL / SYSTEM CDCs - EVENT 14.1.3: INCREASE IN STEAM FLOW (Mode 1) The Critical Design Characteristics for the Increase in Steam Flow Event, Mode 1, are presented below. 2.3.5.1 REACTIVITY CONTROL CSF Functional / System CDCs - Insert control rods.
Reference:
FSAR Table 14.1.3-3 2.3.5.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF Functional / System CDCs - Included with RCSHR CSF. 1 2.3.5.3 RCS HEAT REMOVAL CSF s Functional / System CDCs - Trip turbine to reduce the steaming rate.
Reference:
FSAR Table 14.1.3-3 q 2.3.5.4 RCS PRESSURE & INVENTORY CONTROL Functional / System CDCs - Identify from event detailed analyses, if applicable 2.3.5.5 CONTAINMENT INTEGRFlY & RADIATION CONTROL CSF Functional / System CDCs - N/A for this event. 2.3.5.6 ELECTRICAL POWER CSF Functional / System CDCs - Transfer loads to offsite power source. , 2.3.5.7 ESSENTIAL SUPPORT SYSTEMS CSF ; Functional / System CDCs - N/A for this event. m 2.3.5.8 ENVIRONMENTAL CONTROL CSF Functional / System CDCs - N/A for this event t I 8
2.3.6 FUNCTIONAIJSYSTEM CDCs - EVENT 14.1.3: INCREASE IN STEAM FLOW (Mode 3) The Critical Design Characteristics for the Increase in Steam Flow Event, Mode 3, are presented below. 2.3.6.1 REACTIVITY CONTROL CSF
- Functional / System CDCs -Identify from event detailed analyses 2.3.6.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF Functional / System CDCs -Included with RCSHR CSF.
- 2.3.6.3 BCS HEAT REMOVAL CSF Functional / System CDC - Steam flow limited to 41%
Reference:
FSAR Section 14.1.3.6 4 2.3.6.4 RCS PRESSURE & INVENTORY CONTROL j Functional / System CDCs -Identify from event detailed analyses, if applicable 2.3.6.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF Functional / System CDCs - N/A for this event. 4 2.3.6.6 ELECTRICAL POWER CSF Functional / System CDCs - Plant loads power from offsite power source which is maintained through out event. No additional functional / system CDCs. 2.3.6.7 ESSENTIAL SUPPORT SYSTEMS CSF Functional / System CDCs - N/A for this event. 2.3.6.8 ENVIRONMENTAL CONTROL CSF Functional / System CDCa - N/A for this event 9
2.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 2.4.1 EVENT DESCRIPTION This event is initiated by an increase in steam flow caused by the inadvertent opening of a secondary side safety or relief valve. The inadvertent opening of a steam generator safety valve would result in an : increased steam flow of approximately 6.75% of full rated steam flow. This event is bounded by event 14.1.3, which analyzes steam flow increases of 10% and greater. See discussion in section 1.3 above. l 2.4.2. DESIGN BASIS N/A - Bounded by event 14.1.3 (Relief valve capacity <10% full rated flow) 2.4.3 SYSTEM INTERFACE 4 N/A - Bounded by event 14.1.3 4 2.4.4 EVENT DISPOSITION This event is bounded by the Increase in Steam Flow Event discussed in FSAR Section 14.1.3, and thus
, is not analyzed. However, a functional / system level critical characteristic is appropriate. That is, to
{. validate that the steam generator relief valve capacity is less than the 10% assumed in the event 14.1.3 i analysis, t
)
2.4.5 FUNCTIONAUSYSTEM CDCs - EVENT 14.1.4: INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 1 i . The Critical Design Characteristics for the Inadvertent Opening of a Steam Generator Relief or Safety , Valve Event are presented below. l 2.4.5.1 RCS HEAT REMOVAL CSF Functional / System CDCs - Steam generator relief valve capacity is limited to 10% of full rated steam l1 flow.
Reference:
FSAR sections 14.1.4.4 and 14.1.3.6. 1
. 2.4.5.2 Functional / System CDCs do not apply for the other 7 critical safety functions.
