Text

Proposed Tech Specs Re Reactor Protection Sys (Scram) Instrumentation Requirements & Reactor Protection Sys (Scram) Min Calibr Frequencies for Reactor Protection Instrument Channels
	ML20092H632
Person / Time
Site:	Pilgrim
Issue date:	02/07/1992
From:	; BOSTON EDISON CO.
To:	;
Shared Package
ML20092H629	List: BECO-92-012, Application for Amend to License DPR-35.changing TS to Extend Reactor Protection Sys Instrumentation Surveillance Test Intervals from 1 Month to 3 Months to Provide for 12 & 6 H Allowable out-of-svc Times for Repairs & Tests; ML20092H632;
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{{#Wiki_filter:-. ATTACHMENT B TO BECO 92-012 1 J Revised Technical Specification Paggi 27 28 29 l 30 31 32 33 34 35 36 37 38 39 40 41 45 46 46a (added) 54 54a 58b 63 67 74 76 77 9202210230 920207 PDR _ADOCK 05000293 P PDR

1 PNPS Tablo 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION RE0JIREMENT j Operable Inst. Modes in Which Function i Channels Per Trip Function Trip Level Setting Must Be Ocerable Action (I) l Trio Sys':em (1) Refuel (7) Startup/ Hot Run i Minimum ' Avail. Standby I 1 1 Mode Swit, in Shutdown X X X A j 1 1 Manual Scram X X X A IRM 3 4 High Flux 1 20/125 of full scale X X (5) A 1 i 3 -4 Inoperative X X (5) A APRM 2 3 High Flux (15) (17) (17) X A or B 2 3 Inoperative (13) X X(9) X A or B 2 3 High Flux (151) 115% of Design Power X X (16) A or B l 2 2 High Reactor Pressure 11085 psig X(10) X X A 2 2 High Drywell Pressure 12.5 psig X(8) X(8) X A 2 2 Reactor Low Hater Level 29 In. Indicated Level X X X A SDIV High Hater Level: <39 Gallons X(2) X X A 2 2 East 2 2 West l l 2 2 Main Condenser Low {; Vacuum 223 In. Hg Vacuum X(3) XC3) X A or C l i 2. 2 Main Steam Line High 17X Normal Full Power l Radiation Background (18) X X XC78) A or C 4 4 Main Steam Line Isolation j Valve Closure 1 01 Valve Closure X(3)(6) X(3)(6) X(6) A or C 1 2 2 Turbine Control Valve 2 50 psig Control Oil 1 Fast Closure Pressure at l Acceleration Relay X(4) XC4) X(4) A or D I 4 4 Turbine Stop Valve Closure 110% Valve Closure X(4) X(4) X(4) A or D i i Amendment No. 75, 42, EE, 92, II7. I23, 27

9 NOTES FOR TABLE 3.1.1 1 1. There shall be two operabic or tripped trip systems for each trip function (e.g., high drywell pressure, reactor low water level, etc.). An l instrument channel, satisfying minimum operability requirements for a trip system, may be placed in an inoperable status for up to 5 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. An inoperable channel and/or trip system need not be placed in the tripped condition if this would cause a full scram to occur. When a trip system can be placed in the tripped condition without causing a full scram to occur, place the trip system with the most inoperable channels in the tripped condition, per the table below. If both systems have the same + number of inoperable channels, place either trip system in the tripped condition ser the table below. Condition Required Action _ Cogletion Ti n

a. Hith less than the Place associated trip 12 hours minimum required system in trip operable channels per trip function in one trip system.

or *

b. Hith less than the Place one trip system 6 hours minimum required in trip operable channels per trip function, in both trip systems.

or *

c. If full scram Restore RPS trip 1 hour capability is not capability available for a given trip function or *

Initiate the actions required by Table 3.1.1 and specified in Actions A through D below for that function:

A. Initiate insertion of operable rods and complete insertion of all operable rods within four (4) hours. l B. Reduce power level to IRH range and place mode switch in the startup/ hot standby position within eight (8) hourt. l C. Reduce turbine load and close main steam line isolation valves within eight (8) hours. l D. Reduce power to less than 45% of design. Amendment No. 6, 119, 28

tiQ11$ FOR TABLE 3._1.1 (Cont'd) I 2. PerQissible to bypass, eith control rod block, for reactor protection system reset in refuel and shutdown positions of the reactor mode switch. 3. Permissible to bypass when reactor pressure is (600 psig. 4. Permissible to bypass when turbine first stage pressure is less than 305 psig 5. IRH s are bypassed when APRM's are onscale and the reactor mode switch is in the run position. 6. The design permits closure of any two lines without a scram being initiated. 7. When the reactor is subtritical, fuel is in the reactor vessel and the reactor water temperature is less than 212'F, only the following trip functions need to be operable: L A. Mode switch in shutdown B. Manual scram C. High flux IkH D. Scram discharge volume high level E. APRH (15%) high flux scram 8. Not required to be operable when primary containment integrity is not required. 9. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MH(t).

10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

11. Deleted

12. Deleted

13. An APRH will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of LPRM's to an APRH.

14. Deleted

15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal power.

16. The APRM (15%) high flux scram is bypassed when in the run mode.

17. The APRH flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.

18. Hithin 24 hours prior to the planned start of hydrogen injection with the

. reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the in,jection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power. Amendment No. 6, 15, 27, 42, 26. 117, 118. 133, 29

m TABLE 4.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIO *iAL TEST FRE00ENCIES FOR SAFETY INSTRUMENTATION AND CONTROL CIRCUITS Functional Test Minimum Frecuency (3) Mode Switch in Shutdown Place Mode Switch in Shutdown Each Refueling Outage Manual Scram' Trip Channel and Alarm Every 3 Months 1 RPS Channel. Test Switch (5) Trip Channel and Alarm Once Per Week { l IRM I High Flux-Trip Channel and Alarm (4) Once Per Heek During Refueling 1 and Before Each Startup l 4 Incperative Trip Channel and Alarm Once Per Heek During Refueling 1 and Before Each Startup APRM i High Flux Trip Output Relays (4) Every 3 Months (7) Inoperative Trip Output Relays (4) Every 3 Months l Flow Bias Trip Output Relays (4) Every 3 Months l High Flux (151) Trip Output Relays (4) Once Per Week During Refueling ) and Before Each Startup High Reactor Pressure Trip Channel and Alarm (4) Every 3 Months I High Drywell Pressure Trip Channel and Alarm (4) Every 3 Months i Reactor low Hater Level Trip Channel and Alarm (4) Every 3 Months High Hater Level in Scram Discharge Tanks Trip Channel and Alarm (4) Every 3 Months 1 j Turbine Condenser Low Vacuum Trip Channel and Alarm (4) Every 3 Months i Main Steam Line High Radiation Trip Channel and Alarm (4) Every 3 Months i I Main Steam Line Isolation Valve Closure Trip Channel and Alarm Every 3 Months { l Turbine Control Valve Fast Closure Trip Channel and Alarm Every 3 Months 1 Turbine First Stage Pressure Permissive Trip Channel and Alarm (4) Every 3 Months l Turbine Stop Valve Closure Trip Channel and Alarm Every 3 Months-Reactor Pressure Permissive Trip Channel and Alarm (4) Every 3 Months Amendment No. 6, 79. 97. 117, 30

HOIti_f.0fLIMIL4dd 1, Deleted 2. Deleted l 3. Functional tests are not required when the systems are not required to be operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable status. 4. This instrumentation is exempted from the instrument channel test definition, This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels. 5. Test. CPS channel after maintenance. 6.- Deleted i 7. This APRM testing will be performed once every 3 months when in the RUN mode and within 24 hours af ter entering RUN mode, if not performed within the previous seven days. Amendment No. 6, 79, 31

TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FRE00ENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Calibratica Test (5) Minimum Frecuency (2) IRM High Flux Comparison to APRM on Controlled Note (4) l Shutdowns Full Calibration Once/ operating cycle APRM High Flux Once every 3 Days Output Signal Heat Balance Flow Bias Signal Calibrate Flow Comparator and Each Refueling Outage Flow Bias Network Calibrate Flow Blas Signal (1) Every 3 Months LPRM Signal TIP System Traverse Every 1000 Effective Full Power Hours Note (7) Note (7) High Reactor Pressure Note (7) Note (7) High Drywell Pressure Reactor Low Hater Level Note (7) Note (7) High Water Level in Scram Discharge Tanks Note (7) Note (7) Turbine Condenser Low Vacuum Note (7) Note (7) Main Steam Line Isolation Valve Closure Note (6) Note (6) Main Steam Line High Radiation Standard Current Source (3) Every 3 Months Turbine First Stage Pressure Permissive Note (7) Note (7) Turbine Control Valve Fast Closure Standard Pressure Source Every 3 Months Turbine Stop Valve Closure Note (6) Note (6) Reactor Pressure Permissive Note (7) Note (7) f 32 Amendment No. 6, 40, 79, 99,

ROTES FOR TABLE 4.L2 1. Adjust the flow bias trip reference, as necessary, to conform to a calibrated flow signal. 2. Calibration tests are not required when the systems are not required to be operable or are tripped. 3. The current source provides an instrument thannel aligt..er.' Calibration using a radiation source shall be made caca refueling out, p 4. Maximum frequency required is once per week. 5. Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle. 6. Physical inspection and actuation of these position switches will be performed during the refueling outages. 7. Calibration of these devices will be performed during refueling outages. To verify transmitter outpu;, a daily instrument check wiel be performed. Calibration of the associated analog trip units will be performed concurrent with functional testing as specified in Table 4.1.1. Amendment No. 79, 99, 33

