ML20207J816

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Proposed Tech Specs Re Nuclear Instrumentation,Reflecting Replacement of Aging Equipment & Provision of Shutdown Functions for Reactor W/Numac Instruments from GE
ML20207J816
Person / Time
Site: La Crosse File:Dairyland Power Cooperative icon.png
Issue date: 12/19/1986
From:
DAIRYLAND POWER COOPERATIVE
To:
Shared Package
ML19292G558 List:
References
LAC-12019, NUDOCS 8701080606
Download: ML20207J816 (40)
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Text

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4 December 19, 1986 In reply, please refer to LAC-12019 DOCKET NO. 50-409 ENCLOSURE A Current Technical Specifications P

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2.8.4 'Ihe drive mechanisms shall have individual scram accumulators which drive the hydraulic motors and which shall be connected to the gas charging system and to the hydraulic charging system through individual check valves.

2.8.5 Each control rod shall be decelerated near the end of its scram stroke by an internal hydraulic dashpot action.

2.8.6 A rod test circuit shall permit the withdrawal and testing of one selected rod when the rod test key switch is turned on, when one of the rods has been selected for testing, when all other rods are full in, and when the period exceeds 7 see on both source range ch'annels.

This test circuit shall measure the time from the dropping out of the "All Rods Scram" relay to the indication of full insertion of the rod.

When the rod test key switch is turned on, occurrence of a period less than 7 see on either of the source range channels-shall initiate a full scram.

2.8.7 Sustained loss of voltage to Reactor Building Motor Control Center 1A.(which supplies the hydraulic charging pumps and the control rod drive motors), or low gas pressure, or a low icvel of hydraulic fluid in any scram accumulator shall initiate safety action as specified in

) Table 1.

2.9 CONTROL SYSTEMS I 2.9.1 The power level of the reactor shall be adjusted manually by the positioning of control rods or by changes in forced the circulation pump speed, or automatically by pressure

, controlled changes in the forced circulation pump speed.!

2.9.2 , The position of the scoop tubes which control the forced ion circulation pump speed shall be controlled mariually or automatically by the pressure control system.l 2.9.3 Delete .

2.9.4 An Initial Pressure Regulating System shall assume control of turbine inlet valves at low pressures so as to maintain turbine inlet pressure within a controlled band. The high pressure set point of the control band shall normally be set below'the operating set point of the forced circulation pump automatic control system. To maintain rated turbine inlet pressure, the Initial Pressure Regulating System may be used without use of the forced circulation pump automatic control system. -

ChangeNo./,/,4

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7 f

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2.9.5 Reactor water level shall normally be controlled automatically by a three-element control system which senses unvoided reactor water level, feedwater flow, and main steam flow, and which adjusts feedwater pump speed by positioning scoop tubes within fluid couplings between the feedwater pumps and their motors.

2.10 SAFETY AND MONITORING SYSTEMS 2.10.1 Ceneral 2.10.1.1 The instrumentation and safety systems shall provide the information and automatic protective capability required for safe reactor operation through all power ranges.

2.10.1.2 Nuc1 car and process instrumentation which senses neutron flux and process system parameters shall provide direct inputs to the control or safety systems.

2.10.1.3 Voltage monitors shall be provided to detect and indicate failures in the nuclear instrumentation.

2.10.1.4 Radiation monitors shall be provided to detect and indicate i radiation levels and to cause reactor building ventilation system isolation to excessive radiation levels.

2.10.2 Reactor Scram and Reactor Start and Withdraw Permits 2.10.2.1 A full scram signal shall cause the electric and hydraulic i

motore for all control rods to be actuated for rod insertion except as provided by Sec. 4.2.4.7.

1 2.10.2.2 A partial scram signal shall cause the electric and hydraulic scram motors for 13 preselected control rods to be actuated for rod

. insertion except as provided by Sec. 4.2.4.7. Full insertion of operable l l .

partial scram rods during power operation shall render the reactor sub- .

critical. After a partial scram, the reactor shall either be restarted within 2 hr. or all other operable rods shall be fully inserted. .

l .

( 2.10.2.3 An interlock 'shall prevent control rod withdrawal unless the l

following conditions are met:

i l

(1) The reactor start is " sealed in." .

(2) The period is longer than 7 see on both source range channels, l l and neither channel is bypassed unless a withdraw permit " seal in" .

l' l has been obtained. Bypass switch must be returned to normal (open) to regain a withdraw permit " seal in" after full or partial scram.

1 t .

t .

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_19_

(3) The period is longer than 15 nec on both intermed[ ate range cl.annels, an.! neither channel in hypa sed ualers a withdr.w permit

" scal in" has been obt ained. 3ypass switches must be retdrned to normal (open) to regain a withdraw permit " scal in" after full or partial scram.

