ML20044F506

Proposed Tech Spec Sections 1,2,4 & 5 & Pages B3.3-75 & B3.3-91

ML20044F506
Person / Time
Site:	05200001
Issue date:	05/20/1993
From:	GENERAL ELECTRIC CO.
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ML20044F492	List: ML20044F491 ML20044F506
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NUDOCS 9305280199
Download: ML20044F506 (66)

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{{#Wiki_filter:! Definitions 1.1 (g I.0 USE AND APPLICATION 1.1 Definitions ..................................... NOTE------------------------..-.-........ The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Term Definition f ACTIONS ACTIONS shall be that part of a Specification that i prescribes Required Actions to be taken under ] l designated Conditions within specified Completion Times. 74 g AVERAGE BUNDLE EXPOSURE The AVERAGE BUNDLE EXPOSURE shall be equal to the i sum of the axially averaged exposure of the fuel f v rods in the specified bundle divided by the number of fuel rods in the fuel bundle. AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicable to_ a specific planar height and is equal to the sum T of the exposure of all the fuel rods in the ( Os specified bundle at the specified height divided t by the number of fuel rods in the fuel bundle. AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific [ HEAT GENERATION RATE planar 'ght and is equal to the sum of the (APLHGR) ts-{LHGRs he t generation rate per unit length of fuel o; for all the fuel rods in the specified r bundle at the specified height divided by the number of fuel rods in the fuel bundlegat the heightlT~ CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it i responds within the necessary range and accuracy l to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alam, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors shall consist of an inplace cross calibration of the sensing elements and nomal calibration of the i (continued) O ABWR% STS 1.1-1

1. cv. O 09/28/C?

9305280199 930520 PDR ADOCK 05200001 A PDR

1 i Definitions 1.1 s -[ 1.1 Definitions I CHANNEL CALIBRATION remaining adjustable devices in the channel. (continued) Whenever a sensing element is replaced, the next required inplace cross calibration consists of i comparing the other sensing elements with the' I recently installed sensing element. The CHANNEL CALIBRATION may be perfomed by means of any series of sequential, overlapping, or total-channel steps so that the entire channel is calibrated. ( CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This detemination shall include, where possible, comparison of the chennel i indication and status to other indications or i status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be: a. Analog channels-the injection of-a simulated-i or actual signal into the channel as close to i the sensor as practicable _ to. verify OPERABILITY, including required alam, interlock, display, and trip functions, and channel failure trips. b. Bistable channels (e.g., pressure switches and I switch contacts)-the injection of a ' simulated or actual signal into-the channel as close to. . the sensor as practicable to verify OPERABILITY, including required alarm and trip. functions. l The CHANNEL FUNCTIONAL TEST may be perfomed by. means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested. t CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other j components affecting reactivity within the reactor vessel with the vessel head removed and fuel in .i the vessel. Movement of senace rang local power range monitors, int = 1 g itors,- ...l n. _., 1 t e ts+ro ^ Sh)cN (continued) ABWR'/4 STS. 1.1-2 -an. O, 09/20/02

y L! i Definitions

1.1-l r

i t 1.1 -Definitions i meM4er4, traversin I CORE ALTERATION movable detectors (g incore probes, or special (continued) including undervessel-replacement) is not considered a CORE ALTERATION. j In addition, control rod movement with.other thanL j the nomal control rod drive is not considered a j CORE ALTERATION provided there are.no fuel assemblies in the associated. core cell. Suspension of CORE ALTERATIONS shall not preclude-i completion of movement of a component to a safe i position. ~ t L CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle-specific-limits a shall be detemined for each reload cycle in' l accordance with Specification EWitWc. Plant ~. operation within these limits 's addressed in j individual Specifications. {g,y] q DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration-i of I-131 (microcuries/ gram) that alone would produce the:same thyroid dose'as the quantity ~and q T isotopic mixture of 1-131, I-132, 1-133, I-134, pdera L dalhu and I-135 actually present. The: thyroid dose conversion factors ~ used for this calculation shall b* DethoselistediniTd!: I!! ef TIO-10004, AEC, 1962, "C:!celet h 'Of'O ht = : B:ter: f:r G [""?..' =""$ Id 5'fiE39',DI ".'. 1".".*Ei',F", l '"~ =

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1 E-AVERAGE G shall be the average (weighted in proportion-DISINTEGRATION ENERGY to the concentration of each radionuclidej n the: l, rea'etor coolant at the time of: sampling);oiltheL C sum of the avera a disintegrationi(ge beta and gamma energies perin MeV) f g iodines, with half lives 4115 Minutes, making up, at'least 95% of the total noniodine: activity in the coolant.. s EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be ~ that. time interval i SYSTEM (ECCS) RESPONSE 'from when the monitored parameter exceedsjits ECCS! TIME -initiation setpoint'at the channel' sensor'until the ECCS equipment is capable' of perfoming its -l safety' function (i.e.,~ the valves travel; to their j (continu'd) e O 1 ASWR/4 STS 1.1-3 h. O, 09/2o/u t n i

Definitions 1.1-1.1 Definitions eEMERGENCYCORECOOLING required positions, pump discharge pressures reach SYSTEM (ECCS) RESPONSE their required values, etc.). Times shall include TIME diesel generator starting and sequence loading Gontinuedp delays, where applicable. The response time may. be measured by means of any series of sequential, overlapping, or total steps so that the entire hg.A#u response time is measured. 6A ENDOFCYCLE( The EOC-RPT SYSTEM RESPONSE TIME shall be that RCCRC'JLATIO" PUMP TRIP time interval from initial signal generation by (EOC-RPT) SYSTEM RESPONSE the associated turbine stop valve limit switch or TIME from when the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpointP@e;;;let: : ppr::si:n Of the-e: ectr; :r: bet ;;n th 511; :p;n ::nt :t; cf p;p o f + g* the r::f rc.htien y..ay circ.it br: de. The idg ger response time may be measured by means of any e de-N 3 g t s-series of sequential, overlapping, or tota! steps v e r so that the entire response time is measured. 3p=*) ) v{ r# [ :::pt S r th: bre;Ler erc 3 yyicniv.. i;e, a s M' .r o .r y which is n:t ;;;es..cd i,ut is vaiidaird Le cenfmn naj-fDi[r' t .tt 10 the zon.fedur er i design..;.m]. ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE

1. LEAKAGE'into the drywell such as that from pump seals or valve packing, that is captured and' conducted to a sump or collecting tank; or (continued)

O i prBWR745TS 1.1-4 h. O,00/2S/9L o

Definitions 1.1 /~'T 1.1 Definitions L,I LEAKAGE

2. LEAKAGE into the drywell atmosphere from (continued) sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per (h RATE (LHGR) unit length of fuel rod. It is the integral of g V the heat flux over the heat transfer area g_ associated with the unit length. LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components (i.e., all relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from ac close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.- The LOGIC SYSTEM FUNCTIONAL TEST may be perfomed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested. ~ CTION The MFLPD shall be the larcest vah:: ;f the m 0F LIMITING imiting power density in the core. POWER DENSITY am fr ~ of limiting power density _ shall be the LHGR existin n location divided by the specified LHGR limit for that Tundic 1N_ [ ') (continued) \\_/ ABWR71E STS 1.1-5 A v. O, 09/2*/9?

Definitions 1.1~ ["') 1.1 Definitions (continued) A.J MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core.{fr ::ch th:: Of fe']. The CPR is that power in the assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE-0PERABILITY A system, subsystem, train, component, or device shall be OPERABLE when it is capable of perfoming its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform g its specified safety function (s) are also capable of perfoming their related support function (s). PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are: a. Described in Chaoter 14, Initial Test i Program ft SAR; b. Authorized under the provisions of 10 CFR 50.59; or c. Otherwise approved by the Nuclear Regulatory Comission. PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature lirrits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits (continued) A ABWR/fa STS 1.1-6 ke. 3, Og/20/ M

Definitions 1.1 1.1 Definitions l l PRESSURE AND shall be determined for each fluence period in j TEMPERATURE LIMITS accordance with Specification Mt:$;2. Plant l REPORT (PTLR) operation within these operati g limits is (continued) addressed in LCO 3.4. "RCS ressure and Temperature (P/T) Limits." [s. x-RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of [M27] MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is y f,,,,,. c measured. I f SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: o f e. : A c en t re t r ah / ^ 'T a. The reactor is xenon free; M c..m ;3 %,4 twe c d"' P.J 3 oli ck c re c...ect =J b. ThemoderatortemperatureisI68' hand I

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j pt r c c c % L t o r. A LL All control rods are(fully inserted except for c.

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the s @ control ro Pof highest reactivity ff[r 4. e J 4Yc worth, which is assumed to be fully withdrawn. .s i+s.we c.ct = ut ter-. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of ( ,SDM. STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. x REFUELING INTERVAL REFUELING INTERVAL is the period of up to 24 months duration \\ from startup following a refueling outage until startup following the subsequent refueling outage. SR 3.0.2 is applicable. N

Definitions 1.1 /N ig 1.1 Definitions (continued) THERMAL POWER THEPMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME consists RESPONSE TIME of two components: a. The time for initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and f e-t-b. The time for initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. r gV ABWR/6STS 1.1-8 Av. O, 03/2S/92

Definitions 1.1 T (d Table 1.1-1 (page 1 of 1) MODES REACTOR MODE AVERAGE REACTOR MODE TITLE SWITCH POSITION COOLANT TEMPERATURE ct.("F) 1 Power Operation Run NA 2 Startup Refuel (a) or Startup/ Hot NA Standby 3 HotShutdown(a) Shutdown > 200 73 d 4 Cold Shutdown (a) Shutdown s 200 73 5 Refueling (b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned. (b) One or more reactor vessel head closure bolts less than fully tensioned. 1 l l l l I frBWR76STS 1.1-9

