ML20035B164

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Amend 147 to License DPR-35,revising RPS Instrumentation Surveillance Test Intervals from 1 Month to 3 Months, Provides for 12- & 6-hr Allowable out-of-svc Times for Repairs,Tests & Deletes Water Level Perturbation
ML20035B164
Person / Time
Site: Pilgrim
Issue date: 03/25/1993
From: Butler W
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20035B165 List:
References
NUDOCS 9303310272
Download: ML20035B164 (29)
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oq[o UNITED STATES 3 NUCLEAR REGULATORY COMMISSION -y f, .c W ASHINGTON. D. C. 20555 1 %, '=... / t r i BOSTON EDISON COMPANY t DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATInff AMENDMENT TO FAClllTY OPERATING LICENSE Amendment No.147 License No. DPR-35 1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that: A. The application for amendment filed by the Boston Edison Company (the-licensee) dated February 7, 1992 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 3.8 of facility Operating License No. DPR-35 is hereby amended to read as follows: + i 9303310272 930325 PDR ADOCK 05000293 p PDR-

B i t I i e Technical Specifications I i The Technical Specifications contained in-Appendix A, as revised .through Amendment No.147, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical [ Specifications. l 6 3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days. [ l FOR THE NUCLEAR REGULATORY COMMISSION t [ ) L. t Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Attachment:

( i Changes to the Technical ~ Specifications

}

Date of Issuance: March 25. 1993 J I i l t 3 .L I i l I i

i 6 ATTACHMENT TO LICENSE AMENDMENT NO.147 FACILITY OPERATING LICENSE NO. DPR-35 t DOCKET NO. 50-293 l Replace the following pages of the Appendix A Technical Specifications with l the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages have been provided.* Remove Insert 27 27 28 28 29 29 30 30 31 31 32 32 33 33 34 34 35 35 36 36 37 37 38 38 39 39 40 40 41 41 45 45 46 46 46a 54 54 54a 54a l 58b 58b 63 63 67 67 74 74 76 76 77 77 i l i f i

i PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMENT Operable Inst. Modes in Hhich Function Channels Per Trip Function Trip Level Setting Must Be Operable Action (I) Trio Sys':em (1) Refuel (7) Startup/ Hot Run Minimum Avail. Standby i 1 Mode Switch in Shutdown X X X A 1 1 Manual Scram X X X A IRM 3 4 High Flux 1 20/125 of full scale X X (5) A 1 3 4 Inoperative X X (5) A APRM 2 3 High Flux (15) (17) (17) X A or 8 2 3 Inoperative (13) X X(9) X A or B 2 3 High Flux (15%) 115% of Design Power X X (16) A or 8 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 19 In. Indicated Level X X X A SDIV High Hater Level: <39 Gallons X(2) X X A 2 2 East 2 2 Hest 2 2 Main Condenser low Vacuum 123 In. Hg Vacuum X(3) X(3) X A or C 2 2 Main Steam Line High 17X Normal Full Power Radiation Background (18) X X X(18) A or C 4 4 Main Steam Line Isolation Valve Closure 1 0% Valve Closure X(3)(6) X(3)(6) X(6) A or C 1 2 2 Turbine Control Valve 1150 psig Control Oil Fast Closure Pressure at Acceleration Relay X(4) Xt4) X(4) A or D 4 4 Turbine Stop Valve Closure 110% Valve Closure X(4). X(4) X(4) A or D Amendment No. 15, 42, 86, 92, 117 133,147 27

i s ~ f f NOTES FOR TABLE 3.1.1 i 1. There shall be two operable or tripped trip systems for each trip function 4 (e.g., high drywell pressure, reactor low water level, etc.). An instrument j. j channel, satisfying minimum operability requirements for a trip system, 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 ' l one OPERABLE channel in the same trip system is monitoring that parameter. t i An inoperable channel and/or trip sy, tem need not be placed in the tripped { condition if this would cause a full scram to occur. When a trip system can 1 be placed in the tripped :ondition eithout causing a full scram to occur, place the trip system with the mos', inoperable channels in the tripped i condition, per the table below. '.f both systems have the same number of inoperable channels, place either trip system in the tripped condition, per 1 the table below. 1 Condition Reouired Action Completion Time

a. With less than the Place associated trip 12 hoars r

minimum required system in trip 4 operable channels per l trip function in one a l trip system. or