BASES: t 3.1 The reactor protection system automatically initiates a reactor scram to: 1. Preserve the integrity of the fuel cladding. 2. Preserve the integrity of the reactor coolant system. 3. Minimize the energy which must be absorbed following a loss of-coolant accident, and prevents critichlity. This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels ma? ' out of service because of maintenance. When necessary, one channel nay be made inoperable for brief intervals to conduct required functional tests and talibrations. The reactor protection >ystem is of the dual channel type (Reference FSAR Section 7.2). The system is made up of two independent trip systems, each having two subchannels of tripping devices. Each subchannel has an input from at least one instrument channel which monitors a critical parameter. The outputs of the subchannels are combined in a 1 out of 2 logic (i.e., an input signal on either one or both of the subchannels will cause a trip system trip). The outputs of the trip systems are arranged so that a trip on both systems is required to produce a reactor scram. This system meets the intent of IEEE-279 for Nuclear Power Plant Protection Systems. The system has a reliability greater ihan that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system. With the exception of the Average Power Range Monitor (APRH) channels, the Intermediate Range Monitor (IRH) channels, the Main Steam Isolation l Valve closure, and the Turbine Stop Valve closure, each subchannel has one instrument channel. When the minimum condition for operation on the number of operable instrument channels per untripped protection trip system is met or if it cannot be met and the affected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved (i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor). Three APRM instrument channels are provided for each protection trip system. For some trip functions (e.g. MSIV or Turbine Stop Valve Position Switches).- the loss of one instrument may lead to degradation of both trip systams. In these cases, a 6 hour LCO must be entered. A source range monitor (SRM) system is also provided to supply additional neutron level information during refuel and startup (Reference FSAR Section 7.5.4). Amendment No. 79, 34

3.1 MSES (Cont'd) The requirement that the IRH's be inserted in the core when the APRH's read 2.5 indicated on the scale assures there is proper overlap in the neutron monitoring systcas and thus, sufficient coverage is provided for all ranges of reactor operation. The provision of an APRM scram at 115% design power in the Refuel and Startup/ Hot Standby modes and the backup IRH scram at 1120/125 of full scale assures there is proper overlap in the Neutron Monitoring Systems and thus, sufficient coverage is provided for all ranges of reactor operation. The APRH's-cover the Refuel and Startup/ Hot Standby modes with the APRM 15% scram, and the power range with the flow-biased rod block and scram. The IRH's provide additional protection in the Refuel and Startup/ Hot Standby modes. Thus, the IRH and APRM 15% scram are required in the Refuel and Startup/ Hot Standby modes. In the power range, the APRM system provides the required protection (Reference FSAR Section 7.5.7). Thus, the IRH system is not required in the Run mode. The high reactor pressure, high drywell pressure, reactor low water level, and scram discharge volume high level scrams are required for Startup/ Hot Standby and Run modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation. The requirement to have the scram functions, as indicated in Table 3.1.1, operable in the Refuel mode is to assure shif ting to the Refuel mode during reactor power operation does not diminish the capability of the reactor protection system. The turbine condenser low vacuum scram is only required during power operation and must be bypassed to startup the unit. Below 305 psig turbine first stage pressure (45% of rated), the scram signal due to turbine stop valve closure or fast closure of turbine control valves is bypassed because flux and pressure scram are adequate to protect the reactor. If the scram signal due to turbine stop valve closure or fast closure of turbine control valves is bypassed at lower powers, less conservative MCPR and MAPLHGR operating limits may be applied as specified in the CORE OPERATING LIMITS REPORT. Averaae Power Ranae Hopitor (APRM) APRH's #1 and #3 operate contacts in one subchannel and APRH's #2 and #3 operate contacts in the other subchannel. APRH's #4, #5, and #6 are arranged similarly in the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This allows the bypassing of one APRH per protection trip system for maintenance, testing, or calibration. Additional IRH channels have also been provided to allow for bypassing of one such channel. Amendment No. 79, 35

,1 _ 3i1 BASES (Cdnt'd) f The'APRM system, which is calibrated using heat balance data taken during steady-state conditions, reads-in percent of. design power (1998 MHt). r Because fission chambers provide the basic. input signals, the APRM system responds directly to average neutron flux. During transients, the- -instantaneous rate of heat-transfer from the fuel (reactor _ thermal power) is less than the-instantaneous neutron flux due to.the time constant of .the fuel.- Therefore, during abnormal-operational transients,-the thermal' power of the fuel will be less than that indicated by the neutron flux at the scram setting. ' Analyses demonstrated that with a 120 percent scram -trip setting, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin from fuel damage. Therefore, the use of flow-referenced scram trip provides even additional margin. .An' increase in-the APRM scram setting would decrease the margin present-before the fuel cladding integrity safety limit is reached. The APRM scram setting was determined by an analysis of margins required to provide.a reasonable range for maneuvering during operation. Reducing this operating _ margin would increase the frequency _of sp9d ous scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM setting was selected because it provides proper margin for-the fuel cladding integrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams. Analyses of the limiting transients show that no scram adjustment is required to assure the minimum critical power ratio (MCPR) is greater than _the safety limit HCPR when the transient is initiated from HCPR above the operating limit MCPR. For operation in the-startup mode while the reactor is at low pressure, 1the APRM scram setting of 15 percent of rated power provides proper thermal margin between the setpoint and the safety limit,-25 percent of rated; The. margin is sufficient to accommodate anticipated maneuvers associated with power plant startup. Effects of; increasing pressure at-zero or low void. content are minor, cold water from sources available during startup is not much-colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Horth.of individual rods is very. low in a uniform rod pattern. Thusi of all possible sources'of reactivity input, uniform control rod withdrawal is the most probable case of-significant power rise. Because the-flux-distribution associated with uniform rod withdrawals does not involve - high local peaks ~, and because several rods.must be moved to change power-by a significant percentage of rated power, the rate of power rise _is very slow.- Generally the heat flux is in the near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach.to the scram Amendment No._79, 133, 36

4 3.1 JLASIS (Cont'd) level, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before power could exceed the safety limit. The 15% APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 880 psig. The analysis to support operation at various power and flow relationships has considered operation with two recirculation pumps. Intermediate Ranae Monitor (IRM) The IRH system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRH is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram setting of 120/125 of full scale.is active in each range of the IRM. For example, if the instrument were on Range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on Range 5, the scram would be 120/125 of full scale on that range. Thus, as the IRH is ranged up tu accommodate the increase in power level, the scram setting is also ranged up. The most significant sources of-reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutran flux, and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded. In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed.- This analysis included starting the accident at various power levels. The.most severe case involves an initial condition in which the reactor is just subtritical and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak core power limited to one percent of rated power, thus maintaining MCPR above the safety limit MCPR. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM. Reactor low Hater Level The setpoint for low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results show that scram at this level properly protects the fuel and the pressure barrier, because MCPR Amendment No. 79, 133, 37

4 3.1 MSfJS (Cont'd) remains well above the safety limit MCPR in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 15 inches below the normal operating range and is thus sufficient to avoid spurious scrams. Turbine Stoo Valve Closure The turbine stop valve closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 1 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the safety limit MCPR even during the worst case transient that assumes the turbine bypass is closed. Turbine _ Control Valve Fast Closure The turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the acceleration relay. This setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR. Main Condenser low Vacuum To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low condenser vacuum initiates a scram. The low vacuum scram setpoint is selected to initiate a scram before the closure of the turbine stop valves is initiated. Main Steam Line Isolation Valve Closure The low pressure isolation of the main steam lines at 880 psig (as specified in Table 3.2. A) was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 785 psig requires the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram and APRM 15% scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram pro **-tion over the entire Amendme. No. 6, 79, 99, 123, 38

3.1 BASES (Cont'd) range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase. Hiah Reactor Pressure The high reactor pressure scram setting is chosen slightly above the maximum normal operating pressure to permit normal operation without spurious scram, yet provide a wide margin to the ASME Section III allowable reactor coolant system pressure (1250 psig, see Bases Section 3.6.D), Hiah Drywell Pressure Instrumentation for the drywell is provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the Core Standby Cooling Systems (CSCS) initiation to minimize the energy that must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation. Main Steam Line Hiah Radiat10D High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors that cause an isolation of the main condenser off-gas line. Rgactor Mode Switch The reactor mode switch actuates or bypasses the various scram functions appropriate to the particular plant operating status (Reference FSAR Section 7.2.3.9), i Manual Scram The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation. l l l Amendment No. 6, 79, 133, 39 l

C 3.1 BASES (Cont'd) Scram Discharae Instrument Volume The control rod drive scram syst?m is designed so that all of the water that is discharged from the reactor by a scram can be accommodated in the discharge piping. The two scram discharge volumes have a capacity of 48 gallons of water each and are at the low points of the scram discharge piping. During normal operation the scram discharge volume system is empty; however, should it fill with water, the water aischarged to the piping could not be accommodated which would result in slow scram times or partial control rod insertion. To preclude this occurrence, redundant and diverse level detection devices in the scram discharge instrument volumes have been provided. From a reference zero established by analysis, the instruments will alarm at a water level less than 4.5 gallons, initiate a control rod block before the 18 gallon water level, and scram the reactor before the water level reaches 39 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function properly. 4.1 BASES The reactor protection system is made up of two independent trip systems. There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Spacified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with General Electric Company Topical Report NEDC-30851P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System," as i approved by the NRC and documented in the safety evaluation report (NRC letter to T. A. Pickens from A. Thadani dated July 15, 1987). A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument channels have not been included in the latter table. These are: mode l switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is not applicable (i.e., the switch is either on or off). l The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. This is compensated for in the APRM system by calibrating every three days using heat balance data and by calibrating individual LPRM's every 1000 effective full power hours using TIP traverse data. Amendment No. 42, 122, 128, 40 1