Rev, 0, 09/28/gt

l Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors ( i PURPOSE The purpose of this section is to explain the meaning of 't logical connectors. I Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logica; connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings. BACKGROUND Several levels of logic may be used to state Required i Actions. These levels are identified by the placement (or l nesting) of the logical connectors and by the number i assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement' of the logical connector in the first level of nesting (i.e., left justified with the i number of the Required Action). The successive levels of l O logic are identified by additional digits of the Required l Action number and by successive indentions of the logical Conttectors. l When logical connectors are used to state a Condition, only the first level of logic is used, and the logical connector is left justified with the Condition statement. ^ When logical connectors are used to state a Completion Time, Surveillance, or Frequency, only the first. level of logic is 'j used, and the logical connector is left justified with-the t statement of the Completion Time, Surveillance, or ) Frequency. J i EXAMPLES The following examples illustrate the use of logical i connectors. (continued) O ABWR[$ STS 1.2-10 Rh. O, 09/25/92 i

Logical Connectors 1.2 (~ r 1.2 Logical Connectors EXAMPLES EXAMPLE 1.2-1 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Veri fy... AND A.2 Restore... In this example, the logical connector AND is used to indicate that, when in Condition A, both Required Actions A.1 and A.2 must be completed. P (continued) pr3WR)(STS 1.2-11 Rcv. O, 09/20/92-l

i Logical Connectors I 1.2 e 1.2 Logical Connectors i EXAMPLES EXAMPLE 1.2-2 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Trip... E A.2.1 Ve ri fy... l AND A.2.2.1 Reduce... M A.2.2.2 Perfom... M A.3 Align... This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be perfomed as indicated by the use of the logical connector M and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be perfomed as indicated by the logical connector AND. Required Action A.2.2 is met by perfoming A.2.2.1 or A.2.2.2. The indented position of the logical connector E indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be perfomed. i nU ABWR M STS 1.2-12 hv. v, uv/2eis2

Completion Times 1.3 ( 1.0 USE AND APPLICATION 1.3 Completion Times f PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND LCOs specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action (s) and Completion Time (s). i DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an t ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the A Applicability of the LCO. Required Actions must be 3 V completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability. If situations are discovered that require entry into more than one Condition at a time within a single LC0 (multiple Conditions), the Required Actions for each Condition must be perfomed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition. Once a Condition has been enterd, subsequent trains, subsystems, componeats, or variab es expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition unless t specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. (continued) O A BWRESTS 1.3-13 Rs. O, 09/20/02

l Completion Times 1.3 1.3 Completion Times DESCRIPTION However, when a subsecuent train, subsystem, component, or i (continued) variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time (s) may i be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability: a. Must exist concurrent with the first inoperability; ) I and b. Must remain inoperable or not within limits after the first inoperability is resolved. The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either: a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours; or b. The stated Completion Time as measured from discovery l of the subsequent inoperability. O The above Completion Time extensions do not apply to those i Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications. The above Completion Time extension does not apply to a Completion Time with a modified " time zero." This modified l " time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase " f rom di scove ry... " Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended. I 1 l l Q (continued) prBWR/RSTS 1.3-14 Rev. O, 05/28/92 l

Completion Times 1.3 1.3 Completion Times (continued) EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions. EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met. ,c Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. The Required Actions of Condition B are to be in MODE 3 within 12 hours AND in MODE 4 within 36 hours. A total of 12 hours is allowed for reaching MODE 3 and a total of 36 hours (not 4B hours) is allowed for reaching MODE 4 from the time that Condition B was entered. If MODE 3 is reached within 6 hours, the time allowed for reaching MODE 4 is the next 30 hours because the total time allowed for reaching MODE 4 is 36 hours. If Condition B is entered while in MODE 3, the time allowed for reaching MODE 4 is the next 36 hours. (continued) /bWR74STS 1.3-15 D.e v. O, 09/2S/02 l ^

i Completion Tinies l 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued) j ACTIONS ~ CONDITION REQUIRED ACTION COMPLETION TIME i A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE acatus. B. Required B.1 Be in ynnE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met. t When a pump is declared inoperable Condition A is entered. O If the pump is not restored to OPERABLE status within 7 days, Condition B is entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the i inoperable pump is restored to OPERABLE status after i Condition B is entered, the Required Actions of Condition B may be-terminated. When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered.for l the second pump. LC0_3.0.3 is entered, since~the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for_ Condition A does not'stop ~i after LCO 3.0.3 is entered, but continues to be tracked from 1 the time Condition A was initially _ entered. j While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for i Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A. { (continued) O ASWR/ti STS 1.3-16 Re.. C, 09/28/0? l 1

Completion Times 1.3 V 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued) While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired. On restoring one of the pumps to OPERABLE status, the Condition A. Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days. O E i (continued) @WRhi STS 1.3-17 1c v. O,09/25/92

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7 days Function X Function X subsystem subsystem to AND inoperable. OPERABLE status. 10 days from discovery of failure to meet the LCO B. One B.1 Restore 72 hours Function Y Function Y subsystem subsystem to AND inoperable. OPERABLE stat' J 10 days from discovery of failure to meet the LCO C. One C.1 Restore 72 hours Function X Function X subsystem subsystem to inoperable. OPERABLE status. AND ,0_B One C.2 Restore 72 hours Function Y Function Y subsystem subsystem to inoperable. OPERABLE status. (continued) O ABWR)4STS 1.3-18 Rcv. O, 09/25/92

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train, starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered). If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the a Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A). The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time O measured from the time it was discovered the LCO was not met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. The separate Completion Time modified by the phrase "from discovery of failure to meet the LC0" is designed to prevent indefinite continued operation while not meeting the LCO. This Completion Time allows for an exception to the nortnal " time zero" for beginning the Completion Time " clock". In this instance, the Completion Time " time zero" is specified as commencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered. 4 I (continued) ABWR/li STS 1.3-19 Rn. O, 09/24/42

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-4 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve (s) 4 hours valves to OPERABLE inoperable. status. B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met. P O A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times. Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours. If the Completion Time of 4 hours (including any extensions) expires while one or more valves are still inoperable, Condition B is entered. (continued) O $BWRM STS 1.3-20 Re. O, 09/28/92

Completion Times-1.3 j 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued) ACTIONS ............................N0TE---------------------------- Separate Condition entry is allowed for each inoperable valve. l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours valves OPERABLE status. inoperable. e B. Required B.1 Be in MODE 3. 12 hours Action and associated AND O Completion Time not B.2 Be in MODE 4. 36 hours met. The Note above the ACTIONS table is a method of modifying l how the Completion Time is tracked. If.this method of ~' modifying how the Completion Time is tracked was applicable only to Condition A, the Note may appear in the Condition l column. The Note allows Condition A to be entered separately for i each inoperable valve, and Completion. Times tracked on a per valve basis.- When a valve is declared inoperable, Condition A is entered and its. Completion Time starts. If ~ subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start' and are tracked for each valve. l (continued) i ABWR/4 STS 1.3-21 hv. O,00/28/92 l

3 Completion Times i 1.3 n -( ) 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued) If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. l If the Completion Times associated with subsequent valves in Condition A expire, Con dition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION -COMPLETION TIME (\\ ~( ) A. One channel A.1 Perform Once per inoperable. SR 3.x.x.x. 8 haurs OR A.2 Reduce THERMAL 8 hours POWER to s 50% RTP. B. Required B.1 Be in MODE 3. 12 hours Action and associated Completion Time not met. (continued) h,m ffBWR/(STS 1.3-22 Rr.. O, 09/29/9fh-

i Completion Times 1 1.3 l 1.3 Completion Times i EXAMPLES EXAMPLE 1.3-6 (continued) Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each perfomance after the initial perfomance. If Required Action A.1 is followed and the Required Action is not met within the Completion Time (including the 25% extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered. If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A. O i (continued) O A BWR7{i STS 1.3-23 -Rev. O, 09/28/92 1

1 Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected I hour subsystem subsystem inoperable. isolated. AND Once per 8 hours thereafter AND A.2 Restore subsystem 72 hours to OPERABLE status. x B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met. Required Action A.1 has two Completion Times. The I hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.I. If after Condition A is entered, Required Action A.1 is not met within either the initial I hour or any subsequent 8 hour interval from the previous f (including the .25% extension allowed by SR 3.0.2)per ormance, Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 (continued) 7-~\\ j \\~ J ABWR/(STS 1.3-24 Rev. O, 09/28192 ) i 1

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered Condition B is exited and operation may continue in accordance with Ccndition A, provided the Completion Time for Required Action A.2 has not expired. Since the second Completion Time of Required Action A.1 has a modified " time zero" (i.e., after the initial I hour, not from time of Condition entry), the allowance for a Completion Time extension does not apply. IMMEDIATE When "Imediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner. O O hBWR/t STS 1.3-25 Re.. O, 00/23/92

Frequency 1.4 r r 1.0 USE AND APPLICATION 1.4 Frequency-PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements. -l DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to-meet.the l associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. i The "specified Frequency" is referred to throughout this i section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified-l Frequency" consists of the requirements of the. Frequency t column of each SR, as well as certain Notes in the Surveillance column that modify perfomance requirements. Sometimes special situations dictate when the requirements [ of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated O as clarifying Notes in the Surveillance,~as part of the Surveillance, or both. Example 1.4-4 discusses these special situations. Situations where a Surveillance could be required (i.e., its e Frequency could expire), but where it is not possible or not desired that it be perfomed until sometime after.the q associated LCO is within its Applicability, represent j potential SR 3.0.4 conflicts. To avoid these conflicts, the 1 SR (i.e., the Surveillance or the Frequency) is stated such j that it is only " required" when it can be and 'should be i performed. With an SR satisfied, SR 3.0.4 imposes no l restriction. j -l The use of " met or performed" in these instances conveys specified meanings. A Surveillance is " met" only when the r acceptance criteria are satisfied. Known-failure of-the requirements of a Surveillance, even without a Surveillance l specifically being " performed," constitutes a Surveillance not " met." "Perfomance" refers only to the requirement to i specifically determine the ability to meet the acceptance l (continued) O fBWR/6STS 1.4-26 Rev. O, 09/2S/92 i i

m Frequency 1.4 [&' 1.4 Frequency DESCRIPTION criteria. SR 3.0.4 restrictions would not apply if both the (continued) following conditions are satisfied: a. The Surveillance is not required to be perfomed; and b. The Surveillance is not required to be met or, even if required to be met, is not known to be failed. EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LC0 (LCO not shown) is MODES 1, 2, and 3. EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perfom CHANNEL CHECK. 12 hours Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time. Perfomance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the perfomance of the Surveillance is not otherwise modified I (continued) AV hBWR/ft STS 1.4-27 Rw. O, 09/fs/92