( (4) The partial scran circuit is clear.

(5) The rod test key switch is not in " Test On" position. ,

-(6) The scram accunulators' low oil 1cyc1 and the low gas pressure -

parti.1 scraci protection are not bypassed.

2.10.2.4 Depressing the reactor start pushbutton shall seal in a reactor start permit only if the following conditions are met: -

(1) All rods are fully inserted.

(2) The building locks are secured or bypassed.

(3) Heither intermediate ranae period scram is bypas. sed.

(4) The count rate on both source range channels is above a setpoint which is at least twice the level of instrument noise and not less than -

one count per second. ,

(5) Partial scrams are cicared or bypassed. -

(6) All-rod scrams are cleared or bypassed.

).h

( ^j 2.10.2.5 A full scram shall remove the " scal in" of the reactor start circuit.

2.10.2.6 Manual scram capability shall be provided in cach scram circuit.

2.10.3 _ Nuclear Instrumentation 2.10.3.1 The nuclear instru=entation'shall provide two source range channels which extend dowa to source level.(chanacis 1 and 2), two intermediate range channels (channels 3 and 4), two linear flux channels which extend to 150 parcent rated power (channals 5 and 6), and two standard power range chanacis which extend to 150 percent rated power (channels 7 and 8). . a a "A' nuclear instrucentr .'on power range autcantic calibration sub- r system shall be provided. Tnis su'c-systea ( ucicar Ins'tre .antation Automatic Cain Control Systen) shall consist of four cocpletely independent channe .s, one for each of the four pawcr range channels, i.e., channels 5, 6, 7 and 8. Each sub-systen channel shall ocasure the reactor thernal pcuer by process ir.stru entation. Tne thermal pouer signal is then cc= pared ui th a sigani from the nuc1 car instruner.: channel and the gai, is adjusted elt.ctronically as required to r. eke the two equivalent. Tae autccatic gain adj us teent rate shall be 0.01 y,0.C01 volts per sec.

(#. 4  !

2.10.3.2 Ch..r.ne l t 3, 4, 5, and 6 shall overlap cha incts I and 2 by a mininma of ) dec.n!c. Channels 7 and 8 shall overlap channels 3 and 4 by a r..inimum of i decade.

Change No. 3

.C ~ o The nucicar instruncatation with its autonctic gain "2.10.3.3 control s.ub-systen shall be capabic of detecting and indicatin; peuer Icceis and of initiating s:ran actions

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as specified in Table 1 ar.d shall pro'aibit reactor start or control red withdrawal as specified in Sections 2.10.2.3 and 2.10.2.4."

2.10.3.4 Scram actions and interlock prchibits in the reactor start and rod withdecwal circuits shall be capabic of being initiated directly by cach of the nucicar instrumentation channels, except that the scram c.utputs of channels 5, 6, 7, and 8 shall be in 2 of 4 coine'idence above 5 percent rated poucr. When a channel in the coincidence circuit is being tested, a signal from any of the other three channels shall be capabic of initiating scram. ,

The dcccctors shall be positioned in eight horizontal instrument 2.10.3.5 tubes tangent to the reactor vessel. Detectors for each pair of ,similar intermediate and poner range channels shall ba located 180 degrees apart; detectors for chanacls 5 and 6 shall each be located 90 degrecs apart from the detectors of channels 7 and 8. Dctcetors may be placed incore during refueling and zero power operation.

2.10.3.6 Channels 7 and 8 shall respectively supply independent power-to-recirculation flow safety channels with an output current proportional

( to neutron flux. ,

2.10.3.7 Each power-to-recirculation flow safety channel shall receive a flow sinnat from two of four forccd circulation loop flow transmitters and one of two reactor power signals. The flow signals shall be added, and two-thirds of the total flow compared with the reactor power signal.

- The resultant signal shall be compared with a setpoint, manually established, to produce the required trip signal if the setpoint is exceeded. Each power-to-recirculation flow safety channel shall be completely independent

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of the other. .

. 2.10.4 Process Ins t rur.cnta tion ,

2.10.4.1 The process safety instrumentation shall be capabic of detecting and Indicating nafety pararcters and shall provide safety actions under abnorc.a1 cunditiens as rpecified in Tabic 1.

2.10.4.2 Additional process instrumentation shall be provided to accomplish control functions and to furnish information necessary for safe plant operation.

2.10.5 Radi. tion it.a t t':- ca t :t ion 2.10.5.1 !bdf..tton ..onitoring instru.acntar. ion shall be provided which is

, capabic of detecting and indicatics: rr.diation levels within the plant and in the efficent strears from the restricted area.

Change No. g

9 - "

. TABLE 4.0.2.2.1-1 -

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REACTOR SAFETY SYSTEM 9EACTOR SHUTDOWN INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT

1. Manual Reactor Trip -

Not applicable

2. Source Range, Neutron Flux ,

(Nuclear Channels 1 and 2) Not applicable-3.