,.g, This page is intentionally left blank, j e f Amendment No. 41

.i PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Operable Instrument' Channels Per Trio System (1) Minimum l Available Trio Function Trio Level Settina Action (2) 2(7) 2 Reactor low Hater Level 29" indicated level (3)' A and D 1 1 Reactor High Pressure 1110 psig D 2 2 Reactor Low-Low Hater Level at or above -49 in. A indicated level (4) 2 2 Reactor High Hater Level 148" indicated level (5) B 2(7) 2 High Drywell Pressure 12.5 psig A 2 2 High Radiation Main Steam-17 times normal rated o Line Tunnel (9) full power background 2 2 Low Pressure Main Steam Line 2880 psig (8) B 2(6) 2 High Flow Main Steam Line 1140% of rated steam flow B 2 2 Main Steam Line Tunnel ' Exhaust Duct High Temperature 1 70*F B 1 2 2 Turbine Basement Exhaust Duct High Temperature 1 50*F B 1 1 1 Reactor Cleanup System High Flow 13001 of rated flow C 2 2 Reactor Cleanup System High Temperature 1150*F C Amendment No. 26, 45

=. - NOTES FOR TABLE 3.2.A 1. Whenever Primary Containment integrity is required by Section 3,7, there shall be two operable or tripped trip systems for each function. An instrument channel may be placed in an inoperable status for up to 6 hours for' required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter; or, where only one channel exists per trip system, the other trip system shall be operable. 2. Action If the minimum number of operable instrument channels cannot be met for one of the trip systems, the appropriate conditions listed below shall be followed: If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within one hour (twelve hours for Reactor Low Water Level, High Orywell Pressure, and Main Steam Line High Radiation) or initiate the action required by Table 3.2.A for the affected trip functions. If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to operable status within two hours (six hours for Reactor Low Hater Level, High Drywell Pressure, and Main Steam Line High Radiation) or initiate the Action required by Table 3.2.A for the affected trip function. If the minimum number of operable instrument channels cannot be met for both trip systems, place at least one trip system (with the most inoperable channels) in the tripped condition within one hour or initiate the appropriate Action required by Table 3.2.A listed below for the affected trip function. A, Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours. B. Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours. C. Isolate Reactor Hater Cleanup System. D. Isolate Shutdown Cooling. i-i Amendment No. 86, 105, 179, 46

~ O 3. Instrument set point corresponds to 128.26 inches above top Of active fuel. 4. Instrument set point corresponds to 77.26 inches above top of attive fuel. 5. Not required in Run Mode (bypassed by Mode Switch). 6. Two required for each steam line. 7. These signals also start SBGTS and initiate secondary containment isolation. 8. Only required in Run Mode (interlocked with Mode Switch). 9. Within 24 hours prior to the planned start of hydrogen '.jection with the reactor power at greater than 20% rated power, the norr full power radiation background level and associated trip setpoir, may be changed ' based on a calculated value of the radiation level expi'ted during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power. Amendment No. 46a

PNPS TABLE 3.2.C-1 INSTRUMENTATION THAT INITIATES ROD BLOCKS Operable Instrument Channels Required Trio Function per Trio Function ODerational Conditions Notes Minimum i Available -l APRM Upscale (Flow Biased) 4 6 Run (1) APRM Upscale 4 6 Startup/ Refuel (1) APRM Inoperative 4 6 Run/Startup/ Refuel (1) APRM Downstale 4 6 Run (1) Rod Block Monitor 2 2 Run, with limiting control rod (2) (Power Dependent) pattern, and reactor power > LPSP (5) Rod. Block Monitor 2 2 Run, with limiting control rod (2) Inoperative pattern, and reactor power > LPSP (5) Rod Block Monitor 2 2 Run, with limiting control rod (2) Downscale pattern, and reactor power > LPSP (5) IRM Downscale 6 8 Startup/ Refuel, except trip is (1) bypassed when IRM is on its lowest range IRM Detector not in 6 8 Startup/ Refuel, trip is bypassed (1) Startup Position when mode switch is placed in run IRM Upscale 6 8 Startup/ Refuel (1) IRM Inoperative 6 8 Startup/ Refuel (1) SRM Detector not in 3 4 Startup/ Refuel, except trip is by-(1) Startup Position passed when SRM count rate is 2 100 counts /second or IRMs on Range 3 or above (4) SRM Downscale 3 4 Startup/ Refuel, except trip is by-(1) passed when IRMs on Range 3 or above (4) Amendment No. 15, 27, 42, 65, 72, 79, II0, 729, 128, 54 1

PNPS TABLE 3.2.C-1 (Con't) Operable Instrument Channels Required I Trio Function ' oer Trio Function Operational Conditions Notes Minimum i Available SRH Upscale 3 4 Startup/ Refuel, except trip is by-(1) passed when the IRM range switches are on Range 8 or above (4) SRM Inoperative 3' 4 -Startup/ Refuel, except trip is by-(1) passed when the IRH range switches are on Range 8 or above (4) Scram Discharge 2 2 Run/Startup/ Refuel (3) Instrument Volume Hater Level - High Scram Discharge 1 1 Run/Startup/ Refuel (3) Instrument Volume-Scram Trip Bypassed Recirculation Flow 2 2 Run (1) Converter - Upscale Recirculation Flow 2 2 Run (1) Converter - Inoperative Recirculation Flow 2 2 Run (1) Converter - Comparator Mismatch 4 Amendment No. 128, 54a

TABLE 3.2.F (Cont'd) l SURVEILLANCE INSTRUMENTATION Minimum # of Operable Instrument . Parameter and Rance Notes Type Indication 1 Channels Instrument # 1-RI 1001-609-Reactor Building Vent Indicator /Multipoint (4) (7) .I RR 1001-608 Recgrder 4 R/hr 10- to 10 L 1 RI 1001-608 Main Stack Vent Indicator /Multipoint (4) (7) RR 1001-608 Recyrder 4 R/hr 10- to 10 1 RI 1001-610 Turbine Building Vent. Indicator /Multipoint (4) (7) RR 1001-608 Retgrder 10- to 10'R/hr l i I Amendment No. 3, 58b 8

PNPS TABLE 4.2.C MINIMUM TEST AND CALIBRATION FREOUENCY FOR CONTROL R00 BLOCKS ACTUATION Instrument Channel' Instrument Functional Calibration Instrument Check _IfLLt APRM - Downscale Once/3 Months Once/3 Months-Once/ Day APRM - Upscale Once/3 Months. Once/3 Months Once/ Day APRM - Inoperative Once/3 Months Not Applicable Once/ Day IRM - Upscale (2) (3) Startup or Control Shutdown (2) IRM - Downscale (2) (3) Startup or Control Shutdown (2) IRM - Inoperative-(2) (3) Not Applicable (2)' RBM - Upscale Once/3 Months Once/6 Months Once/ Day RBH - Downstale Once/3 Months Once/6 Months Once/ Day RBM - Inoperative Once/3 Months Not Applicable Once/ Day SRM - Upscale (2) (3) Startup or Control Shutdown (2)' SRM - Inoperative (2) (3) Not Applicable (2) SRM - Detector Not in Startup Position (2)'(3) Not Applicable (2) SRM - Downscale (2) (3) Startup or Control Shutdown (2) IRM - Detector Not in Startup Position (2) (3) Not Applicable (2) Scram Discharge Instrument Volume Once/3 Months Refuel Not Applicable. Hater Level-High Scram Discharge Instrument Once/3 Months Not Applicable Not Applicable Volume-Scram Trip Bypassed Recirculation Flow Converter Not Applicable Once/ Cycle Once/ Day Recirculation Flow Converter-Upscale Once/3 Months Once/3 Months Once/ Day Recirculation Flow Converter-Inoperative-Once/3 Months Not Applicable Once/ Day Recirculation Flow Converter-Comparator Once/3 Months Once/3 Months Once/ Day Off Limits Recirculation Flow Process Instruments Not Applicable Once/ Cycle Once/ Day Loaic System Functional Test (4) (6) System Logic Check Once/18 Months Amendment No. 110, 120, 63

4 c NOTES FOR TABLES 4.2.A THROUGH 4.2.G 1. Initially once per month until exposure hours (H as defined on Figure 5 4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not less than one month nor more than three months. 2. Functional tests, calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped. Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations of IRMs and SRMs shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per day during those periods whea the instruments are required to be operable. 3. This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel. 4. Simulated automatic actuation shall be performed once each operating cycle. Where possible, all logic system functional tests will be performed using the test jacks. 5. Reactor low water level, high drywell pressure and main steam line high radiation are not included on Table 4.2.A since tney are tested on Tables 4.1.1 and 4.1.2. l 6. The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems. l l 7. Calibration of analog trip units will be performed concurrent with l functional testing. The functional test will consist of injecting a simulated electrical signal into the measurement channel. Calibration of associated analog transmitters will be performed each refueling outage. Amendment No. 67

18515: 4.2 The instrumentation listed in Table 4.2.A thru 4.2.H will be functionally tested and/or calibrated at regularly scheduled intervals. The same design reliability goal as the Reactor Protection System of 0.99999 is generally applied for all applications of_(1 out of 2) X (2) logic. Therefore, on-off sensors are tested once/3 months, and bi-stable trips associated with analog sensors and amplifiers are tested once/ week. Conservatively assuming that those instruments which have their contacts arranged in 1 out of n logic cannot be used during a testing sequence, there is an optimum test interval that should be maintained in order to maximize the reliability of a given channel (7). This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by: '+ Where: i-the optimum interval between tests. t-the ti N the trip contacts are disabled from performing their function while the test is in progress. r-the expected failure rate of the relays. To test-the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed this task requires an estimated 30 minutes to complete in a thorough a workmanlikemannerandthattherelayshaveafailurerateof10gd failures per hour. Using this data and the above operation, the optimum test interval is 1 x 103 hours i 2(0.5) 104_ - 40 days For additional marain a test interval of once Der month will be used initially, (7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16, 1968, page 10 Equation (24), Lawrence Radiation Laboratory. Amendment 74

4.2 f! ASIS (Cont'd) is shown by Curve No, 2. Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability. A more unusual case is that the testing is not done independently. If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3. Note that the minimum occurs at about 40,000 hours, much longer than for cases 1 and 2. Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel. Bypassing both channels for simultaneous testing should be avoided. The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately folicwing, the second channel be bypassed, tested and restored. This is shown by Curve No. 4. Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel. The best test procedure uf all those examined is to perfectly stagger the tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5. The difference between Cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in hu; nan error. The conclusions to be drawn are these: 1, A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and 2. more than one channel should not be bypassed for testing at any one time. The radiation monitors in the refueling area ventilation duct which initiate building isolation and Standby Gas Treatment operation are I arranged in two 1 out of 2 logic systems. The bases given above apply here also and were used to arrive at the functional testing frequency. Based on experience with instruments of similar design, a testing interval of-once every three months has been found adequate. l Amendment No. 89, 76