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued) (refer to Examples 1.4-3 and 1.4-4), then SR 3.0.3 becomes applicable. If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4. j EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within O 12 hours after e 25% RTP b.!LD 24 hours thereafter Example 1.4-2 has two Frequencies. The first is a one time - performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to a 25% RTP, the Surveillance must be performed within 12 hours. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2. (continued) O $8d/4STS 1.4-28 ae.. c, og/2m i

Frequency 1.4 (y r j 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued) "Thereafter" indicates future perfomances must be established per SR 3.0.2, but only af ter a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP. EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I N0TE-----....----..... Not required to be performed until 12 hours after a 25% RTP. Perfom channel adjustment. 7 days f The interval continues whether or not the unit operation is < 25% RTP between performances. As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches a 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified frequency." Therefore, if the Surveillance were not performed within the 7 day (plus 25% per SR 3.0.2) interval, i but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power a 25% RTP. (continued) O hBWRMSTS 1.4-29 fev. O, 09/20/02-

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.a-3 (continued) Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency; MODE changes then would be restricted in accordant' with SR 3.0.4 and the provisions of SR 3.0.3 would apply. EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE------------------

Only required to be met in MODE 1. Verify leakage rates are within limits. 24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour (including the 25% extension allowed by SR 3.0.2) interval, but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR. O hBWR?fi.STS 1.4-30 4. O, 09/23/92

4 SLs i 2.0 O 2.0 SAFETY LIMITS (SLs) l 2.1 SLs [55.1/<y/cd } y 2.1.1 Reactor Core SLs j Withtherear.'.orsteamdomepressure<[785psig)orcore f 2.1.1.1 flow < 10% rat'd core flow: THERMAL POWER shall be s 25% RTP. SS A Sl ( g r-2.1.1.2 With the reactor steam dome pressure a 85psig)andcore flow e 10% rated core flow-MCPR shall beDI.07br tre h;p r::irc;.hti -

r tier. Or

[1.00] fei si.,;'; '.;p racimhtien i

irrt'r. 2.1.1.3 Reactor vessel water level shall be greater than the. top of active irradiated fuel.

93. /5 $ /CMJ 2.1.2 Reactor Coolant System Pressure SL s

Reactor steam dome pressure shall be maintained s (k325 psigf 2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: .l 2.2.2.1 Restore compliance with all SLs; and l 2.2.2.2 Insert all insertable control rods. 2.2.3 Within 24 hours, notify the [ General Manager-Nuclear Plant and-Vice President-Nuclear Operations) and the [offsite reviewers specified in Specification [0ffsite] Review and Audit"]. [ 5.XJ (continued) i O ASWRASTS 2.0-1 -Re v. O, 03/2E/;I-j

SLs 2.0 l' k 2.0 SLs 2.2 SL Violations (continued) ( 8' 2.2.4 Within 30 days, a Licensee Event Report (LER) shall b prepared pursuant to 10 CFR 50.73. The LER shall be submitte to the NRC, the [offsite reviewers specified in Specification ], and the [ General Manager-Nuclear Plant and Vice President-Nuclear Operations]. 2.2.5 Operation of the unit shall not be resumed until authorized by the NRC. O I L i f I O flBWR76 STS 2.0-2 h. O, 09/25/92

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs) B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s). The fuel cladding integrity SL is set, such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, stepback approach is used to establish an SL, such that the MCPR is no ess than the limit specified in Specification 2.1.1.2 for boetf General Electric (GE) Corporation =d 'iW lMhar I.;d ',V:T) C:g:ntion fuelJP MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thennal stresses, which occur from reactor operation significantly above design conditions. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during nonnal operation and during A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling. (continued) ABWR?ik,STS B 2.0-1 Rev. O, 03/25/92-

Reactor Core SLs B 2.1.1 () BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker fom. This weaker fom may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant. APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation and A00s. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR is to be established, such that at least 99.9% of the feel rods in the core would not be expected to experience the onset of transition boiling. The Reactor Protection System setpoints (LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"), in combination with all the LCOs, are designed to prevent any (^') anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR. 2.1.1.1Y Fuel Claddino Inteority (General Electric 2 ' ' Corooration (GE) Fuel) S S.'L b GE critical power correlations are applica e for all For operation at low pr)essuresQ85 critical power calculations at pressures flows a 10% of rated flow. and low flows, another basis is used, as follows: Since the pressure drop in the bypass region is .t essentially all elevation head, the core pressure 3 /6 KJIC"'E drop at low power and flows will always be D.5 psi). AnaJyses (Ref. 2 show that with a bundle flow of'48 x 10' lb/h bundle pressure ,N6 Kt M : drop is nearly. independent o bundle power and I'l.g [ D ~g gg has a_ vplue of'(3.5 psi). Thus, the bundle flow wiJh a'(4.5 psi) c riving head will be J [2B x 10) lb/hr Full peale ATLAS.t st data g.7 rog taken at pressures fro (14.7 psig to 800 psia) 1 H/Cm A-S t. 3. J IbI* f (continued) mL) [tBWR?fs STS B 2.0-2 Rev. O, CS/28/92

i Reactor Core SLs B'2.1.1 ~! BASES 2.1.1.1( Corooration (GE) Fuel Claddino Intecrity (General Electric APPLICABLE Fuel) (continued) l SAFETY ANALYSES indicate that the fuel assembly critical power at ) this flow is approximately 3.35 Mwt. With the j design peaking-factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMA -POWER l limitof25%RTPforreactorpressure<,85psig) l is conservative. Q t 1.1.1.1b Fuel Claddina In<:ecrity (Advanced Nuclear Fuel Corocration (ANF l Fuel) !'i The u of the XN-3 correlation is. valid for cri cal power calcula 'ons at pressures > 580 psig and bundl mass fluxes > 0.25 x lb/hr-ft' (Ref. 3). For-operati at low 6 pressures low flows,- the fuel cladding i egrity SL is - i established a limiting condition on co. THERMAL POWER, I with the folio 'ng basis: Provided that he water level i the vessel _(' downconer is ma' tained above he top of thel active fuel, nat al circul ion is sufficient-to l ensure a minimum b die f for all fuel. 1 assemblies that have atively high power and .i potentially can appro 1 a critical heat flux condition. For the F x9 fuel design, the. j 3 minimum bundle f1 is > x 10 lb/hr. For the ANF 8x8 fuel des' n, the a inum bundle flow is - 8 > 28 x 10 lb/ For all d igns the coolant minimum bund flow and maxi flow area are j such that t mass. flur.'is alwa "j > 0.25 x lb/hr-ft'. - Full. sea critical power tests ta n at pressures down to 1 7 psia indica that the fuel assembly crit al power at 2 0.25-10' lb/hr-ft is t 3.35 Mwt. A 25% RTP, a bun e power of 3.35 Mwt corresponds to bundle' ~ ra al peaking factor of > 3.0, which is-gnificantly higher than the expected pea 'ng-actor. Thus, a THERMAL POWER limit of 25% TP o for reactor pressures < 785' psig is conservat e. i l l (continued)- O 1 O BWRhk STS B_2.0-3 Rev. G, C^/C/92 { k ~ n

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2( MCPR (GE Fuel) SAFETY ANALYSES (continued) The fuel cladding integrity SL is set, such that no significant fuel damage is calculated to occur if the limit-is not violated. Since the parameters that result in fuel damage are not directly observable. during reactor operation, the thennal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage-could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods,' the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in i the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. j Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties. -O The MCPR SL is detemined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is detemined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also t includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis. j hLl.2b MCPR (ANF Fuell The MCPR sures sufficient conservatis the operating MCPR limit tha, ' the event of an rom the limiting condition of operat at less .9% of the fuel rods in-the core would be expec avoid boiling transition. The margin between calcu b ' ing transition (i.e., MCPR = 1.00) a e MCPR SL i sed on a detailed statisti rocedure that conside he uncertainties in i non ng the core operating-state. specific j ertainty included in the SL is the unce inty inherent !j ,I (continued) j O d /t BWRM STS B 2.0-4 Rev. G 09/23/02

-m Reastor Core SLs I B 2.1.1 BASES l APPLICABLE T.1.1.2b MCPR (ANF Fuel) '(continued) SAFETY ANALYSES in e XN-3 critical power correlation. Referenc 3 i descr es the methodology used in detemining t MCPR SL. The XN-3 ritical power correlation is base on a significan body of practical-test data, viding a high-degree of as rance that the'eritical p r, as evaluated by the correlatio is within a small per ntage of the actual critical power be' g estimated. As ng as the core pressure and flow a within the r ge of validity of the XN-3 correlation, the ssumed re tor conditions used in defining.the SL introdu cons vatism into the limit.' because bounding high ra 1 ower factors and bounding flat local peaking distributions re used to estimate the number 1 of rods in boiling transi o Still further conservatism is induced by the tende y.of e XN-3 correlation to overpredict the numbe of rods i boiling-transition. These-i conservatisms and t inherent acc acy of the XN-3 i correlation provi a reasonable deg of assurance that there would be transition boiling i he core during l sustained oper ion at the MCPR~SL. If b ling transition were to occ, there is reason to believe t it the integrity-l of the fue would not be compromised.. Signi ~ ant test data accumula d by the NRC and private organizatio indicate that t use of a boiling transition limitation protect agai t cladding failure is a very conservative approach. E M of the data indicate that BWR fuel can survive for an tended period of time in an environment of boiling-transition. .lJ 2.1.1.3 Reactor Vessel Water Level During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level. requirements due to the effect. of decay heat.. If the water level should drop below the top of the active irradiated fuel during this period, the -l ability to remove decay heat is reduced. This reduction in l cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the j water level becomes < % of the core height.. The reactor vessel water level SL-has been established at.the top' of the j (continued) 8BWR?SSTS B 2.0-5 5. C, 09/20/02

t Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued) i SAFETY ANALYSES active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action. i SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioa:tive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel, thus maintaining a coolable geometry. APPLICABILITY SLs 2.1.1.1, 2.1.1.2 and 2.1.1.3 are applicable in all MODES. However, in MODES 3, 4, and S, with the reactor shut down, it is unlikely that fuel cladding integrity SLs would be violated. O v SAFETY LIMIT W VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref.p). 3 E 4 Exceeding an SL may cause fuel amage and create a potential for radioactive releases in e cess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal. (Continued) (] U flBWRhi STS B 2.0-6 h. O, 09/25/03 l

i i Reactor Core SLs B 2.1.1 lO BASES I %) l SAFETY LIMIT L.L3 VIOLATIONS (continued) If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified witnin 24 hours. The 24 hour period provides time for plant operators and staff to take the appropriate inunediate action and assess the condition of the unit before reporting to the senior management. I W If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC, the senior management of the nuclear plant, and the utility Vice ~i President-Nuclear Operations. This requirement is~in accordance with 10 CFR 50.73 (Ref. ). L.Zd If any SL is violated, restart of the unit shall not f commence until authorized by the NRC. This requirement t f-( ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to } nonnal operation. j t REFERENCES 1-10 CFR 50 Appendix A, GDC 10. 2. NEDE-24011-P-A, (latest approved revision). 4. m-;iF;24 (".), S r! eiea 1, t. ;;-i : r 1 000. l 3 4. 10 CFR 50.72. i y '5. 10 CFR 100. 5 't. 10 CFR 50.73. i O 8BWR A STS B 2.0-7 Re,. O, 00/2S/92

~ RCS Pressure SL B 2.1.2 Iv) B 2.0 SAFETY LIMITS (SLs) l B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the (RCS) against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15. " Reactor Coolt.nt System Design" (Ref.1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during nonnal operation and anticipated operational occurrences (A00s). During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. During nonnal operation and A00s, RCS pressure is limited ( from exceeding the design pressure by more than 10%, in v accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core is done under LCO 3.10.1, " Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result in a breach of the RCPB. If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, " Reactor Site Criteria" (Ref. 4). (Continu2d) ftBWRWi STS B 2.0-8 R;:v. O, 09/2a/M-- i

RCS Pressure SL c B 2.1.2 BASES (continued) APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded. g g Js g7,9g3fcgg TheRCSpressureSLhasbee[selectedsuhthatitisata pressure below which it can'be shown tha the integrity of the system is not endangere The reacto pressure vessel is designed to ASME, Boiler d Pressure V sel Code, Section III, [T*" TW) cluding Adden through the 23 [ey -

f WL

f. 5),

i pemits a max pressure g,7 k-transient pf 110% 375 psi of esign pressur seamdome[ 250 psig The K1 of,(1325 psi as mea red b the react dessure indicator, is equivalent to (l375 psig at the lowest cf 3,J f elevation of the RCS. The RCS is designed to SME Code, l(3 /C rd Section III, 1974 Edition (Ref. 6), for the react f 79 Kjhg recirculation piping, which pemits a ma ' regsure transient of 110% o design pressures of 250 psig)for suction piping and 650 psig The RCS pressure SL is sel cted to b)for discharge piping. e the lowest transient overpressure allo ed by the applicable codes. Ilb K3}C d3 l I05.5 % Jcd5 O SAFETY LIMITS The maximum transient pressure allow le in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the R ld250 psig) valves, and fittings isfor suction T ' Ping and"

piping, 10% of esign pressures 13 65f of g

?M piping. The most limiting of these wo allowances is the 110% of design pressure; therefore, the Sj. on maximum allowable RCS pressure is established a 375psigh 9/.7 # 9 /c[9 APPLICABILITY SL 2.1.2 applies in all MODES; however, in MODE 5, because the reactor vessel head closure bolts are not fully tightened, it is unlikely the RCS would be pressurized. i SAFETY LIMIT 2,2.1 VIOLATIONS i If any SL is violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref. 7). (continued) [iBWR?6 STS B 2.0-9 k.. O 09/25/92

i t RCS Pressure SL i B.2.1.2 BASES SAFETY LIMIT 2.2.2 VIOLATIONS l (continued) Exceeding the RCS pressure SL may cause inmediate RCS j failure and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits i (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action. l 2.2.3 i If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within - 24 hours. The 24 hour period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the l senior management. l W If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC, the senior i management of the nuclear plant, and the utility Vice President-Nuclear Operations. This requirement is in accordance with 10 CFR 50.73 (Ref.-8). I L2d 1 i If any SL is violated, restart of the unit shall not i 4 commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. { t i REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28. 2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000. I i (continued) bBWR76.STS B 2.0-10 Rev. O, 09/28/92 j 1 4

RCS Pressure-SL'- l B 2.1.2 i OQ BASES REFERENCES 3. - ASME, Boiler and _ Pressure Vessel Code, Section XI, (continued) Article IW-5000'- j f 4. 10 CFR 100. [ 5. ASME, Boiler and Pressure Vessel Code [Iti44datts], i Addenda, [ C t. 9 ivii]. l 6. ASME, Boiler and Pressure Vessel Code [ISP4Midttizm]. f 7. 10 CFR 50.72. l t 8. 10 CFR 50.73. i -i + .I t t + O 1 i i t i

l t

.l t i i

i

-i [4BWR/6-STS B 2.0-11 Rev. O, 09/26/92'

a q Design Features 4.0 4.0 DESIGN FEATURES' 4.1 Site 4.1.1 Site and Exclusion Area Boundaries 4 The site and exclusion area boundaries [shall be as described or as shown in Figure 4.1-1). 4.1.2 Low Population Zone (LPZ) l The LPZ [shall be as described or as shown in Figure 4.1-2]. l 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium O dioxide (U0,) as fuel material?f, and water rodsW Limited i substitutions of zirconium alloy or stainless steel filler rods j for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC _i staff ~ approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead t test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 Control Rod Assemblies og The reactor core shall contain D937 cruciform shaped control rod l assemblies. The control material shall be 0:r:r carbi6. +aN" - -t;& as approved by the NRC. I (continued) A BWRA STS. 4.0-1 Rs. O, 00/28/62

.j .ii E~ Design Features { 4.0 4.0 DESIGN FEATURES (continued) i; 4.3 Fuel Storage 1 4.3.1 Criticality 4.3.1.1 The spent' fuel storage. racks are designed and shall.. be maintained with: a. Fuel assemblies having a maximum k-infinity of [3eKI]. in ~ the nonnal reactor core configuration at cold conditionsy f [r-- :- " '"5 ---':'z;; ;f M) .y;.;,,, r;;;;] ; b. k,$1udes an allowance for uncertainties as described in N in v{Section9.1ofthe/gSAR]'i' j [c. ', ::ir:1 feel ess--t!y c--ter t: ;;nt:r ;ter;;; ;;;;;i g 1 i ef p] in:h:r -i+h4= rer: : d [12.25] i :h:: b:te:: ma n injhe[le.4up;;f 5,ter ge recks] ;n Un:.,,,,; r wn6aan=wu6 puuai anuj-l [u. A nominal fuel assembly center to center storage spacing of [8d5] inches, within a neutron poison material between torage spaces, in the7high d in the spe,nt-fuel' storage poolg,ensity storage racks ed th; ;;;:r - ,...._.4._..=_.. _...,....a 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: f.{, 3,5-a. Fuel assemblies having a maximu k-infinity of [2H)Q in the nonnal reactor core configuration-atq:ld rrditkr.;F r_ --- = ....f_t___.....-..... r. ; -= i v... u.......1,. nge.__ s e' +i_ 2 .r..7 9 0 *c. b. k,,,ludes an allowance for uncertainties as described ins; inc / 3Section9.1oftheFSART 4 [- k,,, c 0.9S 'i' ::d:r;ted b, eq.ee.; fee, hich ir.cl. des i e e!!:;;r.ce fur uncertaintin u hicrit;d it. [S::tica 0.1 ei ine is,^, ], omt y C 11. A nominal [596] inch center. to center distance between i fuel assemblies placed in storage racks. i (continued) l DBWR/n STS 4.0-2 MeV

a. no m /92' _

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.2 Drainace The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below e!: : tier. [^^^ ft 5.25 inch::]. /0 fe e,t gj,, n p 4.3.3 Cavacity sP eh +fe 4.3.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more 2351-{ than % fuel assemblies. ".77..,m.+,4-- -+than ["^0] f ;l :: ^ elf;;; ai Le noreo in t h 4.2.2.2 nani O I O I k BWRN STS 4.0-3 -R:;. O, 09/29/92

m r i ] Design Features }l 4.0 k..

i i

s i i - g ; s -9 : g a r.e s La t L be. E qn L;a) by th< COL i d Ggt; cant, b This figure shall consist of [a' map of] the site area and' i provide, as a minimum, the information described in Section ~

O

[2.1.2] of the fSAR relating to [the map). - S i 1 -.1 } l t t 1 Figure 4.1-1-(page'Iof_1). Site and Exclusion Area Boundaries ABWRh STS 4.0-4 Rav. O, 0^/28/92- [

l r. i t Design Features - ( ): 4.0 - i i 1 1 4 i I q-h <. y S upp G' j gf;can4 I ha LL goL Mf re 3 5-his 4?; 3 tu "This figure shall consist of'[a map of] the site area showing the LPZ boundar .O and recreational' areas.y. Features such as towns, roads, shall be indicated ~in sufficient detail to allow identification of significant shifts-in: population distribution within the LPZ. i i Figure 4.1-2 (page 1 of.1) Low Population Zone MM i (kBWR/6 STS 4.0-5 -R... 0, 00/20/02-

p t p i Ni s ' 'S e cti bA is'fhvibh N_j'YY F C C L-7tM L, ca y /. .,.f ~ " :pe :ibil4ty 5.1 {- '{ 5.0 ADMINISTRATIVE CONTROLS 1 %. Responsibility [ { 4 5.1.1 The'[ Plant Superintendent] shall be responsible for overall unit ~ operation and shall delegate in writing _ the succe ion to this responsibility during his absence. Th [ Plant Superintendent], or his designee, ' accordance with app d administrative procedures, shall ap rove, prior to imple tation, each proposed test or exper ment and proposed changes d modifications to unit systems or equipment that-affect nuclear sa ty. f f 5.1.2 The[ShiftSupe isor (SS)] shall b responsible for the control- } room connand funct' n. ~ A manage t directive to this.effect,. signed by the.[highe " level of rporateorsite' management] .j shall be issued. annua to all station. personnel. During anv absenceofthe[SS).from he ntrol room while the. unit-is in~ MODE 1, 2, or 3, an indivi with a valid Senior Reactor -l Operator (SRO) license sha e designated to assume the control j . room connand function. D in ny' absence of thel [SS] from the control room while the it is 1 MODE 4 or 5, an individual'with.