  • j
b. With less than the Place one trip system 6 hours minimum required in trip i

operable channels per i trip function, in =l both trip systems. or *

c. If full scram trip Restore RPS trip I hour l

capability is not capability j available for a given i trip function or

  • i 2
  • 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. I B. Reduce power level to IRM range and place mode switch in the startup/ hot standby position within eight (8) hours. I 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. l l ) ) Amendment No. 6, 119, 147 28 I -.. i

t r-NOTES FOR TABLE 3.1.1 (Cont'd) i 2. Permissible to bypass, with control rod block, for reactor protection system reset in rifuel and shutdown positions of the reactor mode switch. 3. Permissible to bypass when reactor pressure is <600 psig. ~ 4. PermissiF'S to bypass when turbine first stage pressure is less than 305 psig. 5. IRH's are bypassed when APRH's are onstale 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 ta be operable: A. Mode switch in shutdown B. Manual scram I C. High flux IRH D. Scram discharge volume high level E. APRM (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 APRM 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 APRM.
14. Deleted
15. The APRM high flux trip level r,etting shall be as spet:fied 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 APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.
18. Within 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 : hanged 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 20% rated power. t 8 Amendment No. 6, 75, 27, 42, E6,177, IIE,123,147 29

TABLE 4.1.1 REACTOR PROTECTION SYSTEM (SCRAH) INSTRUMENTATION FUNCTIONAL TESTS HINIMUM FUNCTIONAL TEST FRE0VENCIES FOR SAFETY INSTRUMENTATION AND CONTROL CIRCUITS Functional Test Minimum frequency (3) Mode Switch in Shutdown Place Mode Switch in Shutdown Each Refueling Outage Manual Scram Trip Channel and Alarm Every 3 Honths RPS Channel Test Switch (S) Trip Channel and Alarm Once Per Heek IRH High Flux Trip Channel and Alarm (4) Once Per Heck During Refueling and Before Each Startup Inoperative Trip Channel and Alarm Once Per Heek During Refueling and Before Each Startup APRH High Flux Trip Output Relays (4) Every 3 Honths (7) Inoperative Trip Output Relays (4) Every 3 Honths Flow Bias Trip Output Relays (4) Every 3 Honths High Flux (151) Trip Output Relays (4) Once Per Heek During Refueling and Before Each Startup High Reactor Pressure Trip Channel and Alarm (4) Every 3 Honths High Drywell Pressure Trip Channel and Alarm (4) Every 3 Honths Reactor low Hater level Trip Channel and Alarm (4) Every 3 Honths High Hater Level in Scram Discharge Tanks Trip Channel and Alarm (4) Every 3 Honths Turbine Condenser low Vacuum Trip Channel and Alarm (4) Every 3 Honths Main Steam Line High Radiation Trip Channel and Alarm (4) Every 3 Honths Main Steam Line Isolation Valve Closure Trip Channel and Alarm Every 3 Honths Turbine Control Valve Fast Closure ' Trip Channel and Alarm Every 3 Honths Turbine First Stage Pressure Permissive Trip Channel and Alarm (4) Every 3 Honths Turbine Stop Valve. Closure Trip Channel and Alarm Every 3 Honths Reactor Pressure Permissive Trip Channel and Alarm (4) Every 3 Honths Amendment No. 6, 79, 99, 117 147 30 .~.

i i. NOTES FOR TABLE 4.1.1 I 1. Deleted 2. Deleted 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. i 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, f 6. Deleted } 7. This APRM testing will be performed once every 3 months when in the RUN mode and within 24 hours after entering RUN mode, if not performed within f the previous seven days. I t I l Amendment No. 6, 79, 147 31 1

~ L 1 TABLE 4.1.2 REACTOR FROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FRERJENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS l instrument Channel Calibration Test (5) Minimum Freauency (2)- IRM High Flux Comparison to APRM on Controlled Note (4) -Shutdowns Full Calibration Once/ operating cycle APRM High Flux 1 Output Signal Heat Balance Once every 3 Days Flow Dias Signal Calibrate Flow Comparator and Each Refueling Outage Flow Blas Network Calibrate Flow Blas Signal (1) Every 3 Months 1 LPRH Signal TIP' System Traverse Every 1000 Effective ~ Full Power Hours High Reactor Pressure Note (7) Note'(7) f High Drywell Pressure Note (7) . Note (7) Reactor Low Hater Level Note (7) Note (7) c High Hater Level in Scram Discharge Tanks Note (7) Note (7) Turbine Condenser low Vacuum Note (7) Note (7) I Main Steam Line Isolation Valve Closure -Note (6) Note (6) j ' Main Steam Line High Radiation Standard Current Source (3) Every 3 Months 3 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) l Reactor Pressure Permissive Note (7) . Note-(7) ] f -Amendment'No.' 6, 40, 79. 99,- 147' 32 l l: .., c,. .....u.. ,_,,,..,.._._.,;._......_.._.,_.___,.-_.u_...;._.._:.. .,,m.