4.2 DASIS (Cont'd) The automatic pressure relief instrumentation can be considered to be a 1 out of.2 logic system and the discussion above applies also. The instrumentation which is required for the recirculation pump trip and alternate rod insertion systems incorporate analog transmitters. The transmitter calibration frequency is once per refueling outage, which is consistent with both the equipment capabilities and the requirements for similar equipment used at Pilgrim. The Trip Unit Calibration and Instrument functional Test is specified at monthly, which is the same-frequency specified for other similar protective devices. An instrument check is specified at once per day; this is considered to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip and alternate rod insertion systems are backups to existing protective instrumentation. Control Rod Block and PCIS instrumentation common to RPS instrumentation have surveillance intei.als and maintenance outage times selected in accordance with NEDC-308 SIP-A, Supplements 1 and 2 as approved by the NRC and documented in SERs (letters to D. N. Grace from C. E. Rossi -dated September 22, 1988 and January 6, 1989). A logic system functional test interval of 18 months was selected to minimize the frequency of safety system inoperability due to testing and to minimize the potential for inadvertent safety system trips and their attendant transients. Based on industry experience and BHR Standard Technical Specifications, an 18 month testing interval provides adequate assurance of operability for this equipment. Amendment No. 42, 727, 730, 77

_v e p ma es 4 +b4-,~-,-w,--4 a a-e-- 1,J,- hua A+nem+-- -a,42---o, he~ a 3 Inserts for Marked-up Pages Insert A on Paae 23 An inoperable r.hannel and/or trip system need not be placed in the tripped condition if this would cause a full scram to occur. When a trip system can be placed in the tripped condition without causing a full scram to occur, place the trip system with the most inoperable channels in the tripped condition, per the table below. If both systems have the same number of inoperable channels, place either trip system in the tripped condition, per the table below. Condition Reauired Action Comnletion Time

a. With less than the Place associated trip 12 hours minimum required system in trip operable channels per trip function in one trip system.

or *

b. With less than the Place one trip system 6 hours minimum required in trip operable channels per trip function, in both trip systems, or *

c. If full scram Restore RPS trip i hour capability is not capability available for a given trip function or *

Initiate the actions required by Table 3.1.1 and specified in Actions A through D below for that function:

Insert B on Paag_4A The reactor protection system is made up of two independent trip systems. There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with General Electric Company Topical Report NEDC-30851P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System," as approved by the NRC and documented in the safety evaluation report (NRC letter to T. A. Pickens from A. Thadani dated July 15, 1987).

Inserts for Marked-up Pages insert C on Page_46 2. Action If the minimum number of operable instrument channels cannot be met for one of the trip systems, the appropriate conditions listed below shall be followed: If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within one hour (twelve hours for Reactor low Hater Level, High Drywell Pressure, and Hain Steam Line High Radiation) or initiate the action required by Table 3.2.A for the affected trip functions. If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to ) operable status within two hours (six hours for Reactor Low Hater Level, High Drywell Pressure, and Hain Steam Line High Radiation) or initiate the Action required by Table 3.2. A for the affected trip function. If the minimum number of operable instrument channels cannot be met for both trip systems, place at least one trip system (with the most inoperable channels) in the tripped condition within one hour or initiate the appropriate Action required by Table 3.2.A listed below for the affected trip function. l

e .1 1 i ATTACHMENT C TO BECO 92-012 Marked-up Technical Specification Paagi l 27 28 29 30 31 32 33 34 35 36 37 38 39-40 41 45 46 54 54a-5Bb 63 67 74 76 -- 77 .-----.a

M A3 J ~ d NLE3.1.3 381TiTidiffWumlief ~~ CTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMENT Hodes in Hhich Function [OperableInst. Trip Function Trip Level Setting Must Be Ooerable Action (I) Channals per Refuel (7) Startup/ Hot .Run . Tri Ofl, Oyste Standby b ' Il Q Hode Switch in Shutdown X X X A 1 i / Hanual Scram X X X A IRH 3 Y High Flux 1 20/125 of full scale X X (5) A 1 3 et Inoperative X X (5) A APRM 2 3 High Flux (15) (17) (17) X A or B 2 3 Inoperative (13) X X(9) X A or B f2 J High Flux (15%) 1 5% of Design Power X X (16) A or B 1 2 2 High Reactor Pressure 11085 psig X(10) X X A 2l Z High Drywell Pressure 12.5 psig X(8) X(8) X A b 2 2 Reactor Low Hater Level 29 In. Indicated Level X X X A 2f 2 High Hater Level:in-Scram j 1f, Digarge-4nstr,-Volume {139 Gallons q Tl(2)j LXj 113 ty 1 2-2 wee r 2 2, Main Condenser Low i Vacuum 223 In. Hg Vacuum X(3) X(3) X A or C l 2 a Main Steam Line High 17X Normal Full Power Radiation Background (18) X X X(18) A or C 4 y Main Steam Line Isolation Valve Closure 1101 Valve Closure XC3)(6) XC3)(6) X(6) A or C 2 2_ Turbine Control Valve 1150 psig Control 011 1 Fast Closure Pressure at Acceleration Relay X(4) X(4) X(4) A or D hy Turbine Stop Valve Closure 1101 Valve Closure X(4) X(4) X(4) A or D RtHTis@t No. ~ 4v 15, 42, 86, 92, II7, AN, AmW men 27

3 -) m (e, cy., E 4 d-y "/// ~ * " *- *d ' O /** #* * # ** *'e ~

M/

H NOTES FOR TABLE 3.1.ll t i There shall be two operable or tripped trip-systems for each functior[An

l 1.

instrument; channel aay be placed in an-inoperable status for'up to 6 iours-l' f hh, f y. or required surviiillance without placing the trip system in the-tripped l S,,,,,, m - condition provided at least~one-OPERABLE channel in the same trip system of eg.4+7 is monitoring that parameter. +f-4heatnimum-number-of-openble-a-

<e"* ~ 6-ins t rumen t--c hanne l s-pe r-t r i p-sys t em a a nnot-be-me t - for-bo t h-t ri p -sy s t ems &

4 -

f-/ if'% the-appropriate-lons-li sted-belorshaltbe-takent ^--

s

H-lu s e t-We A.

Initiate in ion of-operable rods and complete insertion of all operable. rods within fourehours. B.. Reduce power level to IRH ran e and place mode switch in the startup/ hot standby position within r3 o q_5) C. Reduce turbine load and close main steam line isolation valves within f ' ( 8')hou rs.- W"r O D. Reduce power to less than 451,of design. V. I)ermissible-to bypass, with control rod block, for reactor protection 2 system reset in refuel and shutdown positions of the reactor mode-switch, p3. Permissible to bypass when reactor pressure is (600 psig. ]4'.:Permissibletobypasswhenturbinefirststagepressureislessthan305 psig. 1 5. IRH's are bypassed when APRM's are onscale and the reactor mode switch is in the run position. -6. The design permits closure of any two lines without a scram being initiated. 7.- When the reactor is subcritical, fuel is in-the _ reactor vessel and the reactor water temperature is less than 212'F, only the following trip -functions need to be operable: A. -Mode switch in shutdown B. : Manual scram I } C. - High flux IRM f f D.- Scram discharge volume high level l l E. APRM (15%) high flux scram [

8..Not required-to be operable when primary containment integrity is not j-required.

9. Not required while performing low power physics tests at atmospheric I pressure during or after refueling at power levels not to exceed 5 MW (t). / ^"~ I ~ Nw h mt W m,a 22 n Amendment'No 9', 28 L

4 NOTES FOR TABLLLld. (LAM 7'@ y-- " a t <t n - p~ ' PH ' ^ ~ ' ' 10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel. 11. Deleted 12. Deleted 13. An APRH will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of LPRH's to an APRH. 14. Deleted 15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS ;EPORT, but shall in no case exceed 120% of rated thermal power. 16. The APRH (15%) high flux scram is bypassed when in the run mode. 17. The ALRH flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes. 18. Hithin 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on'either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 201 rated power. l l Revtriorr-1464 p Amendment No. 15,27,42,86,177, IIB,A23; 29

^d ' # " "

N

~ TABLE 4.1.1 REACTOR PROTECTICJ SYSTEM (SCRAM) INSTRUMENTATI FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREOUENCIES FOR SAFETY INSTR _ AND CONTROL CIRCUITS Group-(2)C:^- Functional Test-Minimum Frecuency (3) Mode Switch in Shutdown A' Place Mode Switch in Shutdown Each Refueling Outage I Manual Scram A Trip Channel and Alarm Every 3 Months c,. e < Lbcef { RPS Channel Test Switch (5) Trip Channel and Alarm -Each-Re eHng-Outam IRM ,/ High Flux C Trip Channel and Alarm (4) Once Per Week During Refueling and Before Each Startup Inoperative C Trip Channel and Alarm Once Per Heek During Refueling.- and Before Each Startup High Flux B j Trip Output Relays (4) Once/ Week (7) w.f J A-A Inoperative B l Trip Output Relays (4) Once/Heek Flow Bias B l +4M4brate-F4ew-Blas-SignaF Once/ Month -41 7 ' f T~/_ Trip OntnjJt Re_ lays (4) Once Per Heek During Refueling High Flux (157.) B