O-e v iid sao iice or ctor one tar 14ce sheii:6-designated to' assume e control ro command function.

j 3 ~! i 1 L q l q lO A BWRhi-STS.' 5.0-1 L. .O, 00/20/12' a b .i

PAM Instrumsntation e B 3.3.3.1 BASES ( LCO Continued ) 4,-51 Drywell Pressure Drywell pressure is a Category I variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. -7we-Pour wide range drywell pressure signals are transmitted'from separate pressure transmitters and are continuously displayed 11n.1the" main cont'olTroom recorded and dispicyed n tU ::nt'rbl' rod r ' ' These displays recorder are the primary reed:de :.' indication used by'the" operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. t 5. 617;'Drywell/wetwell "rinary C -tai-rmt Area Radiation (Hioh Rance) r Pri= ry centairment cre

diction (higP rznge) i:.Drywell and.wetwe11' radiation measurements and displaystare provided

~ to' monitor ~for the potential'~of'aignificant~ radiation ~ releases and to provide release assessment for use by operators in determining the need to invoke, site emergency ^ plans.(Two divisions'~of instrumentation?are provided"with both drywell and wetwellimonitors in each division >andTate continuously displayed 11n the main control' room.oThese displays;are the; primary?indicationfused byithe operator during.an: accident..ThereforeifthelPAM Specificatientdeals l specificallyLwith:this' portion of theTinstrument/ channel. For thic pl nt, pri=:ry cent:ir.m:nt Orc ::dicti:n (high range; PAM instrumentation ecncist cf the following; f Dr ::11 Cum = Level Oryw;11 cump level i: -Category : variable provided for verificatien cf ECCO feneti n that Operate t: :irt ir nCE integrity. Tor thic plant, the dryvell cump 107:1 PAM in trrtentation conciet: Of the fOllowing: Trvu:11 Drair Cu - L val Dryuell dreir :: p ler:1 i: : C:tegory ! v riable previded t; detect bres:P cf th OCPO nd fer verificati;n and 1:ng ter-Ourveillan:: Of 5000 functi;n that Oper:t to m intain 300 int 0grity. For thic plant, the dryvell dreir cump level PLM in trument tion ;cncitt: Of the fcilcuing; 8. Primary Containment Isolation Valve (PCIV) Position PCIV position is provided for verification of containment integrity. In the case of PCIV position, the important I (continued) SEE/5 ABWR STS B 3.3-75 03/31/93 nev 0, 09/fS/92 )

1 i PAM Instrumsntation l B 3.3.3.1 1 BASES ( LCO Continued ) 8. Primary Containment Isolation Valve (PCIV) Position (continued) information is the status of the containment penetration. The LCO requires one position indicator for each active 7 t PCIV. This is sufficient to provide redundant [ indications of r rify redund:ntly the isolation status of each isolatable penetration via indicated status of the active valves ~and prior knowladge of passive valve or system boundary status. If a penetration is isolated, position indication for both of the PCIVs in the associated penetration flow path is not needed to determine status. l Therefore, the position indication for valves in an isolated penetration is not required to be OPERABLE. l Indication ~of'the completionof'the containment';;isolat' ion i function;is~providedlby valvetelosed/not: closed ~ indications for individual valves on safety;related video displays. Annunciators are:provided'toealert the operator'to'anyjlines that: may not-be. isolated. For this plant, the PCIV position PAM instrumentation consists of the following: 9,10. Wide Rance Neutron Flux Wide range neutron flux is a Category _I, variable,provided to verify reactor shutdown.zThe(displayLeontrollerfusesidata from two' APRM divisions ~and two: SRNM' divisions !tiofproVideTa display:of' neutron flux fon.:the mainLeontro1lroomi: panel:with a range:of:10 T to full power ( These.displaystare the primary indication.used-bycthe operator l:duringtan accident. 3 Therefore,1the PAM: Specification: deals specifical.ly;,with' this; portion.of;the'instrumentLchannel. ~ ' Tcr this pl:nt,.id reng; ncuteen flux PT.M in tru=cnt: tion consiet: Of the following: 44v-l l', 612. Drywell and Wetwbil CO,tci ment Hydrocen and oxvoen Analyzer Drywell and wetsell' cent in=ent hydrogen and oxygen [ analyzers are Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach. -4 hic veri:Me-+e-The se parameters _are also important in verifying the adequacy' of mitigating ' actions. There'are' two" divisions "of a H:/O analyzerse Samples?of.eitherjthefdrywell'or!wetwelllare..... i drawn ~ into' the _ analyzers -based: en tthe position of ta selector switch 'in the ' main control room.1 Displays : and ~ alarms'are providedli.n'the main; control room. These:. displays are the primary indication used by?the'operatoriduring,an? accident > Therefore,1the PAM' Specification dealskspecifica11yiwith' this' portion:ofLthe instrument 1 channel. (continued)__ BWRf6-ABWR STS B 3.3-76 03/31/93 nov 0, 09/2S/92

PAM Instrumsntation B 3.3.3.1 BASES ( LCO Continued ) My--11; :'12. Drvwell and Wetwell Cent inment Hydrocen and Oxvoen Analyzer (continued) Tcr thi; plent, the dryu 11 :nd centsin= nt hydr:g:n and cuyg:r :ncly:cr: PLM in trumcatsti:n consist; cf the fellering: 12. "rincrv Centair-nt "rcacur: Primary centainment pre: ur i: : C:tegory ! v:ri ble provided t: verify nOO nd cente+nment integrity :nd :: verify th Off;;tivents: cf 2000 ::ti:n: tchen t pr;v:nt centtint:nt 5::: 5 Tu .:id: : ng pri=:ry centninn:nt prc::ur cign :: ::: tr:n;=itted fr:r ::parct: prc::ure tran==itter: :nd cr; : ntinuou ly rc ;rd:s and di p1:y;d :n tu: : ntr:1 rocr re;;rder;. Th;;; :::;rder; cr: th pri= ty indic: tier u Od by th Operater during n :: ident. Ther fere, the PLM Cpecif! : tier 30:10 Opecifically wit-h this p;rtion of the in:trum:nt ch nn:1.

13. Suppression Pool Water Temperature Suppression pool water temperature is a Category I variable provided to detect a condition that could potentially lead to containment breach, and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression pool water temperature instrumentation allows operators to detect trends in suppression pool water temperature in sufficient time to take action to prevent steam quenching vibrat.i.ons in.the. suppression _po.ol.

There are'four divisions'of; suppression poo11 temperature monitoring;with twenty foui (5 ) temperature sensors in_'e'ach ~ division, s The temperatnrels'ensorstin' each ' division "are ' mounted' at1[15 )::.circumf rential positionsfwith : [y)":i sensors 'at various' elevations atieachy. position., crc' rranged is^5 x i group:ef feui l'ndependent":nd~r:dUndant ch:nnel:, The sensor 'are ' located (to! provide LanTindication"oflthe average pool temperature and withllbeit d;;;F thht'ther: 15 "a~~ group of 'serssore 'within ~ a' 9. metei-M-f+ line of sight of each relief valve discharge ~' location. Thus, et*-li f) groups of sensors are sufficient to monitor each relief' valve discharge location. TheLindividuaifsensors and bulk average ^ temperature-maytbeVselected:for displayfin the control' room. These.. display: h :Fgicup~cf fewe-eenesee id 1udd:'td5 sin;;r feb~narm:1~;uppr:::ica p :1 temp : turc monitoring :nd tu: ::n::r: f or r.*.M. Th cutput; f:r the PLM esseers-se: ::: rded on four ind:p:nd:nt rc crder; ir th: centrol reer. (Ch:nn:1:

nd C :r: redund:nt-t: ch:rnel: S and 0, re;pectively.;

• 11 four of th::: re;;rder: must b; OPEn?.0LC t o f urnich tu channels of PLM indic tion for :::h of the relic! v:1== di: h rg: 10:: tion:.

Th=== ::: rd:r: are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels. 1 (continued) Ern/E ABWR STS B 3.3-77 03/31/93 nev. O, 09/23/02 1 l I a i l

PAM Instrumsntation B'3.3.3.1 I i BASES ( LCO Continued ) i 4 i 14; Dr@well" Air"TesceratOFe i DrywellY airTempe ritureTi s7a } catei;idry?I Wsriabis @r6Vidsd? t o verify RCS^and containmentiintegrityfand;toiverifylthe ~ ~ ~ effectiveness ofiactions taken:toiremove'onergyffromaths containment."1There are two:divisionsfoffdrywelistemperatsife monitoring with temperature sensors! distributed!throughout~ 1 thejdrywellito: provide lconfidenceithatsthereii~efanfadequate i representation of thesstate of thefdryuellh controlfroom'~ displays:offthe temperaturesyare}thej;primaryfindication1"0 sed by,the operator!duringianiaccidenth JThereforelStbalPAM Specification"dealsispecific_ ally;withXthis}portionjfjthe instrument'. channel.