l 0 I t NOTES FOR TABLE 4.1.2 1. Adjust the flow bias trip reference, as necessary, to conform to a i 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 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. Amendment No. 79, 99, 147 33

-~ y i BASES: ~3.1 The reactor protection' system automatically initiates a reactor scram to: 1. Preserve the integrity of the fuel cladding. -i 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 criticality. t 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 may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to - conduct required functional tests and calibrations. The reactor protection system is of the dual channel type (Reference FSAR l Section 7.2). The system is made up of two independent trip systems, each having two subchannels of tripping devices.. Each subchannel has an j 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 I a trip on both systems is required'to produce a reactor scram. 1 This system meets the intent of IEEE-279 for Nuclear Power Plant Protection Systems. The system has a reliability greater than that of a ) 2 out of 3 system and somewhat less than that of a 1 out of 2 system. -i i Hith the exception of the Average Power Range Monitor (APRM) channels-i the Intermediate Range Monitor (IRM)' channels, the Main Steam Isolation Valve closure, and.the Turbine Stop Valve, closure, each subchannel has l 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 i system is placed in a tripped condition, the effectiveness of the i protection system is preserved (i.e., the system can tolerate a single } failure and still perform its intended function of scramming'the i reactor). Three APRM instrument channels are provided for each j protection trip system. j for some trip functions (e.g. MSIV or Turbine Stop Valve Position U Switches), the loss of one instrument may lead to degradation of both .i trip systems. 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 j Section 7.5.4). { Amendment No. 79,147 34

3.1 EASH (Cont'd) The requirement that the IRM's be inserted in the core when the APRM's read 2.5 indicated on the scale assures there is proper overlap in the neutron monitoring systems and thus, sufficient coverage is provided for all ranges of reactor operation. The provision of an APRM scram at 115% design nower in the Refuel and Startup/ Hot Standby modes and the backup IRM s iam 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 151 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 Stanaby 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 shifting 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 Rance Monitor (APRM) l APRM's #1 and #3 operate contacts in one subchannel and APRM's #2 and #3 operate contacts in the other subchannel. APRM'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 APRM per i protection trip system for maintenance, testing, or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. Amendment No. 79, 147 35

4 3.1 BASES (Cont'd) The APRM system, which is calibrated using heat balance data taken dur steady-state conditions, reads in percent of design p + During transients, the responds directly to average neutron flux. instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the tim power of the fuel will be less than that indicated by the neutron flux at the fuel. Analyses demonstrated that with a 120 percent scram trip setting, none of the abnormal operational transien the scram setting. l Therefore, the use of flow-referenced scram trip provides even damage. additional margin. An increase in the APRM scram setting would decrease the margin presen The APRM before the fuel cladding integrity safety limit is reached. Reducing provide a reasonable range for maneuvering during operation. this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM setting was selected because it p 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 MCPR when the transient is initiated from HCPR above the operating limit NCPR. For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power The margin is sufficient to accommodate anticipated maneuversEf rated. associated with power plant startup. 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 patte worth minimizer. Thus, of Horth of individual rods is very low in a uniform rod pat Because the flux is the most probable case of significant power rise. distribution associated with uniform rod withdrawals does no high local peaks, and because several rods must be moved to change po by a significant percentage of rated power, the rate of very slow. In an assumed uniform rod withdrawal approach to the scram fission rate. 36 Amendment No. 79, 133, 147

l 3.1 BASES (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 Rance Monitor (IRM) The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM 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 IRH 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 IRM is ranged up to 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 equilibriuw with the neutron 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 i 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 Water 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, 123, 147 37