-f/'# N+t >

and Before Each Startup High Reactor Pressure D rip hannel and ATK (4) g u,,4 44 }a-High Drywell Pressure .D / Trip Channel and Alarm (4) u -41 h Reactor Low Hater Level D Trip Channel and Alarm (4) n -414ca_ High Hater Level in Scram Discharge Tanks D Trip Channel and Alarm (4) Every 3 Months Turbine Condenser Low Vacuum D Trip Channel and Alarm (4) 44 % Main Steam Line High Radiation B Trip Channel and Alarm (4) -Once/Heek" Main Steam Line Isolation Valve Closure IA Trip Channel and Alarm 6-7 3 M'-MM-Turbine Control Valve Fast Closure A Trip Channel and Alarm -{+)* Turbine First Stage Pressure Permissive Df Trip Channel and Alarm (4) Every 3 Months Turbine Stop Valve Closure A Trip Channel and Alarm /< 41 Reactor Pressure Permissive Trip Channel and Alarm (4) Every 3 Months -Redslon-4M-AmendmentNo.p79,99,AI{ 30

1 HQJ1S FOR tab 1L4dd . f) /rfe d. 1. ([Initiallyoncepermonthuntilexposure(Hasdefinedonfigure4.1.1)is (P ' L2.0 x 10d; thereafter, according to figure 4.1.1 with an interval not jlessthanonemonthnormorethanthreemonths. The compilation of f instrument failure rate data may include data obtained from other boiling / L water reactors for which the same design instrument operates in an \\ environment similar_tO. thaLof PNPS. / 2. A-description of-the-four-groups-is-included-in-the~ Bases-of-thite l Specificationce 0, / c t, d. 3. Functional tests are not required when the systems are not required to be operable or are tripped. If tests.re missed, they shall be performed prior to returning the systems to an oph able status. 4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels. 5. Test RPS channel after maintenance. [jy 4 y,,,s , ~.- - - - ~ ~ - ~ - - 6. The water level in the reactor vessel will be perturbed and them,N ' corresponding level indicator changes will be monitored. This j Nrturbation test will c performed every month af ter completion of the j monthly functional test progru. 7. Thi*APRMteutingwillbeperformedonceperweek#whenintheJT8modef C if- , reactor-15-out-of-the-run-mode-for-more-than one~veck,-the testing ' wl-r7erformedas soor. as practicable-after-returning-to-the-run-mode o GG~[US (oD7'f T.,. n t4],, a4 4 osa > F H. n d

y'.

/ p e h , cf w a f4, 14e gevs o n.: sev en Amendment No. 7,9', 31

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w C w mm m D D C O O m O O eqt w t-6L ~- SU ' s, CE a w g u bs e 50 = o a U E E "} c 8 w*' t d snas&e c M> O O E se O V M e D e -(( - 2 b

s. aE

.8 s c m a c O h N g I b D D Y W z a u u = a e: Y 55555E E 'O 55 c$ fs R &* 3 4 3 3 t t 4 c a c "m 2 c a ?, "5 O t 2 e ? c u .E sus M 0 .c m -p s u ~ r 5 0 h E 0 ?N ..u 4 o E gg-5 .c x v .c o a e o o o u .e. a a. 2 3'.s b h .b c s.- a z z cx x x x t-r -= cp

NOTES FOR TAELE 4.1.2 1. A-description-of-four; groups-is-included in-the bases.of thisS1_-- , cc < h Spectf1cationna (P4 at HET % hn H:7 W Ti~iW Cu ~n*n h o 70 Q, a c - /.f - + t J th w.g. 2. Calibration tests are_n_ot required wheti'1.he systems are not required to be opeiable or are tripped. t 3. The current source provides an instrument channel alignment. Calibration using a radiation source shall be made each refueling outage. 4. Maximum frequency required is once per week. 5. Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle. 6. Physical inspection and actuation of these position switches will be performed during the refueling outages. 7. Calibration of these devices will be performed during refueling outages. To verify transmitter output, a daily instrument check will be performed. Calibration of the associated analog trip units will be performed concurrent with functional testing as specified in Table 4.1.1. l l l Revision-103"%. Nlt) % 33 pma,g,,,,7 l L

l BASfS: BASES: 'N 3.1 The reactor protection system 4.1 A. The minimum functional testing autcmatically initiates a reactor frequency used in this scram to.: i specification is based on a reliability analytts using the \\

1. Prs:e/ a a integrity of the concepts developed in refeJence j

fuel '). ng. (6). This concept was j specifically adapted to the one .2. Preser>e the Integrity of the out of two taken twice loglC of reactor ccolant system. the reattor protection systefD. The analysis shows that the

3. Minimize the energy which must sensors are primarily be absorbed following a loss responsible for the reliability of coolant accident, and of the reactor protection prevents criticality, system.

This analysis mates use of " unsafe failure" rate This specification provides the emperience at conventional and limiting conditions for operation nuclear power plants in a necessary to preserve the ability reliability tredel for the of the system to tolerate single system. An " unsafe failure" is failures and still perform its defined as one which negates intended function even during channel operability and which, periods when instrument channels due to its nature, is revealed ( may be out of service because of only when the channel is maintenance. When necessary, one functionally tested or attempts channel may be made inoperable to respond to a real signal, for brief intervals to conduct failures such as blown fuses, required functional tests and ruptured bourdon tubes, faulted calibrations. amplifiers. faulted cables, Mn# etc. which result in " upscale" The reactor protection syst,em s or "downstale* readings on the of the dual cha 1 type n Re A reactor instrumentation are (SectionTTG The sys'em s " safe" and will be easily made up of two ependent trip recognized by the operators systems, each having two during operation because they subchannels of tripping devices. are revealed by an alarm or a Each subchannel has an input from

scram, at least one instrument channel which monitors a critical The channels listed in Tables parameter.

4.1.1 and 4.1.2 are divided into four groups for l The outputs of the subchannels functional testing. These are: are ccobined in a 1 out of 2 logict(1.e., an input signal on A. On-Off sensors that provide eitbe'r one or both of the a scram trip function. subchannels will cause a trip systemtrip) The outputs of the B. Analog devices coupled with trip systems are arranged so that bi-stable trips that p a trip on both systems is required ( provide a scram functio. AmendmentNo.Jk 34

h 3.1 BASfS (Cont'lD A.I BASEf iCont'd h ( to produce-a reactor scram. C. Devices which only serve a \\ useful function during some \\ / restricted made of operation, \\/ This system meets the intent of i IEEE - 279 for Nuclear Power such as startup or shutdown, id" Plant Protection Systems. The or for which the only system has a reliability greater practical test is one that than that of a 2 out of 3 system can be performed at shutdown, and somewhat less than that of a 1 out 2 system. D. Diverse Analog Transmitter / trip unit devices that provide With the exception of the Average alarms, trips or scram functions. Power Range Monttor (APRM) channels, the Intermediate Range The sensors that make up group h Honttor (IRM) channels, the Majn (A) are specifically selected h i from among the whole family of E SteamIsolationValveclosufe;af,d the Turbine Stop Valve closbre industrial on-off sensors that r each subchannel has one have earned an excellent l instrument channel. When the reputation for reliable 4 minimum condition for operation operation. During design, a goal L on the number of operable { of 0.99999 probability of success f Instrument channels per Sntripped 1 (at the 50% confidence level) was { protection trip system 15 t.;et or I adopted to assure that a balanced -if it cannot be met and the and adequate design is achieved. -i m affected protection trip system is The probability of success is i i placed in a tripped condition, j primarily a function of the 7 lv the effectiveness of the f sensor failure rate and the test 7 protection system is preserved; interval. A three-month test j

/* *^kr l'

1.e., the system can tolerate a interval was planned for group b ,, w single failure and still perform (A) sensors. This-is in keeping its intended function of with good operating practices, M*** $ cramming the reactor. Three and satisfies the design goal for i_ .l 8 *

APRH instrument channels are the logic configuration utilized provided_for each protection trip in the Reactor Protection System.

system. . N N PRW s #)/ r % f L J m (J N /t)and #3 operate confait's' objective of maintaining an W A.w~ To satisfy the long-term in one subchannel and APRM's #2 adequate level of safety and #3 operate contacts in the throughout the plant lifetime, a other subchannel. APRM's#4,#5Q minimum goal of 0.9999 at-the 95% i and #6 are arranged similarly l'n confidence level is proposed. the'other protection trip Hith the (1 out of 2) X (2) system. Each protection trip . logic, this requires that each L system has one more APRH than is sensor have an availabl11ty of necessary to meet the minimum . 0.993 at the 95% confidence l number required per channel. level. This level of A This allows.the bypassing of one / availability may be maintained by APRM per protection trip system j adjusting the test interval as a-for maintenance, testing or l function of the observed failure calibration. Additional IRM / history (6). channels have also, /T~~ ~ ~ H T Reliability of Engineered Safety j 6 f.,, t,.f 6 f,..,(c 9. Assiv.