15. 7 Main" Steam ~Line"Radlation i

Main" steam:11neiradiktion")1sTaicategory Y variabl'e7provided l F to monitor; fuel integrity %:Radiationfin;the main ateamjline i tunnel <pfwhichsi'sLmeasuredibystheiprocess1 radiation"~' i s monitoring' system Miis"anEindicatorToficcolant' radiation'.' There 'are :four available:Edivisionstof' main 2 steam' tunnel c radiation monitoring 1with"contro19roomldisplayoffrom'ieach l t divisionA These displays'are the7primarylindicationbusediby the operatortduring an accident.11Therefore,lthejPAM" l _ i Specification! deals;r erEcallyiwith thisyportionfof?the i instrumentichannelt. APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1 and 2. These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low-therefore, PAM instrumentation is not required to be OPERABLE in these MODES. ACTIONS Note 1 has been added to the ACTIONS to exclude the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE while relying on the Actions i even though the Actions may eventually require plant l shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to diagnose an accident using alternate instruments and i methods, and the low probability of an event requiring these j instruments. A Note has also been provided to modify the ACTIONS related to PAM instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, 4 subsequent trains, subsystems, components, or variables expressed in the condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required l (continued) Ern/E ABWR STS B 3.3-78 03/31/93 ner. O, 00/22/92 i

PAM Instrumentation B 3.3.3.1 BASES ( ACTIONS Continued ) Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate inoperable functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function. b1 When a' fun'etion' has~ ens 'reqbfred ^dfsision'/ channel-ene->r mer runctien: have'che' : quired":henn:1^that~is" inoperable, the required inoperable division / channel must be restored to OPERABLE status within 31-4G~ days. ~Th: 20 dey Completion Time.is based on thelhigh~ reliability'of]theTremaining devices 'for monitoring "the cema*eing^ OPEPJ.EL: ~ the parameter operiting ~nperien:: and takes into account ch:nnel (or ir th: :::: ;f Function th t h: Only en: required Chennel, other non negulatory cuid: 1.07 in:t rue.:nt ch nnel: to meniter the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. L1 if : ch:nn:1 h:: not b: r :::te::i 1: C? r.2L

t;: in 20 d:y:, If the;requiredTactionsiand?associa'ted~ completion time-of? condition A:are?not~meth thl's Required' Action' specifie's~1nitiation'of~ action's in accordance with Specification 5.9.2.c, "Special Reports," which requires a written report, approvei by the [onsite review committee),

to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This Action is appropriate in lieu of a shutdown requirement since alternative Actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation. Ezl i When a' function hast.two; required"channeis/divisi6nsjthat;are INOPERABLE.then_one' channel / division :n ^ r~ic :~Tdn: tion: hri: tO : qui:Od channel: 'that'irc~in0perable (i.e., tt:0 ch:nnel: in per:ble ir th: ::me Tunctien;, :n ch:nnel i-the runction should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to (continued) BWRf6-AEWR STS B 3.3-79 03/31/93 nee. O, 00/22/92

PAM Instrumantation B 3.3.3.1 BASES ( ACTIONS Continued ) C1 (continued) 2 1 the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur. Condition C is modified by a Note that excludes _ hydrogen / oxygen monitor ch:nnel:7 divisions;"since theyTareEcovered by Condition;D. Condition'O'pr:9 ids:~ ~ epprsprict: n:qdired l'. tion: 'fe'r tu: insp rchic hydrogen monit:0 ch nnele. E:1 When two hydrogen / oxygen monitor disisions channel are inoperable, one division hyd :ger~d5hitsk"chenn:1 must be restored to OPERABLE status within 72 hours. The 72 hour i completion Time is reasonable, based on the backup capability of the Post Accident Sampling System to monitor the hydrogen concentration for evaluation of core damcge and to provide information for operator decisions. Also, it is unlikely that a LOCA that would cause core damage would occur during this time. L.1 This Required Action directs entry into the appropriate condition referenced in Table 3.3.3.1-1. The applicable condition referenced in the Table is Function dependent. If the required-Actions *and' associated:Completi~n TimesYfor o Conditions:C or D is not met for--a: Function:then E::E'ti== in ptrabid':hahn:1^h: n t i:F :n'y"ncquiFdd '.;ti n f an Condition C er 0,

Opplicable, nd th: :::::icted C pletion Ti== h:2 enpiredr Condition E is entered for that function 1 and. table'3.3.3 21-1'used? te chann:1 nd provide:

fec~ transferto the" appropriate ~ subsequent Condition. Esl For the PAH :jerity of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of condition C or D is not met, the plant must be placed in a MODE in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on cperating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems. L Since alternate means of monitoring reactor vessel water level and primary containment area radiation have been developed and tested, the Required Action is not to shut down the plant but rather to follow the directions of (continued) i b EME/5 ?.BWR STS B 3.3-80 03/31/93 nev. O, 09/29/02

PAM Instrumsntation B 3.3.3.1 1 BASES ( ACTIONS Continued ) l Gzl (continued) Specification 5.9.2.c. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. SURVEILLANCE The following SRs apply to each PAM instrumentation Function REQUIREMENTS in Table 3.3.3.1-1. SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross instrumentation failure has not occurred. A CHANNEL _ CHECK is a comparison of the parameter indicated on one instrumentation channel to a similar parameter on other ~ instrumentation channels. It is based on the assumption that'independentTdisplaysTof in:trum:nt chann:1; :: nit: ring the same parameter shosid read approximately the same value. Significant deviations between displ'ay: in trument hrnnele could be an indication of excessive ~ instrument drift or other* faults in one of the channels cr cf :::: thing Ob;'n ~ "ihrich:. A CHANNEL CHECK will detect gross channel cr; failure; thus,~it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high r dictica in:trum:ntet.cn chculd bc cmper d t; cimilar pl nt in:trum:nt; icested thr ughc;t the pl nt. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the match criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. Performance of the CHANNEL CHECK provides" confidence guar:nte : that undetected outright channe1~failh e^is~1imited to 31 days. The rrcquency of 21_dsy: i: 5 ::d upon Thdlhigh#relisbiliff ofLthe"devicesfusedito imp"lementitheiPAM functionsIprovides confidence pl:E.t'spehhEihg Experihu b"uith~r6g:id t: ~ h:nn:1 GPEn?.E!LITY and drift, uhich d:: nctrat : that failure of more than one channel of a given function in any 31 day interval is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the i required channels of this LCO. SR 3.3.3.1 2 A CHANNEL CALIBRATION is performed every refueling ~ interval? IS month =, er :ppr =i=:tely :t Or ry : fuell.ig. ~' CHANNEL" 2 (continued) SER/S ABWR STS B 3.3-81 03/31/93 ner. O, 09/29/92

PAM Instrumentation i B 3.3.3.1 l BASES ( SURVEILLANCE REQUIREMENTS Continued ) SR 3.3.3.1.2 (continued) CALIBRATION is a complete check of the instrument loop including the sensor. The test verifies that the channel responds to the measured parameter with the necessary range and accuracy. The Frequency is based on thellowLdrif tL of thefdevices used toLimplement the'Tunctions_ covered _'by this Loo. Note that calibration cf'these channels. overlaps;oriis encompassed by' calibrations: required byLother.LCOs that' address-some of.the same components: required.by(the.,'PAM displays. Ope : ting experiene: nd~ :ncietency :ith th: typical inductry refueling cycles. REFERENCES 1. Regulatory Guide 1.97, " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," [Date). 2. ABWR'SSAR, Section'15.2 2. [Plent pe ci f ic d cument e (e.g., FC7 ", "nC Regulatory Ouide 1.07, CED ictter;.] EE'n/5 AEWR STS B 3.3-B2 03/31/93 ne C, 09/29/92 4

Rsmote Shutdown Syctcm B 3.3.3.2 B 3.3 INSTRUMENTATION B 3.3.3.2 Remote Shutdown System BASES i BACKGROUND The Remote Shutdown System provides the control room operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the n ::ter Cer.!:01. tic.COnling (ECIC) Cycte= High Pressure" Core Flooder[ System, the safety / relief valves, and the Residual Heat' Removal Shutdown Cooling System can be ~ used to remove core decay heat and meet all safety

• famotie requirements. AdditionalTsystemsTassisting"inithe shutdownicapabilityLare portions:ofithe Nuclear. Boiler System,i the Reactor ' Building; Cooling: Water System,7the Reactor 1 Service. Water:Systemi(the Electrical-Power Distribution System, Sand the Flammability ControlLSistiem?

The'long' term supply of" water for the RGIG HPCF and the ~ ~ ability to operate shutdown cooling from outside the control room allow extended operation in MODE 3. In the event that the control room becomes inaccessible, the operators can establish control at the two remote shutdown panel and place and maintain the plant in~ MODE 3. Not all controls and no ::: ry trenefer cuitche are located at the two remote shutdown panels. 0:2: centrol :nd tr:n:fer suitche:.ill h::: to b; 6perated locally at th; zwitchgear, = ter centrol pen:10,

Other 1:01 :t:tiene.