1-a i j 10 l l
- - - .~. - - . -.- -- .- -
i U 2.5 MAIN STEAM LINE BREAK (MSLB) ACCIDENT, RCS ANALYSIS 2.5.1 EVENT DESCRIPTION This event is initiated by a rupture in the main steam piping upstream of the MSIVs which results in an ; 4 uncontrolled steam release from the secondary system. He increase in energy removal through the t 4 secondary system results in severe overcooling of the primary system. In the presence of a negative ' Moderator Temperature Coefficient (MTC), this cooldown causes a decrease in the shutdown margin (following reactor trip) such that a return to power might be possible following a steam line rupture
! . assuming that the most reactive control rod is stuck in its fully withdrawn position. The MNPS-2 limiting MSLB from a safety standpoint is a Hot Zero Power (HZP) double-ended guillotine break inside containment between the steam generator and the flow restrictors. He MNPS-2 MSLB-RCS analysis is described in FSAR Section 14.1.5. l l
2.5.2 DESIGN BASIS I i The MNPS-2 MSLB-RCS ir ' -
. the following primary assumptions:
- a. Most reactive cont' ' - - Adrawn position.
1
Reference:
FSAR w
- b. Single failure crbria for t . loss of one HPSI pump and one charging pump.
l Refererice: FSAR Section 1,. 3, and Table 14.1.5-3.
- c. Single failure criteria for LOOP ( '" one diesel generator, resulting in the loss of one
' HPSI pump and one charging pump.
Reference:
FSAR Section 14.1.5.4,14.1.5.5.1.3, and Table 14.1.5-3 . d. Safety injection actuation signal (SIAS) actuated by low pressurizer pressure.
- e. Secondary isolation signal (MSI) actuated by low steam pressure.
(
- 2.5.3 SYSTEM INTERFACE I
.ll The following systems interface during the postulated MSLB-RCS recovery analysis:
4
- a. Reactor Protection System i
- b. Control Element Drive
- c. Reactor Coolant System !
- d. Turbine I 4
- e. Electro-Hydraulic Control
- f. Main Steam l
- g. ~ Main Feedwater System (hot full power case) l h. AuxillaryFeedwater
- i. Chemical & Volume Control System
- j. ' High Pressure Safety injection 4 k. ElectricalDistribution
- l. Emergency Power System (loss of offsite power case)
~
'2.5.4 EVENT DISPOSITION , The limiting break is a double-ended guillotine break inside containment between SG and flow restrictors. Two modes are analyzed. For mode 2, the limiting case is with the reactorjust critical.
- 11 l
}
The mode 2 analysis is performed wi h offsite power available and also concurrent with a loss of offsite power, For mode 1, the lim.Y . cas :is with the reactor at full rated power. The mode 1 analysis is performed with offsite power avanable and also concurrent with a loss of offsite power. i 1 12
2.5.5 FUNCTIONAL / SYSTEM CDCs - EVENT 14.1.5.6.1: MSLB @ HZP The Critical Design Characteristics for the MSLB Event @ HZP, with offsite power available, are presented below. 2.5.5.1 REACTIVITY CONTROL CSF Functional / System CDC -Insert control rods within 3.9 seconds of reaching reactor trip setpoint. Refersce: FSAR Table 14.1.5-4 Functional / System CDC - Inject boron to limit peak power.
Reference:
FSAR Table 14.1.5-7 Note: HPSI and charging pump actuation times included under RCS Pressure & Inventory CSF. Functional / System CDC - Inject boron to achieve core subcritical ec,ndition beyond 600 seconds.
Reference:
FSAR Table 14.1.5-7 2.5.5.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF - FunctionaVSystem CDCs included in 2.5.5.3 and 2.5.5.4 below. 2.5.5.3 RCS HEAT REMOVAL CSF Functional / System CDC - Close MSIV on intact main steam line to limit cooldown from non-affected steam generator blowdown.
Reference:
FSAR Table 14.1.5-7 Functioaal/ System CDC - Limit AFW flow to affected steam generator for first 180 seconds of event.