l Features as a Function of lesting

~ L o, e S t.f M A< A n b,,. 44,,j o, Frequency, I.M. Jacobs, " Nuclear j \\{ Safety." Vol. 9, No. 4 July-Aug. / /ou,Fo** i i. ~.. f.., 4.J e t 1968, pp. 310-312. l s,., a s. rm / h/4 n.fsp% ah,.s)y

r., /, nm ym 44%W

,/' ..a -~~ JtendmeEUfoJ9f 35

e e.--*.-_ l 3.1 RSLS (Cont'd) 4.1 DAS15 (Cont'd) N been provided to allow for 10 facilitate the implementation i bypassing of one such channel, of this technique, figure 4.1.1 is provided to indicate an 424-{> appropriate trend in test The 4verage-power-f ange9 interval. The procedure is as follows: wmonitorlog4APRM6sy5 tem, which is calibrated using heat balance 1. Like sensors are pooled into data taken during steady-state one group for the purpose of conditions, reads in percent of data acquisition. design power (1998 MHt). Because j fission chambers provide the 2. The factor H is the exposure basic input signals, the APRM hours and is equal to the system responds directly to number of sensors in a j average neutron flux. During group, n, times the elapsed transients, the instantancous time 1 (H ni). rate of heat transfer from the fuel (reactor thermal pcmer) is 3. The accumulated number of less than the instantaneous unsafe failures is plotted neutron flur due to the time as an ordinate against H as constant of the fuel. 1herefore, an abscissa on figure 4.1.1. during abnormal operational transients, the thermal power of 4. After a trend is the fuel will be less than that established, the appropriate indicated by the neutron flux at monthly test interval to the scram setting. Analyses satisfy the goal will be the demonstrated that with a 120 test interval to the left of percent scram trip setting, none the plotted points. of the abnormal operational transients analyzed violate the 5. A test interval of one month fuel safety limit and there is a will be used initially until substantial margin from fuel a trend is established. dama e. Therefore, the use of flo eferenced scram trip Group (B) devices utilize an prov es even additional margin. analog sensor followed by an j amplifier and a bi-stable trip An increase in the APRH scram circuit. The sensor and setting would decrease the margin amplifier are active components present before the fuel cladding and a failure is almost always integrity safety limit is accompanied by an alarm and an reached. The APRH scram setting indication of the source of was determined by an analysis of trouble. In the event of margins required to provide a failure, repair or substitution reasonable range for maneuvering can start immediately. An during operation. Reducing this "as-is" failure is one that operating margin would increase " sticks" mid-scale and is not the frequency of spurious scrams, capable of going either up or which have an adverse effect on down in response to an reactor safety because of the out-of-limits input. This type j resulting thermal stresses. offailureforanalogdevicesijs Jev444w-M6-P AmendmentNo.77,[//, 36

3.1 LASLS (Cont'd) 4.1 Mili (Coned) 1hus, the APRM setting was a rare occurrence and is selected because it provides detectable by an operator who _@/ v edM wa W margin for the fuel observes that one signal does not m, cladding-integrity safety limit tract the other three, for yet allows operating margin that purpose of analysis, it is reduces the possibility of assurned that this rare failure unnecessar m s will be detected within two hours. w.am wK<4m &&;- p Afralyses of the limiting The bi-stable trip circuit which transients show that no scram is a part of the Group (B) tment is requirtt[afetytoJt11m @ devices can sustain unsafe j jP Feat F {h~a'n Me failures which are revealed only m HCPR when the transient is on test. Therefore, it is nitiated from HCPR above the necessary to test them operating limit HOPR. periodically. For operation in the startup mode A study was conducted of the while the reactor is at low instrumentaticn channels included pressure. the APRM scram settD 9 in the Group (B) devices to of 15 percent of Jat.tLpmxQ~/- calculate their " unsafe" failure providesedequahthermalmargin rates. The analog devices between the setpoint and the (sensors and ampitfiers) are safety limit, 25 percent offff E h, 8 predicted to have an unsafe rated. The margin is adequa W To fa1{'Urerateoflessthan20X 10' failure / hour. The bi-stable accommodate anticipated maneuvers associated with power plant trip circuits are predicted to startup. Effects of increasing haveanunsafefa{1urerateof pressure at zero or low void less than 2 X 10' content are minor, cold water failures / hour. Considering the from sources available during two hour monitoring interval for startup is not much colder than the analog devices as assumed that already in the system, above, and a weekly test interval temperature coefficients are for the bi-stable trip circuits. -small, and control rod patterns. the design reliability goal of are constrained to be uniform by 0.99999 is attained with ample operating procedures backed up by

margin, the rod worth minimizer.

The bi-stable devices are Horth of individual rods is very monitored during plant operation low in a uniform rod pattern, to record their failure history Thus, of all possible sources of and establish a test interval reactivity input, uniform control using the_ curve of_ Figure 4.1.1. rod withdrawal is the most There are numerous identical probable case of significant bi-stable devices used throughout power rise. Because the flux the plant's instrumentation distribution associated with system. Therefore, significant uniform rod withdrawals does not data on the failure rates for the involve high local peaks, and bi-stable devices should be because several rods must be accumulated rapidly, moved to change power by a significant percentage of rated C l AsA+sss-H W Amendment No. M. ///, M l

3.1 ff_iLS (Cont'd) [4.1 M515 (Cont'd) power, the rate of power rise is { 1he frequency of calibration of very slow. Generally the heat the APRH flow Biasing Network has flux is in the near equilibrium ) been established as each with the fission rate. In an refueling outage. The flow assumed uniform rod withdrawal biasing network is furictionally { approach to the scram level, the tested at least once per month rate of power rise is no more and, in addition, cross than five percent of rated power calibration checks of the flow per minute, and the APRM system input to the flow biasing network would be more than adequate to can be made during the functional assure a scram before power could test by direct meter reading, exceed the safety limit. The 15% There are several instruments APRM scram remains active until which must be calibrated and it the mode switch is placed in the will take several days to perform RUN position. This switch occurs the calibration of the entire when reactor pressure is greater network. While the calibration than 880 psig. is being performed, a zero flow signal will be sent to half of The analysis to support operation the APRM's resulting in a half at various power and flow scram and rod block condition, relationships has considered thus, if the calibration were operation with two recirculation performed during operation, finx mps. shaping would not be possible. WM % Am.WO { Based on experience at other generating stations, drift of instruments, such as those in the The IRH system consists of 8 Flow Biasing Network, is not chambers, 4 in each of the avoid spurious scrams, a Significant and therefore, to reactor protection system logic channels. The IRH is a 5-decade calibration frequency of each instrument which covers the range refueling outage is established, of power level between that covered by the SRM and the APRM. Group (C) devices are active only The 5 decades are covered by the during a given portion of the IRM by means of a range switch operational cycle. For example, and the 5 decades are broken down the IRH is active during startup into 10 ranges, each being and inactive during full-power one-half of a decade in size. operation. Thus, the only test that is meaningful is the one The IRH scram setting of 120/125 performed just prior to shutdown of full scale is active in each or startup; i.e., the tests that range of the IRM. For ex le, are performed just prior to use if the instrument were on ge of the instrument. 1, the scram setting woul ea 120/125 of full scale for that Group (D) devices, whP e similar range; likewise, i.the in description to those in Group instrument were o jnge5,the (B), are different in use because ~ scram would be 12 55 of full they (the analog transmitter / trip g scale on that range. Thus, as unit devices) provide alarms, the IRH'is ranged up to k trips or scram functions, An accommodate the increase in power availability analysis is detailed ( in NEDD-2161?A (12/78). j -Revh4en-446 4 Amendment No 79,/d 3B g Y, %

_~ h1 (LASIS (Cont'd) 3.1 M515 (Cont'd) level, the scram setting is also Surveillance frequencies for the ranged up. The most significant SDV system instrumentation is sources of rtactivity change detailed in Amendment Number 65. during the power increase are due NRC concurrence with this to control rod withdrawal. For surveillance program is contained in-sequence control rod in the Safety Evaluation Report withdrawal, the rate of change of and its associated Technical power is slow enough due to the Evaluation Report (1ER-C-5506-66) physical limitation of dated 11/10/82. withdrawing control rods that heat flux is in equilibrium with Calibration frequency of the the neutron flur, and an IRM instrument channel is divided scram would result in a reactor into two groups. These are as shutdown well before any safety follows: limit is exceeded. 1. Passive type indicating Ir. order to ensure that the IRM devices that can be compared provided adequate protection with like units on a against the single rod withdrawal continuous basis, error, a range of rod withdrawal . accidents was analyzed. 1his 2. Vacuum tube or semiconductor analysis included starting the devices and detectors that accident at various power drift or lose sensitivity. j levels. The most severe case f involves an initial condition in Experience with passive type which the reactor is just instruments in generating subtritical and the IRH system is stations and substations not yet on scale. This condition indicates that the specified exists at quarter rod density. calibrations are adequate. For Additional conservatism was taken those devices which employ in this analysis by assuming that amplifiers, drift specifir.ations the IRH channel closest to the call for drif t to be less than withdrawn rod is bypassed. The 0.4%/ month; i.e., in the period results of this analysis show of a month a drif t of.4% would that the reactor is scrammed and occur and thus providing for peak core power limited to one adequate margin. For the APRM percent of rated power, thus system, drift of electronic mintainingMCPRabovetheIafety apparatus is not the only mit HCPR. Based on the a'bove consideration in determining a alysis, the IRH provides calibration frequency. Change in protection against 1ccal control power distribution and loss of rod withdrawal errors and chamber sensitivity dictate a continuous withdrawal of control calibration every seven days. rods in sequence and provides Calibration on this frequency backup protection for the APRH. assures plant operation at or below thermal limits. l Reactor low Water Level L The set point for low level scram i is above the bottom of the separator skirt. This level has I serwm o AmendmentNo.gP,[h 39

0 '~' L7 \\ 3.1 ILAlf3 (Cont'd) 4.1 M515 Kont1(f) t,n. V 1,o t been used in transient analyses A comparison of Tables 4.1.1 and l dealing with coolant inventory 4.1.2 indicates that two ) decrease. The results show that instrument channels have not been scram at this level adequajeJy included in the latter Jable. protects the fuel and the 'V)~/@s These are: mode switch in pressure barrier, because HC R shutdown and manual scram. All remains well above the safety of the devices or sensors limit MCPR in all cases, and associated with these scram system pressure does not reach functions are simple on-off the safety valve settings. The switches and, hence, calibration scram setting is approximately 25 during operation is not in, below the normal operating applicablef(i.e., the switch is range and is thus adequateto avoid spurlous scrams. ^f(u hE@i either on or off), ~ l l j The sensitivity of LPRH detectors Turbine Ston Valve Closn t decreases with exposure to i neutron flux at a slow and The turbine stop valve closure approximately constant rate, scram anticipates the pressure. This is compensated for in the neutron flurand heat flux APRM system by calibrating every increasethibcouldresultfrom three days using heat balance rapid closure of the turbine stop I data and by calibrating valves. Hith a scram trip individual LPRH's every 1000 setting of i 10 percent of valve effective full power hours using closure from full open, the TIP traverse data, resultant increase in surface ( heat flux is limited such that N-HCPR remains above the safety limit MCPR even during the worst g y,,e _y, / 4,,,, case transient that assumes the ? d " '* ' turbine bypass is closed. Turbine Control Valve fast Closure The turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejectionexceedingthe capability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the acceleration relay. This setting and the fact that control valve closure time is approximately twice as long as l L l Revision-153 o-l Amendment No. 42, 123, 138) 40 L