The pl:nt avt matic:lly re: h : MOCS 2 f II uing : pl:nt chutdeu- :nd

n b; meietoined :fely in MOO 3 for en extended period Of t*mer The' postulated f conditiions [ssenmed [ tof exi'st?as^ the 1Malii Control # Room'becomes[ inaccessible,Ithelplant]:Lisjoperating initi~allyiat ~ or-. lessl than Mdesign' power. l: The; plant;Cis f not_...

experiencing' anyltransient: situationsh i Even; thoughjthe Lloss of offsite. power-is"consideredTunlikelyD the;remotefshutdown panels-are powered lfrom; Class;1E[ power;systemibuses5Itand II J so that' backup AC; power:would:be. automatically:suppliediby' thel plant dieseligenerator.;EManual controls offtholdiesel generator are alsojavailablei.locallyg ~ The" plant isTnotlexperie'ncing'any; accident"conditionsieuch T as a design basis-' accident solthaticomplete;contro1 Eof;E engineered safeguard; feature : system,s/from;outside;the main controlyroom_is.notfrequiredf ' A11TplantRpersonnelJis~ assumed to haveievaenatodithe"mais 7 contro17 room and the main contro1Lroom continues to be inaccessible:forsseveral hours.. Thetinitialfeventsthati _ causes the main controitroom to become7 inaccessible?asoumes the reactor operator can manually:scramithe' reactor before ~Ifithisi s:not}possiblo dthe leavingL the' mainj controliroom.f i (continued) cen/E rTG B 3.3-83 E:v. O, 09/2S/92 ABWR'STS B 3.3-83 Rev;.'s ; Bf; 03/31/93 ~ ~

m Remote Shutdown System. = 5 B 3.3.3.2 BASES ( BACKGROUND Continued ) l capabilitiyfsf a' backup"means'tF achieve seactics?reliict'ivitly 1 ~~ " ~ ~ I shutdown;is-available. Some ofsthe? existing system's used?forEnorma17 reactor ~ shutdown"; operations; aref also utilized)in?the(remote shutidown capability;to shutidown :the;reactordfrom outside1theTmain q n J controliroom? "The' functions needed forrremote shutdown" controllarefprovided;with:manualJtransferfdevices1which I override lcontrolsifrom;the main: control 4 room and-l transfer 1 the controls:tolthe remote shutdown / control'.y(contro11and process ! sensor l signals r are interrupted}by;;the ; transfer T^ devicesi at(thel hardwired, L analog f.loopN sensor l signalsGhich interfacetwith the remotehshutdownisystem'arearoutedffrom i the sensor,ithrough>theitransferfdeviceslon the? remote shutdown panels,LandLthen:tolthefaultiplexingtsystemlromote i multiplexing;unitsL(RMUs)/ fort ransmission':'tojthelmain" t control' room.'.'Similarly,':controlfsignalsifrom(theimain control: room are enacted:through'the:RMUs'to' remote shut,down transferldevices, land'thenLtoEthehinterfacing{; system equipment. 1 Actuation of:the:transferidevices interrEpteIthe a connection'to and from theTRMUafand transfers: control'tosthe remote; shutdown system.' : All^necessary power;supplyjcircuits are also; transferred-to'other;sourceh D Remote.~ shutdown contro1Eisinot possible<without<-actuationsofithe(transfes devices.cf 0peration t ofithef transfer; devices causes c an--(alarm in the mainicontroicroom..:..The:remoteishutdown control i panelsf are -located 1outside7the:maint control? roomO Accenss~ts j' this;pointcisladministrativelyfandlprocedur. ally; controlled; i The OPERABILITY of the Remote Shutdown System control and l instrumentation Functions ensures that there is sufficient i j information available on selected plant parameters to place and maintain the plant in MODE 3 should the control room become inaccessible. 1 APPLICABLE The Remote Shutdown System is required to provide equipment { SAFETY ANALYSES at appropriate locations outside the control room with a design capability to promptly shut down the reactor to j MODE 3, including the necessary instrumentation and controls, to maintain the plant in a safe condition in ) MODE 3. j i The criteria governing the design and the specific system j requirements of the Remote Shutdown System are located in 10 CFR 50, Appendix A, GDC 19 (Ref. 1). The Remote Shutdown System is considered an important contributor to reducing the risk of accidents; as such, it has been retained in the Technical Specifications (TS) as indicated in the NRC Policy Statement. l 4 LCO The Remote Shutdown System LCO provides the requirements for the OPERABILITY of the instrumentation and controls necessary to place and maintain the plant in MODE 3 from a (continued) i SEn/E ET& B 3.3-84 n:c. O, 07/29/?2 I AEWR-STS B 3.3-84 Rev4}Bg;03/31/93 i i a 4

I l 1 i Remote Shutdown System B 3.3.3.2 1 BASES ( LCO Continued ) i 1 i location other than the control room. The instrumentation i and controls typically required are listed in ) Table 3.3.3.2-1 in the accompanying LCO. v__ _u,_ .u. s.3<133 m. , es C. li,2I. Z Zl.._.!.. Z.~, ~ 2..E._1..El. u 2 2 '.< A.E. rE.2.E.i.c ' IE...I.I..I, I.'.!_17 : _2.... i 3 depend: upon the pl nt': licencing b: i: : d::: ibed ir th: "90 plant pecific C:fety Evaluation n per {CEn;. Cen:::ll, tu divicien: ::: quired to be OPIn.'2LE. j heu:ver, Only :n: Chinn1 per given ";ncti n i: : quir:d if 4 the plant h:: ju tified cu;P d::ign nd th; "nC CEP h:=

pted the justification.

i ._ The controls, instrumentation, and transfer switches are those required for: Reactor pressure vessel (RPV) pressure control; + Decay heat removal; RPV inventory control; s Flammability Control; ~ i Atmospheric-Control" Monitoring; and e Safety support systems for the above functions, including service water, component cooling water, and onsite power, including the diesel generators. I The Remote Shutdown System is OPERABLE if all instrument and control channels divisions needed to support the remote shutdown function are OPERABLE. In some cases, Table 3.3.3.2-1 may indicate that the required information or control capability is available from several alternate sources. In these cases, the Remote _ Shutdown System is OPERABLE as long as one channel division of any of the alternate information or control sources for each Function is OPERABLE. The Remote Shutdown System instruments and control circuits ( covered by this LCO do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instruments and control circuits will be OPERABLE if plant conditions require that the Remote Shutdown System be I placed in operation. I APPLICABILITY The Remote Shutdown System LCO is applicable in MODES 1, and 2. This is required so that the plant can be placed and maintained in MODE 3 for an_ extended period of time from a location other than the main control room. This LCO is not applicable in MODES 3, 4, and 5. In these MODES, the plant is already suberitical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore (continued) BWn/5 STE B 3.3-85 Rev. O, 09/29/92 ABWR STs B 3.3-85 Rev; [B,[03/31/93

Remote Shutdown System B 3.3.3.2 BASES ( APPLICABILITY Continued ) necessary instrument control Functions if ma'in control room instruments or control becomes unavailable. ' Consequently, the TS do not require OPERABILITY in MODES 3, 4, and 5. ACTIONS A Note is included that excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a plant shutdown. This exception is acceptable due to the low probacility of an i event requiring this system. Note 2 has been provided to modify the ACTIONS related to Remote Shutdown System Functions. Section 1.3, completion Times, specifies that once a Condition has been entered, subsequent treins, subsystems, components, or variables expressed in the condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the condition. However, the Required Actions for inoperable Remote Shutdown System Functions provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate condition entry for each inoperable Remote Shutdown System Function. h.r 1 Condition A addresses the situation where one or more required Functions of the Remote Shutdown System is inoperable. This includes any Function listed in Table 3.3.3.2-1, as well as the control and transfer switches. The Required Action is to restore the. Function (both divisions, if applicable) to OPERABLE status within 30 days. 4 The Completion Time is based on operating experience and the ~ low probability of an event that would require evacuation of the control room. Er1 If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within (12) hours. The allowed completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an ordt:ly manner and without challenging plant systems. (continued) i Ern/E STS B 3.3-86 E: c. O, 09/2S/92 ABWRLSTS B 3.3-86 RevhlB,j03/31/93 L s

Remote Shutdown Syotem B 3.3.3.2 BASES ( Continued ) i SURVEILLANCE SR 3.3.3.2.1 4 REQUIREMENTS Performance of the C"7J NEL DIVISION CHECK once every 31 days ensures that a gross failure'of instrumentation has not occurred. A GdAhMEL DIVISION CHECK is a comparison of the parameter indicated on one ehannel division to a similar parameter on other ch:nn:10 divisionsi ~It"is based on the assumption that instrument ch nh:10disi~sions monitoring the same parameter should read approximately"the same value. Significant deviations between the instrument ch:nnel: divisions could be an indication of excessive instrument drif t 'in 'one of the ch:nnele divisions or something even ~ more serious.. A C"72:NCL DIVISION" CHECK will detect gross channel oridivision failure)'thus, it is key to verifying the instrumentation continues to operate properly between each C"?' NEL DIVISION CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the ch:nn:1 division instrument uncertainties,. including indicatien'and~ readability. If a channel division is outside the match criteria, it may be an i indication that the sensor or the signal processing equipsent has drifted outside its limit. As specified in i the Surveillance, a CHA"NEL DIVISION CHECK is only required for those channel divisions'that'are normally energized. j Ferformance of a C"A!NEL DIVISION CHECK guarantees that undetected outright ch:nner' division failure is limited to i 31 days. The Frequency is based upon plant operating experience that demonstrates ch nnel division failure is rare. j SR 3.3.3.2.2 i SR L3.3.3.2.2 verifles'the' performanceof'the1 Remote Shutdown-System-display is'operableiand within:calibfation.~ The test.. verifies the displayLresponds, to measured: parameter values with the necessary; range,; accuracy,fand) response time.- The frequency :is based'. upo:Creli'abilityfansifses(oflt3 equipment ~ performance. i SR 3.3.3.2.73 1 SR 3.3.3.2.33 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended 'l function. This verification is performed from the remote shutdown panel and locally, as appropriate. This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 4 4 from the remote shutdown panel and the local control stations. However, this Surveillance is not required to be performed only during a plant outage. Operating experience (continued) d ? Etm/5 CTS B 3.3-87 " ^ ^ ' ' * ' ^ ' ABWR'STS B 3.3-87 Revd!Bf;03/31/93

R: mote Shutdown Syotcm B 3.3.3.2 BASES ( SURVEILLANCE REQUIREMENTS Continued ) demonstrates that Remote Shutdown System control ch:nn:10 division's usually, pass.the._ Surveillance when performed at th: 10 7th everyfrefuellng Frequency. SR 3.3.3.2.4 4 GHANNEL DIVISION CALIBRATION is a complete check of the instrument loopfand the sensor. The test verifies the ~ ch:nnel division responds to measured parameter values with the necessary"' range and accuracy. The 20 7.cnSh Refuelin' [ cycle Frequency is based upon operating expe'rience~gand is~ consistent with the typical industry procedures; cefu ling cycle. REFERENCES 1. 10 CFR 50, Appendix A, GDC 19. SM?/5 STE B 3.3-88 n: c. O, 09/09/92 ABWR'.STS B 3.3-88 Revf;;B((03/31/93 4