Reference:
FSAR Section 14.1.5.5.1.4 Functional / System CDC - Isolate AFW flow to affected steam generator at 600 seconds.
Reference:
FSAR Table 14.1.5-7 2.5.5.4 RCS PRESSURE & INVENTORY CONTROL CSF Functional / System CDC - Initiate HPSI flow per design basis HPSI pump head curve.
Reference:
FSAR Table 14.1.5-7 Functional / System CDC - initiate charging flow per design basis pump capacity. Reference FSAR Table 14.1.5-7 2.5.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF - RCS analysis only. No applicable functionaUsystem CDCs. 2.5.5.6 ELECTRICAL POWER CSF - Plant loads power from offsite power source which is maintained through out event. No additional functional / system CDCs. 2.5.5.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functional / system CDCs. 2.5.5.8 ENVIRONMENTAL CONTROL CSF - RCS analysis only. No applicable functional / system CDCs. 13
--- _ _ - - - - ~ _ _ _ - -
2.5,6 FUNCTIONAL / SYSTEM CDCs - EVENT 14.1.5.6.2: MSLB @ IlZP The Critical Design Characteristics for the MSLB Event @ HZP, with a loss of off presented below. 2.5.6.1 REACTIVITY CONTROL CSE 6 ) Insert control rods within 3.9 seconds of _F,nnetionet/ System CDC - (Similar to event 14.1.5. 1
- eaching reactor trip setpoint.
Reference:
FSAR Table 14.1.5-4 j Functional / System CDC - (Similar to event 14.1.5.6.1) Inject boron to limit peak po
Reference:
FSARTable 14.1.5-8 Functional / System CDC - (Similar to event 14.1.5.6.1) Inject boron to ach conditionbeyond 600 seconds.
Reference:
FSARTable 14.1.5-8 2563 2.5.6.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSE - Functional / S and 2.5.6.4 below. 2.5.6.3 RCS HEAT REMOVAL CSE Close MSIV on intact main steam line to Functional / System CDC - (Similar to event 14.1.5.6.1) limit co Functional / System CDC - Limit AFW flow to affected steam generator for
Reference:
FSAR Section 14.1.5.5.1.4 Functional / System CDC -Isolate AFW flow to affected steam generator at 60
Reference:
FSAR Table 14.1.5-7 2.5.6.4 RCS PRESSURE & INVENTORY CONTROL CSE Initiate HPSI flow per design basis HPSI Functional / System CDC - (Similar to event 14.1.5.6.1) pump head curve.
Reference:
FSARTable 14.1.5-8 initiate charging flow per design basis pump Functional / System CDC - (Similar to event 14.1.5.6.1) capacity. Reference FSAR Table 14.1.5-8 2.5.6.5 CONTAINMENT INTEGRITY &_ RADIATION CONTROL CSE - RC applicable functional / system CDCs. 2.5.6.6 ELECTRICAL POWER CSE Functional / System CDC - Diesel Generator start and load to supply pow AFW pumps. 14
. ._ .. _. . . . . _ . . - . _ . . . _ . _ . . . . _ . . . _ _ _ _ ~ . . . . _ . . _ _ . . _ _ . _ . _ - ~ . . . . _ . . . . . _ _
- a. . .. .
t 4 i ?
' 2.5.6.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functional / system .
s CDCs. i i ! (' 2.5.6.8 ENVIRONMENTAL' CONTROL CSF - RCS analysis only. No applicable functional / system CDCs. 1 1 + 2 4. i 5 ) i e i l I i l 4 1 i
)
l
)
i i 15 i
2.5.7 FUNCTIONAL / SYSTEM CDCs - EVENT 14.1.5.6.3: MSLB @ HFP The Critical Design Characteristics for the MSLB Event @ HFP, with offsite power available, are presented below. l 2.5.7.1 REACTIVITY CONTROL CSF I Functional / System CDC - (Similar to Es ent 14.1.5.6.1) Insert control rods within 3.9 seconds of reaching reactor trip setpoint.
Reference:
FSAR Table 14.1.5-4 Functional / System CDC -Inject boron to limit peak power.