4 3.1 BASES (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 i The turbine stop valve closure scram anticipates the pressure, neutron I flux, and heat flux increase that could result from rapid closure of the l turbine stop valves. With a scram trip setting of 110 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 l of the turbine control valves due to load rejection exceeding the i capability of the bypass valves. The reactor protection system initiates 1 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 i valve closure. HCPR remains above the safety limit MCPR. 1 Hain CorJenser Low Vacuum l To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turoine stop valves and turbine l bypass valves. To anticipate the transient and automatic scram resulting j from the closure of the turbine stop valves, low condenser vacuum i initiates a scram. The low vacuum scram setpoint is selected to initiate l l a scram before the closure of the turbine stop valves is initiated. Main Steam Line Isolation Valve Closure f 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 reattor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRH high neutron flux scram and APRH 157. scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire Amendment No. 6, 79, 99, 122,147 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 i closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase. Hich 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). f 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 Hich Radiation 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, 1 Reactor Mode Switch i The reactor mode switch actuates or bypasses the various scram functions appropriate to the particular plant operating status (Reference FSAR I Section 7.2.3.9). 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. Amendment No. 6, 79, I22, 147 39

3.1 j_AE S (Cont'd) Scram Discharoe Instrument Volume The control rod drive scram system 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 discharged 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 j 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 i 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). 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 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). 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, 722, 728, 147 40

.P s-t t I '7 t - i r t. This page is intentionally left blank, I l r i 9 k i i P f l Amendment No. 147 43

~ p. PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRlHARY CONTAINMENT ISOLATION Operable Instrument Channels Per Trio System (1) Minimum 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 B Line Tunnel (9) full power background 2 2 Low Pressure Main Steam Line 1880 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 1170*F B 2 2 Turbine Basement Exhaust Duct High Temperature 1 50*F B 1 1 1 Reactor Cleanup System High Flow 1300% of rated flow C 2 2 Reactor Cleanup System High Temperature 1150*F C Amendment No. 86, 147 45-

3 aj 'l .i 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 t instrument channel may.be placed in an inoperable status for up to 6 hours I 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, I the other trip system shall be operable. 2. Action t If the minimum number of operable instrument channels cannot be met for one i of the trip systems of a trip function, the agropriate conditions listed' l below shall be followed: i 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 Drywell Pressure, and Main Steam Line j High Radiation) or initiate the action required by Table 3.2.A for the i affected trip functions, i If placing the inoperable channel (s) in the tripped condition would i cause an isolation, the inoperable channel (s) shall be restored to operable status within two hours (six hours for Reactor Low Water Level, j High Drywell Pressure, and Main Steam Line High Radiation) or initiate j the Action required by Table 3.2.A for the affected trip function. i 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. i A. Initiate an orderly shutdown and have the reactor in Cold Shutdown l Condition in 24 hours. i B. Initiate an orderly load reduction and have Main Steam Lines isolated j within eight hours. i C. Isolate Reactor Water Cleanup System. D. Isolate Shutdown Cooling. Amendment No. 86, 105, 119,147 46

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3. ' Instrument; set point corresponds to.128.26 inches' above top of active. fuel, i

4. Instrument set point corresponds to 77.26 inches above top of active fuel. 5. Not required in Run Mode (bypassed by Mode Switch). j 6. Two required for each' steam line. i 7. These signals also start SBGTS and initiate secondary containment j isolation. 8. Only required in Run Mode'(interlocked with Mode Switch). i 9. Within 24 hours prior to the planned start of hydrogen injection with the -l 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 .i actual radiation levels'resulting from hydrogen injection. The background i 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 i reactor power levels below 20% rated power. i i r t f i Amendment No. 147 46a a

1 r PNPS TABLE 3.2.C-1 INSTRUMENTATION THAT INITIATES R00 BLOCKS Operable Instrument Channels Required Trio Function oer Trio Function Doerational Conditions Notes Minimum l Available APRM Upscale (Flow Blased) 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 Honitor 2 2 Run, with limiting control rod (2) Downscale pattern, and reactor power > LPSP (5) IRH Downscale 6 8 Startup/ Refuel, except trip is (1) bypassed when IRH is on its lowest range IRH Detector not in 6 8 Startup/ Refuel, trip is bypassed (1) Startup Position when mode switch is placed in run IRH Upscale 6 8 Startup/ Refuel (1) IRH 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 1 100 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, 110, 129, 138, 147 S4 a