O 3.1 EMLS (Cont'd) that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. HCPR remains above the safety limit M:rR, t Main Condenser Lew Varmtm To protect the main condenser against overpressure, a loss of condenser i vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. lo anticipate the transient and automatic scram resulting low condenser vacuum from the closure of the turbine stop valves,Ipoint is selected to initiate initiates a scram. The low vacuum stram set ascrambeforetheclosurtoftheturbinesl5pvalvesisinitiated. Main Steam Line Isolation Valve Closure The low pressure isolation of the main steam lines at 880 psig (as specified in Table 3.2.A) was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shut'down so that high power operation-at low reactor pressure does not occu Athus providing protection for the fuel cladding integrity safety imit. Operation of the reactor at pressur ower than 785 psig requires that " the reactor mode switch be in the positiongwhere protection of is provided by the IRH hinh the fuel cladding integrity safet l neutron flux scram and APRH 15% scram. Thus, the combination of m.n steam lina low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protectkn over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. Hith the scrams set at 10 percent of valve closure, neutron flux does not increase. Hiah Reactor Pressure The high reactor pressure scram setting is chosen slightly above the maximum normal operating pressure to permit normal operation without allowablereactorcoolantsystempressure(1250psig,seeBas[III spurious scram, yet provide a wide margin to the ASHE Section t Section 3.6,0). e Hiah Drywell Pressun ,w Instrumentation 4p4sturt.twitcheW for the drywe11M,r+e provided to detect a loss of coolant accident and initiate the core stendby cooling equipmeM. A high drywell pressure scram is provided at the same setting as the Xoreff601ing A) stems (CSCS) initiation to minimize the energy %whith* must be accohmodated during a loss of coola:.t accident and to prevent w r return to cri ticality. This instrumentation is a backup to the reactor vessel water level instrumentation. y%a l 'An.endment-en-+33f gg > -40ap l i

3.1 MSLS (Cont 'd) Main Siem_ Lint Hich Radhlie High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector of f-gas monitors whicht cause an isolation of the main condenser off-gas line. 'gg ('^h Q T VXreattor mode switch is-provided-whtWettuates or bypasses the various (Ref 7ection 7.2.3.$p$jate to the particular plant operating status? " scram functionL m,ro r F t g %two $tnalScram The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation. Scram Discharae_fnstrument Volym_g Y Thecontrolroddrivescramsystemisdesigned50(hatallofthewater 'r4.J4hietPis discharged from the reactor by a scram cad be accommodated in the discharge piping. The two scram discharge volumesAat-comodate-in-enessf J--of-99 gallons of water each and are at the low points of the scram discharge piping. 40-ct+dit-vet-taken-for-these-volumes 4n-the-design-of-(> the-discharge-pipiniras concerns-the-amotmt-of-water-whith-mutt-bef Q -attomodated-during-a-scramy f -m During normal operation the scram discharge volume system is empty; -[FMa Y, however, should it fill with watey, the water disc arged to the piping 3 r could not be accommodatedr which ould result in s ow scram times or 'ef4 gM partial control rod insertion. T preclude this o currence, redundant and _ M divmtlev1Lde_tec110. devices i he scram disch 4ge instrument volumes (g m - M @ff have been provide @l) whit #,will alarm whene ater levelirea D7. initiate a contro rod b ock etf18 gallon WMWitiFTeactortwheAMTie m th'j V water level reaches 39 ga ns. As indic ed above, there is sufficient LW volume in the piping to accommodate the scram without impairment of the r scram times or amount of insertion of the control rods. This function f b2 / shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately ' pm./ 7 ~ ,c w c A pource range monitor ($RH) system is-also provided to supply additional nettronlevelinfqrmationduringstargtI'pbuthasnoscramfunctionsh l RefF 4ection 7 5 9) m 4 /uy -upy/ efue / and Ttartuptfiot Standby # modes with the Th7AbIM'scov[the APRM15% scram,andthepowerrangewiththeflog_ s ae M onasLP Mw sq wt 4 g w M Dt /A y (/of M b C&As GW i l 1

e o Mbbbbbtitk dk i biase)Refuelpockgndscram. rod The IRH's provide additional protection in the' and Startup/yptStandby4 modes. Thus,theIRgandAPRM15% scram are required in the ' Refuel'd and $ tartup/ Hot Standby modes I Section 73)gegjhe APRH system provides the required p the power ran %src 5JA Thus, the IRH system is not required in the 'Run gs The high reactor pressure, high drywell pressure, reactor low water leve 9 and t. cram discharge volume high level scrams are required for Startup/ Hot Standby and Run modes of plant operation. They are, therefore, required to be operational for these modes of' reactor operation. i The requirement to have the scram functions, as indicated in Table 3.1.1, i operable in the Refuel mode is to assure thateshifting to the Refuel mode during reactor power operation does not diminish the need for the reactor protection system. The turbine condenser low vacuum scram is only required during power operation and must be bypassed to startlijp the Unit. Below 305 psig turbine first stage pressure (45% of rated), the scram signal due to turbine stop valve closure or fast closure of turbine control valves is I bypassed because flux and pressure scram are adequate to protect the reactor. If the scram signal due to turbine stop valve closure or fast closure of turbine control valves is bypassed at lower powers, less conservative HCPR and HAPLHGR operating limits may be applied as specified in the CORE OPERATING LIMITS REPORT,, p TLpt Vo g. </06 i /"The requirement that the IRH's be inserted in the core when the APRH's read 2.5 indicated on the scale assures-thatethere is proper overlap in the neutron monitoring systems and thus,4hal provided for all ranges of reactor operation.padequate coverage is f. N

d Lhe provision of an AfRH scram at 115% design power in the / Refuel,P k

and Tartup/ Hot Standby"~rnodes and the backup IRH scram at 1120/125 of full scale assures tha4t there is proper overlap in the neutron monitoring ] systems andf thus,4hatpa,depi egyerage is provided for all ranges off g (reactoroperation.f > M&W tb %4 4 g,3f l M .k b 'Wm f. N b hO M t I . Ravision453-P _lmdent-Ntn-133r-43BP b y W e, -40ef ts. c{cG.ts/ g

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l TABLE 3.2.A 2NPS') INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATICH i Minimum)# of. t Operable Instrument r,p p., .n.., Channels Per Trio System (1) -IMtremeet-Trio Level Settina Action (21 I-Assat e l2(7){ a Reactor low Hater Level 19' indicated level (3) A and D \\ ,1 i / Reactor High Pressure 1110 psig D 1 \\ 1 2 2 Reactor Low-Low Water Level at or above -49 in. A indicated level (4) 2 g Reactor High Hater Level 1 8" indicated level (5) B 4 j 2(7) 2 High Dryv. Pressure 12.5 psig A l ]- p2 j 2. High Radiation Main Steam 17 times normal rated B Line Tunnel (9) full power background I I 2 2. Low Pressure Main Steam Line 1880 psig (8) B 2(6)) 2. High Flow Main Steam Line 11407 of rated steam flow B l 2 2 Main Steam Line Tunnel Exhaust Duct High Teg erature 1170*F B i l f ! 2 2 Turbine Basement Exhaust Duct High Temperature 1 50*F B 1 I 1 f I Reactor Cleanup System High Flow 13007, of rated flow C j. d 2 Reactor Cleanup System ^ High Temperature 1 50*F C 1 Amendment No. E5', 45

EQII5l 0 U ADL U L A 1. Whenever Primary Contain:ent inte rity is required by Section 3.7. there shall be two operable or tripped rip systens for each function. An -) instrument channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter; or, where only one channel exists per trip system, the other trip system shall be operable.

2. LC.tLCD y(W l.ae V c

Lae 'If Thi first coiumn~ c~annot be met for onf of ihe trip systems, that tri~p 'o - ~ ~~ / Ifthefirstcolumncannotberetforbothtrip) (.systemshallbetripped. systems,theappropriateactionlistedbelowshallbeta[e A. Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours. B. Initiate an orderly load reduction and have Main $ team t.ines isolated l! within eight hours. C. Isolate Reactor Hater Cleanup System. D. Isolate Shutdown Cooling. f3. Instrument set point corresponds to 128.26 inches above tap of " active fue. I 4. Instrument set point corresponds to 77.26 inches above top of active fuel. / i 5. Not required in Run Mode (bypassed by Mode Switch). 6. Two required for each steam line. v { 7. These signals also start SBGTS and initiate secondary containment isolation. I 8. Only required in Run Mode (interlocked with Mode Switch). 9. Hithin 24 hours prior to the planned start of hydrogen inject'on with the .l reactor power at greater than 201 rated power, the normal full power L radiation background level and associated trip setpoints may be changed i based on a calculated value of the radiation level expected during the injection of h drogen. The background radiation level and associated trip j setpoints may o adjusted based on either calculations or measurements of i actual radiation levels resulting from hydrogen injection. The background f radiation level shall be determined and associated trip setpoints shall be i set within 24 hours of re-establishing normal radiation levels after I completionofhydrogeninjectionandpriortowithdrawingcontrolrodsat/ i reactor power levels below 20% rated power. y l '~__ _ l s y f ? e, ,% + yt 3 Revision-122 " -[ l Amendment No. 86,705.!!8.<119; 46 l