Reference:
FSAR Table 141.5-9 Note: HPSI and chargingpumo actuation times included under RCS Pressure & Inventory CSF. i l Functional / System CDC - Inject boron to achieve core subcritical condition beyond 600 seconds. I
Reference:
FSAR Table 14.1.5-9 l 2.5.7.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF - Functional / System CDCs included in 2.5.7.3 I and 2.5.7.4 below. l 2.5.7.3 RCS HEAT REMOVAL CSF 1 Functional / System CDC - (Similar to Event 14.1.5.6.1) Close MSIV on intact main steam line to limit cooldown from non-affected steam generator blowdown.
Reference:
FSAR Table 14.1.5-9 l Functional / System CDC - FW flow limited to that allowed by secondary pressure decrease for the first 30 seconds of the event.
Reference:
FSAR Section 14.1.5.5.1.4 ; Functional / System CDC - Isolate FW flow 30 seconds after the reactor trip. s Functional / System CDC -(Similar to Event 14.1.5.6.1) AFW flow initiated at runout value at t=180 j sec.
Reference:
FSAR Section 14.1.5.5.1.4 Functional / System CDC - (Similar to Event 14.1.5.6.1) Isolate AFW flow to affected steam generator at 600 seconds.
Reference:
FSAR Table 14.1.5-9 2.5.7.4 RCS PRESSURE & INVENTORY CONTROL CSE I Functional / System CDC - (Similar to Event 14.1.5.6.1) Initiate HPSI flow per design basis HPSI pump head curve.
Reference:
FSAR Table 14.1.5-9 Functional / System CDC - (Similar to Event 14.1.5.6.1) initiate charging flow per design basis pump capacity. Reference FSAR Table 14.1.5-9 16
r [ 2.5.7.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF - RCS analysis only. No applicable functional / system CDCs. 2,5.7.6 ELECTRICAL POWER CSF Epnctional/ System CDC - Transfer loads to offsite power source.
^
' 2.5.7.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functional / system CDCs.
2.5.7.8 ENVIRONMENTAL CONTROL CSE - RCS analysis only. No applicable functional / system CDCs. j 17
l l i
- 2.5.8. FUNCTIONAL / SYSTEM CDCs - EVENT 14.1.5.6.4
- MSLB @ HFP w/ LOOP !
1 The Critical Design Characteristics for the MSLB Event @ HFP, with a loss of offiste power, are l presented below. 1 l ' 2.5.8.1 REACTIVITY CONTROL CSF i 1
- Functional / System CDC - (Similar to Event 14.1.5.6.1) Insert control rods within 3.9 seconds of
- reachmg reactor trip setpoint.
[
Reference:
FSAR Table 14.1.5-4 i j Functional / System CDC - Inject boron to limit peak power. [
Reference:
FSAR Table 14.1.5-10 Note: HPSI and charging pump actuation times included under RCS Pressure & Inventory CSF. l Functional / System CDC -Inject boron to achieve core suberitical condition beyond 600 seconds. l
Reference:
FSAR Table 14.1.5-10 f
- j. 2.5.8.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF - Functional / System CDCs included in 2.5.8.3
! and 2.5.8.4 below. l 2.5.8.3 RCS HEAT REMOVAL CSF i i ! , Functional / System CDC - (Similar to Event 14.1.5.6.1) Close MSIV on intact main steam line to I i limit cooldown from non-affected steam generator blowdown.
Reference:
FSAR Table 14.1.5-10 l l 4 Functional / System CDC - FW flow limited to that allowed by secondary pressure decrease for the first } 30 seconds of the event.
Reference:
FSAR Section 14.1.5.5.1.4 L Functional / System CDC - Isolate FW flow 30 seconds after the reactor trip. Functional / System CDC -(Similar to Event 14.1.5.6.1) AFW flow initiated at runout value at t=180
- sec.
]
Reference:
FSAR Section 14.1.5.5.1.4 i Functional / System CDC -(Similar to Event 14.1.5.6.1) Isolate AFW flow to affected steam generator
; at 600 seconds.