PNPS TABLE 3.2.C-1 (Con't) Operable Instrument Channels Required Trio Function oer Trio Function Doerational Conditions Notes Minimum 1 Available SRH Upscale 3 4 Startup/ Refuel, except trip is by-(1) passed when the IRH range switches are on Range 8 or above (4) SRH 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 Hismatch Amendment No. 138, 147 54a

TABLE 3.2.F (Cont'd) SURVEILLANCE INSTRUMENTATION Minimum # of Operable Instrument Type Indication Channels Instrument # Parameter and Ranae Notes 1 RI 1001-609 Reactor Building Vent Indicator /Hultipoint (4) (7) { RR 1001-608 Recyrder 4 R/hr 10- to 10 1 RI 1001-608 Main Stack Vent Indicator /Multipoint (4) (7) RR 1001-608 Recyrder 10- to 104 R/hr 1 RI 1001-610 Turbine Building Vent Indicator /Multipoint (4) (7) RR 1001-608 Recyrder 4 10- to 10 R/hr Amendment No. 83, 147 58b

o. PNPS TABLE 4.2.C MINIMUM TEST AND__ CALIBRATION FRE00ENCY FOR CONTROL ROD BLOCKS ACTU?,TI_QN Instrument Channel instrument Functional Calibration Instrument Chert leit APRM - Downstale Once/3 Honths Once/3 Honths Once/ Day APRM - Upscale Once/3 Honths Once/3 Honths Once/ Day APRH - Inoperative Once/3 Honths Not Applicable Once/ Day IRH - Upscale (2) (3) Startup or Control Shutdown (2) IRH - Downstale (2) (3) Startup or Control Shutdown (2) IRH - Inoperative (2) (3) Not Applicable (2) RBH - Upscale Once/3 Honths Once/6 Honths Once/ Day RBH - Downscale Once/3 Honths Once/6 Honths Once/ Day RBH - Inoperative Once/3 Honths Not Appilcable 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 - Downstale (2) (3) Startup or Centrol Shutdown (2) IRH - Detector Not in Startup Position (2) (3) Not Applicable (2) Scram Discharge Instrument Volume Once/3 Honths Refuel Not Applicable Hater Level-High Scram Discharge Instrument Once/3 Honths Not Applicable Nut Applicable. Volume-Scram Trip Bypassed Recirculation Flow Converter Not Applicable Once/ Cycle Once/ Day Recirculation Flow Converter-Upscale Once/3 Honths Once/3 Honths Once/ Day Recirculation Flow Converter-Inoperative Once/3 Honths Not Applicable Once/ Day Recirculation Flow Converter-Comparator Once/3 Honths Once/3 Honths Once/ Day Off Limits Recirculation Flow Process Instruments Not Appilcable Once/ Cycle Once/ Day Loaic System functional Test (4) (6) System logic Check Once/18 Honths Amendment No. 110, 130, 147 63

4 a e NOTES FOR TABLES 4.2.A THROUGH 4.2.G 1. Initially once per month until exposure hours (M as defined on Figure 5 4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not lest 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 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. Where possible, all logic system functional tests will be performed using the test jacks. 5. Reactor low water level, high drywell pressure and main ste'am line high radiation are not included on Taole 4.2.A since they are tested on Tables 4.1.1 and 4.1.2. I S. 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. ~ f 1 I l Amendment No. 147 67

a BASES: 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: -{4 Where: i-the optimum interval between tests. t t-the time the trip contacts are disabled from i performir.g their function while the test is in progress. r-the expected failure rate of the relays. To test the trip relayc requires that the channel be bypassed, the test i made, and the system returned to its initial state. It is assumed this task requires an estimated 30 minutes to complete in a thorough a0d workmanlike manner and that the relays have a failure rate of 10-failures per hour. Using this data and the above operation, the optimum test interval is 1 x 103 hours 1 2(0.5) 10-0 - 40 days For additional maroin a test interval of once per 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 147 74

.a 4 4.2 BASES (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, i r 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 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 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 I r 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. I Amendment No. 29, 147 76

j> 2 4.2 BASES (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 4 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 intervals and maintenance outage times selected in accordance with NEDC-30851P-A, Supplements 1 and 2 as approved by the NRC and documented in SERs (letters to D. N. Grace from C. E. Rossi l dated September 22, 1988 and January 6, 1989). l 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 BWR Standard Technical Specifications, an 18 month testing interval provides adequate assurance of operability for this equipment. Amendment No. 42, 727, 120, 147 77}}

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