EOC-RPT Instrumsntation B 3.3.4.1 B 3.3 INSTRUMENTATION l B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation l 1 BASES i i BACF. GROUND The EOC-RPT instrumentation initiates a trip "of ? fourfof.'the ~ Reactor Internal ~ Pumps 1(RIP) recirculeti n' plap~ trip--tRPTT to reduce the' peak ^ reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal MCPR Safety Limits (SLs). The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure -Low, or Turbine Stop Valve (TSV)-Closure, Trip Oil Prc ur: Leu (TEV). The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity. The EOC-RPT Function <is included!in the-Recirculation Flos control'-(RFC); system.!The'RFC system is a. triple; redundant microprocessor based eystem with.the'-data.needed~by the ATWS-RPT Function acquired from-the Reactor' Protection ~~ System.(RPS) Trip Logic Unit-(TLU)~via the" plant multiplexing system..The' discrete trip' data'from^the four independent RPS TLUsLare received ~by;all three..of;the RFC controllersi ~ i The tripidata receivedIfrom[thiRPS TLU.is'a" composite which transmits a trip. state data value.to'the EOC-RPT function on eitherfa Turbine Stop1 Valve-Closure or1 Turbine control ~ Valve Fast Closure, Trip.cil. Pressure-Low'ecram initiation.:The' logic for-these signels.is described in~'the'RPS LCO'{LCO'~ 3.3.1.1). ~ Independent"RPT' signals'are' generated?in'all'three'RPC subsystems using 2/4 logic. RPT data'from allJthree"RFC subsystems are'transmittedjto~four of the. RIP. Adjustable Speed ~ Drives (ASD)Lvia: the multiplexing system.LThe ASDsluse 2/3 logic to. implement the. trip and include:antadjustable delay'on the trip, actuation signals tolthe;1oad~ interrupters; The COC-7r! in:trument tier :: hour ir nefer:n : : i:

==pri= d :* ::n ::: th:t d:tect initiction of cle:ur: ! th: :V:, :: f::t =2=:ur: cf the TCV:, ::= bin d.:itY rel:yer l=gi circuite, and f :t :: ting circuit hreshcr th t interrupt th: pruer fr:r the r eirculation purp st:r gener:t r (PC; ::t g:ncreter: t: :::5 ef the recirculction purp :terc. The chcrn 10 include ! treni; Oguipmene (e.g., trip unito) th t ::xper:: :::ured input sign:1 uith (continued) P"P/5 ?.BWR STS B 3.3-89 Rev?f B;'03/31/93 ner. O, C9/2S/92 7 1

EOC-RPT Instrumentation B 3.3.4.1 i BASES ( BACKGROUND Continued ) pre-::t:bliched ::tpcint. "hcr the : tpcint i: :::: d d, th: ch nnel cutput r01;y ::tuat::, which then cutput: := EOC "PT cignal to the trip logic. " hen th PPT hrc ker trip pen, th: ::irculation p;;p: ::::t deur und:r their cu-inerti:. The IOC "PT h:: tue identie:1 trip cyster, cith=r =! uhich ::n ::tu:L: :n PPT. Be:t LOC "PT trip cy;t:r i: 0 tu: cut of tu: Icgi for :::t Punction; thu=, either tu: TCV C10:ure, Trip C44 Prc :::: -- L:u :: tu: TCV P: t C13:ur, Trip Ci4 Preocur: L:n cign 1; cr required for ; trip cy;ter to If cithcr trip ;y;t:0 ::tust::, b;tt :::icculation purp:.:111 trip. There ::: tu IOC ""! brech::: in : rie i per :::ircul: tion purp. One trip cycter trip: One Of the tus COC-PPT brecher; for :::P recircul ticr pump :nd th

cnd trip ;yster trip; th: Oth r 500 "PT br: her fer :::h recircuinti:n purp.

APPLICANTS The Trv Clecure, Trip Oil Prc :gre Lou :nd the TCV P::S !?PETY EMALYCEO, C10 ure, Trip Oil Prec ure--Lou Punction cr: de igned t-e LCO, and trip the r;;ircui tion pump; in th: :c;nt of : turbin: LPPLICABILITY trip cr g:ncratcp lead ::j;; tion 3: mitiget: thc ncutron APPLICABLE The EOC-RPT offa.specifiedl number of RIPS.isTprovided.to SAFETY; ANALYSES' : mitigate-the neutron flux,' heat' flux'and pressur+ee LCO, and ' transients,'and'to increase the margin to the MCPR SLTfor APPLICABILITY ( events that'cause a' rapid' shutoff.of-th'eIsteam flow to the main turbine.'The' analytical' methods"and' assumptions"used in ~ evaluating ~the turbine trip and generator load rejection, as well as other safety analyses that assume EOC-RPT, are summarized in References 2, 3, and 4. To mitigate pressurization transient effects, the EOC-RPT must trip the RIPS retircui: tion pumps after initiation of initial closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than does a scram alone, resulting in an increased margin to the MCPR SL. Alternatively, MCPR limits for an inoperable ECC-RPT as specified in the COLR are sufficient to mitigate pressurization transient effects. The EOC-RPT function is automatically disabled when turbine first stage pressure is < [40%) RTP. f EOC-RPT instrumentation satisfies criterion 3 of the NRC Policy Statement. The OPERABILITY of the EOC-RPT is dependent on the OPERABIL!TY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in :::t trip ;yste=, with their trip eetpoints within the specified Allowable value givenmin?ths RPS LCO'(LCO'3.3.1.1) of CP 2.2.'.1.2. The actual setpoint is calibrated' consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated load ~1nterrspters1on'the ASDs. EOC "PT 2 brecher. Each channel ~(including the' associated ASDjload (continued) BWaf4-ABWR STS B 3.3-90 Rev'~ ;B, J 03/31/92 "Or. P, 09/2S/92

l EOC-RPT-Instrumentation l B 3.3.4~.1 P BASES (' APPLICABLE SAFETY ANALYSIS,-LCO,.and"APPLICABILITYE Continued.') interruptere EOC '"! brechere) must also respond within its assumed response time. 4 i The allowable. values;'applicab.9,*. %.-y1 analysis,?and, applicability of the Turbine Stcl 0J.w-Closure:and Turbine Control" Valve Fast closuren Trip 0110 Pressure-Low TOC-RPT l functions are' addressed in the RPS LCO?(LCO 3.3.1'1)4 The allowable values,: applicable. safety analysis, hand' applicability of1the.ASD' trip / initiation and41oad interrupters are addressed by;the ATWS-RPT LCO(-LCO.. i 3.3.4.2). This LCO addresses the.EOC-RPT function ofjtheTRFC controller. ,....y....._2 __u.. v ~,,._ m. m.....__.,__ ) specified ir th: LCC. 5:=in:1 trip ::tpcinte-cre :p ified

... ~ _._r_.....

.a..___,... ~... ___._a,_ ...y ...r... v,.. _ -. ___~..: r. .__.._3 . _ -.....:. u....r._.._ 2 r ____.__2 i __.r... ...._.._m_..,_ ..u._........ _. _ __:_ _...___ u_ 2_ __.~~ _.___2 u_ ,,,_..m,_ -____-r--_-.m..~. n.. e t. e.s t~. n n.s.e. 4 n. u..r.. -- __ --nr ee.:_-_ .t. .e.. e,. v. u_._.____ a....__. a...._ ...y ..y.... . ~....... ...y k_.._.. .. !. k..:. m.,.,._..x,_,,.,.. _.. _-_..,.u.i.... .~_ m. m .r..., y_e _.~___ __2 ._3...__ .._.: y_ .~r,.,..._., u..~____ ___ _.... ~.. ~., .y.. ..u...u .__._.:... -...u.~ 32 .m.,~_ .u.._ .~.r-.1_ _.... _____ c.,~..

==p=r:d t=

the ::tu:1 pre ::: par== ter : .g., TD' electrehydrauli: centr:1 (E"C) preocure), nd -h e r the __2 ~ .r u..~_ ..___2._ u...,_, y.-___ r...~_... .u_ _____:~.._2 ,2.~.. :.. _ e.,,., ...e _u_____ _.r_...u..u.._~~._._..,....: o ....t 7

3..:._-.:.

2__: ....._2 .-..._u..._ . _. l u e :ef the pr ces: por ettre ebtained fr - the : fety .u_.~ s.,.,_._a_.,_ 2__: .s __.. -..._2... _ ~ _ _ _ _ _ _ ~ .u_ _2 __.....,y._,__, ..u.. _ a.

==.': the inetrument errer=.

The trip : tpcint: ::: t h e.- r determined ::: cunt 4ng for th: ::=:ining instrument errece

2.._.:

.,.u.~_ __e_ ...._2 . u..s._ ______ ._1_ __.r_.... 2__: ...r ..... ~. 2. 2__.... c_._._-_ u. .... ~ _ _ y. _. . _ v. ,.rw_..:__~._,__

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... ~............... 7____ 2 _ric., ..;. ~... .. _.... ~ _..... s... t .,.~ .u.~...~.. u...... _ _ _ ~ _ _.. : m...__._.u~....:....______._ _2_,_...~4 . ~... .n

e..n.,. n,,,

._. a. m.. ,, n .....,.,...~ ,- u._ ____:,:.. c r. 1._.m..,. _.c,_.,.. .g App.. c....- e-e+e _. s..s.. -.~ __:., _.....___ a 2: _ _. ,.1 _ __2 u_,_.. __ _.. ~..,.. -, . ~. -.. ..____u.. .n.. _ _ : __ u._._..:. 1 Alternately, since this instrumentation protects against a MCPR SL violation with the instrumentation inoperable, modifications to the MCPR limits (LCO 3.2.2) may be applied to allow this LCO to be met. The MCPR penalty for the Conditicn EOC-RPT inoperable is specified in the COLR, 9 Turbine Ster '?:1ee CI cure, Tric 011 "r rcure Leu r.,.___._._..

m. u _.~..e.n.._ __2

....m.:._.._ ...y .._u.~. les: Of : h::t zink that produe:: reacter prc::ure, neutron flux, and h::t flun tr :icnt: th:t mu t be limited. _1_ hereL_._ _, _.. n n.........._.. a.- .r. u. n.:,. (continued) i m'"/E ASWR STS B 3.3-91 P,evj B,[ 03/31/93 ".0c. O, 09/2?/92 4 6 I l 4 ..}}