L
Reference:
FSAR Table 14,1.5-10 2.5.8.4 RCS PRESSURE & INVENTORY CONTROL CSF Functional / System CDC - (Similar to Event 14.1.5.6.1) Initiate HPSI flow per design basis HPSI pump head curve.
Reference:
FSAR Table 14.1.5-10 I Functional / System CDC - (Similar to Event 14.1.5.6.1) initiate charging flow per design basis pump capacity. Reference FSAR Table 14.1.5-10 18
1 4 1 2.5.8.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF - RCS analysis only. No applicable functional / system CDCs. l l i 2.5.8.6 ELECTRICAL POWER CSF ; l Functional / System CDC - (Similar to Event 14.1.5.6.2) Diesel Generator start and load to supply I power to HPSI, charging, and AFW pumps. l 2.5.8.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functional / system CDCs. ) 2.5.8.8 ENVIRONMENTAL CONTROL CSF - RCS analysis only. No applicable functional / system CDCs, i 19
PARSONS POWER t l 1 Critical Safety Function Diagrams i Millstone Unit 2 i l l 1 I i l i t
- PARSONS FOWER -
Tier 2 Revision 0 18 July,1997
i . - 1 TABLE OF CONTENTS
- 1. Definitions 1
- 2. Diagrams 1
e REACTIVITY CONTROL e FUEL INTEGRITY & CORE HEAT REMOVAL e RCS HEAT REMOVAL 4
- RCS PRESSURE & INVENTORY CONTROL s
Rey, O
O A'ITACHMENT 2 CRITICAL SAFETY FUNCTIONS
- 1. REACTIVITY CONTROL Trip the reactor and maintain the reactor core shutdown margin within Techuical Specification Limits.
- 2. FUEL INTEGRITY & CORE HEAT REMOVAL Maintain core cooling to prevent fuel cladding failure and release of fission products to the reactor coolant system. The core should remain covered and reactor coolant subcooling margin maintained > 30 degF.
3 RCS HEAT REMOVAL Utilize the steam generators as an RCS heat sink. Qrx = Qeg = Usg Asg (Tave - Tsat) Deliver reactor core heat output (Qrx) to the steam genentors via forced flow or natural circulation. Maintain steam generator water level to provide sufficient wetted tube area (Asg) for primary to , secondary heat transfer. ' Control steam generator pressure (and thereby Tsat) to maintain the steam generator as an RCS heat sink. I Control steam generator pressure to control RCS temperature (Tave) and its rate / direction of change. l
- 4. RCS PRESSURE & INVENTORY CONTROL Maintain RCS pressure within the limits of EOP Figure 3.2 (RCS Pressure / Temperature Limits) for the given RCS temperature by use of a pressurizer steam bubble (preferred method) or emergency injection system throttling and CVCS/ relief valve operation. i Maintain RCS inventory sufficient to ensure core heat removal and primary to secondary heat transfer.
- 5. CONTAINMENT INTEGRITY & RADIATION CONTR.0L Maintain the integrity of the reactor building as a fission product boundary. For LOCA and steam generator tube rupture events, limit fission product release to maintain exposure within 10CFR100 limits.
Maintain Reactor Building environmental conditions (temperature, pressure, and combustible gas concentrations) within the RB design basis. Isolate non-accident mitigation process lines penetrating the containment structure. Maintain isolation capability for process lines required for accident mitigation. Provide penetration cooling. Maintain the integrity of electrical penetrations. Maintain the integrity of access hatches and their sealing function.
- 6. ELECTRICAL POWER Provide electric power to equipment required to achieve and maintain safe shutdown
- 7. ESSENTIAL SUPPORT SYSTEMS Maintain the operability of cooling water sysic~ essential to the proper operation of safe shutdown systems. (Component-specific systems are included with the safety function provided by that system.)
- 8. ENVIRONMENTAL CONTROL Maintain acceptable environmental conditions (temperature and radiation level) in plant areas requiring personnel access to achieve a safe shutdown condition Provide ventilation and cooling to vital equipment.
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