PNPS TABLE 3.2.C-1 INSTRUMENTATION THAT INITIATES ROD SLOCKS I ( w r.. J.. n l 11M d OperableJChannels Required Trio Function oer Trio function Coerational Conditient Notes i j 1 A .t a ta - APRM Upscale (Flow Biased) N] c Run (!) \\ APRM Upscale {4 1 4 Startup/ Refuel (1) 1 APRM Inoperative k4 4 Run/Startup/ Refuel (1) \\ APRM Downstale 4 { 6 Run (1) i Rod Block Monitor '2 } 2 Run, with limiting control rod (2) l (Power Dependent) pattern, and reactor power > LPSP (5) l l Rod Block Monitor 2l 2 Run, with limiting control rod (2) l 4 Inoperative .l pattern, and reactor power > LPSP (5) { i Rod Block Monitor 2 2 Run, with limiting control rod (2) { Downscale pattern, and reactor power > LPSP (3) i IRM Downscale. 6 I T Startup/ Refuel, except trip is (1) bypassed when IRM is on its lowest range IRM Detector not in 6l / Startup/ Refuel, trip is bypassed (1) Startup Position when mode switch is placed in run IRM Upscale 6 f Startup/ Refuel (1) IRM Inoperative 6 F Startup/ Refuel (1) SRM Detector not in 3 4 Startup/ Refuel,.except trip is by-(1) Startup Position j passed when SRM count rate is 1 100 i counts /second or IRMs on Range 3 or above (4) I ihl 4 Startup/ Refuel, except trip is by-(1) SRM Downscale passed when IRMs on Range 3 or above (4) Revision M3-o-Amendment No. 75, 27, 42, 65, 72, 79, IJO, 129,A35, 54

PNPS TABLE 3.2.C-1 (Con't) 1 _ /I-s t ~cp (iiinimum)DperableMannels Required i Trio Function oer Trio function ODerational Conditions Notes l l n.r+ ue _- SRM Upscale Ti 4 Startup/ Refuel, except trip is by-(1) i passed when the IRM range switches are on Range 8 or above (4) SRM Inoperative 3i V Startup/ Refuel, except trip is by-(1) [ passed when the IRM range switches are on Range 8 or above (4) t Scram Discharge !2+ 2 Run/Startup/ Refuel (3) 1 Instrument Volume Hater I Level - High 7f a l l Scram Discharge f 1 / Run/Startup/ Refuel (3) Instrument Volume-Scram 1 Trip Bypassed / f Recirculation Flow 2 2-Run (1) Converter - Upscale i I Recirculation Flow ,I 2 Run (1) 2 Converter - Inoperative / i k\\ y( Recirculation Flow a Run (1) Converter - Comparato-Mismatch hisien '9-Amendment No. 13F', 54a 1

j!lil j a b BS ) ) ) 7 7 7 s ( ( ( e t ) ) ) o 4 4 4 N ( ( ( t t t n n n i i i o o o p r p r p n i h i h i r o t / t / t h i l R l R l / te u u u R ac M 4 M 4 M 4 cn / 0 / 0 / 0 ia r 1 r 1 r 1 dR or or or n teo t eo teo Id adt adt adt n cr cr cr g ig ig ea i p dc-dc-dc-y ne0 ne0 ne0 T IR1 IR1 IR1 N O IT )A dT N 't E nM oU CR t t (T n n S e e F. I V V N g g

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o. *.. O PNPS TABLE 4.2.C MINIMUM TEST AND CALIBRATION FRE00ENCY FOR CONTROL R00 BLOCKS ACTUATION Instrument Channel Instrument Functional Calibration Instrument Check Ifl.t APRM - Downstale HH4}^ Once/3 Months Once/ Day APRM - Upscale w M)-G}" Once/3 Months Once/ Day APRM - Inoperative jy j,w f(, f (4M3)e Not Applicable Once/ Day IRM - Upscale 1 (2) (3) Startup or Control Shutdown (2) IRM - Downscale (2) (3) Startup or Control Shutdown (2) IRM - Inoperative (2) (3) Not Applicable (2) RBM - Upscale --r- (+)-GF Once/6 Months Once/ Day RBM - Downstale i e UM3)~ Once/6 Months Once/ Day RBM - Inoperative H)-( 3)= Not Applicable Once/ Day SRM - Upscale (2) (3) Startup or Control Shutdown (2) SRM - Inoperative (2) (3) Not Applicable (2) SRM - Detector Not in Startup Position (2) (3) Not Applicable (2) SRM - Downscale (2) (3) Startup or Control Shutdown (2) IRH - Detector Not in Startup Position (2) (3) Not Applicable (2) Scram Discharge Instrument Volume Once/3 Months Refuel Not Applicable Hater Level-High Scram Discharge Instrument Once/3 Months Not Applicable Not Applicable Volume-Scram Trip Bypassed Recirculation Flow Converter Not Applicable Once/ Cycle Once/ Day Recirculation Flow Converter-Upscale Once/3 Months Once/3 Months Once/ Day Recirculation Flow Converter-Inoperative Once/3 Months Not Applicable Once/ Day Recirculation Flow Converter-Ccmparator Once/3 Months Once/3 Months Once/ Day Off Limits Recirculation Flow Process Instruments Not Applicable Once/ Cycle Once/ Day Locic System Functional Test (4) (6) System Logic Check Once/18 Months RevisiefrticM43 Amendment No. 710. J.30' 63

1 o NOTES FOR TABLES 4,2.A THROUGH 4.2.G 1. Initially once per month until exposure hours (H as defined on figure 5 4.1.1) is 2.0 x 10 ; thereafter, according to figure 4.1.1 with an interval not less than one month nor more than three months. 2. Functional tests, calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped, functional tests shall_be performed before each startup with a required frequency not to exceed once per week. Calibrations of IRHs and SRHs shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per day during those periods when the instruments are required to be operable. 3. This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel. 4. Simulated automatic actuation shall be performed once each operating cycle. Hhere possible, all logic system functional tests will be performed using the test jacks. 5. Reactor low water level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2.A since they are tested onTablesg.1.2. 7 6. The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems. 7. Calibration of analog trip units will be performed concurrent with functional testing. The functional test will consist of injecting a simulated electrical signal into the measurement channel. Calibration of associated analog transmitters will be performed each refueling outage. l 4 l l l Revision 113 67

o MSLS: $k$) ,g. TheinstrumentationlisfedinTable4.2.Athru4.2.hwillbe 4.2 functionally tested and calibrated at regularly scheduled intervals. The same design reliability goal as the Reactor Protection System of 0.99999 is generally applied for all applications of (1 out of 2) X (2) logic. Therefore, on-of f sensors are tested once/3 nonths, and bi-stable trips associated with analog. sensors _and amplifiers are_ tested 7--- 1, a I w o' ~ tw <- / I,.. < 14... o t a ri,Jh. J onceiweek. -n lJhose i struments whicht shen-trip Ed,~ result in-a-rod block have-their v L-. tr y e k,

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contacts arranged _in lout of-n-l' gic and all.are capable of being " r bypassed. Jor such a tripping arrangement with bypass-capability" provided;-there is an optimum test interval that shotild be maintained in order to maximize the reliability of a given channel (7). This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by: i-pltr Where: i= the optimum interval between tests. t-the time the trip contacts are disabled from performing their function while the test is in progress. r-the expected failure rate of the relays. To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed this task requires an estimated 30 mihutes to complete in a thorough a workmanlikemannerandthattherelayshaveafailurerateof10'gd failures per hour. Using this data and the above operation, the optimum test interval is 1 x 103 hours - \\ / 210pl i V 10- - 40 days [pr additional marJ n a test interval of on_Le_ler month will be_uigd initially. i (7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodh Inspection, Benjamin Epstein, Albert Shif f, July 16, 1968, page 10 Equation (24), Lawrence Radiation Laboratory. 4.. J._.. /,A

~ _ _ _ _ _ _ _ _ _ o 4.2 ILASIS (Cont'd) is shown by Curve No. 2. Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability. A more unusual case is that the testing is not done independently. If U, both channels are bypassed and tested at the same time, the result is shown in Curve No. 3. Note that the minimum occurs at about 40,000 hours, much longer than for cases 1 and 2. Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel. Bypassing both channels for simultaneous testing should be avoided. The most likely case would be to stipulate that one channel be b) passed, tested, and restored, and then immediately following, the second channel be bypassed, tested and restored. This is shown by Curve No. 4. Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel. The best test procedure of all those examined is to perfectly stagger the tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5. The difference between Cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human error. The conclusions to be drawn are these: 1. A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and 2. more than one channel should not be bypassed for testing at any one time. The radiation monitors in the refpeling area ventilation duct which 'initiatebuildingisolationandTstandby.$sfreatmentoperationare arranged in two 1 out of 2 logic systems. The bases given above for-the^ Tod blockseapply here also and were used to arrive at the functional testing frequency. Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate. AmendmentNo.89; 76

e e' e e 4.2 !} ASIS (Cont'd) The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also. The instrumentation which is required for the recirculation pump trip and alternate rod insertion systems incorporate analog transmitters. The transmitter calibration frequency is once per refueling outage, which is consistent with both the equipment capabilities and the requirements for similar equipment used at Pilgrim. The Trip Unit Calibration and Instrument functional Test is specified at monthly, which is the same frequency specified for other similar protective devices. An instrument check is specified at once per day; this is considered to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip and alternate ro,d insertion systems are backups to existing protective instrumentation. /'// A logic system functional test interval of 18 months was selected to minimize the frequency of safety system inoperability due to testing and to minimize the potential for inadvertent safety system trips and their attendant transients. Based on industry experience and DWR Standard Technical Specifications, an 18 month testing interval provides adequate assurance of operability for this equipment. ^ to Al') Co., f-. / b. J Glo, D p-.I I' C / I i n L ~.. t r ti e n c a sv. -oII+,es i., f a s e t,

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hre< Y e *, + s s pe$ ny e f ". n r j 3 p. fe, f,,f,, i, s v + t.n P - A, s y /*-... fs i+<d z, /-> wef4 A/ E O c: aey si y ,g g g ILf,2v><d I; y fk e k' A C gJ s ,.,,,,, g j,,, ( /e WL3 h 0, it!. G. w e R, c, [ A),,,, * (, S,.y + 4 e.- ? t, f 9 V J' '*'I h n a<r 6 /1 rey ) l f ~~" 'N%___.__..- Revision No.143 Amendment No. 42, 121, 130 77 -}}

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