Source Range, Reactor Period-Short (Nuclear Channels 1 and 2) .

>7 seconds 4.

Intermediate Range, Neutron Flux (Nuclear Channels 3 and 4) Not applicable

5. Intermediate Range, Reactor e Period-Short (Nuclear q)

Channels 3 and 4) > 3 seconds -

g:

6. Power Range, Neutron Flux
a. Reactor Power-High

< 15 ! 21 indicated -

po*er on Nuclear Channel 7 or 8 (Nuclear Channels 5 and 6) < 30% of selected scale

b. Reactor Power-High k > 15 2% indicated '

& power on Nuclear Channel 8 . 7 or 8 (Nuclear Channels  ;

A 5, ' 6, 7 and 8 with Automatic Gain Control) < 120% of RATED THERMAL FOWER ob 94

" .v ..

~ ___

TABLE 4.0.2.2.1-1. * (Continued) .- . ..

REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS..

TRIP

  • SETPOINT FUNCTIONAL UNIT
7. Reactor Pressure-High (Pressure .

- Safety Channels 1 and 2) < 1325 psig

8. Reactor Power-to-Forced Circulation Flow Abnormal, (Power-Flow Safety Channels P-(2/3) F < 48.3*

1 and 2) (Fio. 4.0.2.2.1-1)

Reactor Coolant Flowrate-Low

9. 1 30% of rated forced circula-(Power-Flow Safety Channels 1 and 2) .

" I*

(Fio. 4.0.2.2.1-1)

10. Reactor Water Level-High (Water level IIigh: < 19 inches above "0" incher.

Indicated water level.

safety channels 1 and 2 ) and Low ' Water level safety Low: < 12 inches lbelow"0" inches channels 1, 2 and 3) Indicated water level.

11. Main Condenser Vacuum-Low .

(Main Condenser Vacuum i

> m Switches 1 and 2) - 19 inches He vacuum -

y n

12. Main Steam Isolation Valves i
a. Containment Bldg. MSIV Not Fully .

I Open (Valve Closure Relays 1 and 2) > ,90% full open travel

b. Turbine Bldg. MSIV Not Fully Open (Valve Closure Relays 1 and 2) > 907, full open travel .

. I.

c. Turbine Stop Valve Not Fully Open ,

{ Valve Closure Limit Switch) >_ 90% full open travel l.

[

o l

13-2 r-

*D = % of RATED THER'tAL PONER, F = of rated forced circulation flow.

a

f3en Fic. 4.0.2.2.1-1) .

i

.c l

I

- 27ii - - I 4.0.2.2 \

LIMITING SAFETY SYSTEM SETTINGS - -

1

) I BASES FOR SECTION 4.0.2.2 '

4 . 0. 2. 2.1 REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION The Reactor Safety Systen Reactor Shutdown Instrumentation Setpoints specified in Table 4. 0. 2. 2.1 - 1 are the values at which the reictot trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and Peactor Coolant Systems are prevented from exceeding their Safety Limits.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic instru-mentation channels and provides manual reactor trip capability.

Intermediate and Source Range The Intermediate and Source Range Reactor Period Channels provide reactor core protection during reactor startup.

The Source Range Reactor Period channel provides a prohibit signal to the Control Rod Withdrawal Interlock when the reactor period is less than 7 seconds. It also provides a scram signal to the Control Rod Test Interlock when the reactor period is less than 7 seconds.

1 The Intermediate Range Reactor Period scram provides redundant protec-tion to the low range scrams of the Power Range, Neutron Flux Channels 5 and 6.

A worst possible cause of a significant power rise during startup is con-tinuous control rod withdrawal. This accident has been analyzed for an extremely conservative case: reactor initially subcritical and at ambient temperature, all reactor period and reactor power level scrams inoperative, and two control rods being withdrawn, with a reactivity insertion rate of 18d/sec. The anhlysis shows that a minimum reactor period of 0.045 seconds occurs 6.2 seconds after criticality is achieved and that reactor power will peak at 443 Mw, s 270% of RATED THERMAL POWER, shortly thereaf ter due to the Doppler effect and then decrease further as voids are created. The analysis concluded that even without scrams, fuel damage would not occur since the maximum peak fuel" temperature would be about 16850F.

).

Amendment No. t { ~4

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- 27jj -

LIMITI!!G SAPETY S'/_ STEM SPTTI!;GS BASES FO.R SSCTIO!:- 4.0.2.2 infermediate and Source Dange (Continued) -

Protectaon against this type of startup accident is provided by the Intermediate Range Acactor period and by the high reactor power scram of Po.;er IGnge Cha'nnels 5 and 6 uhile on Ipw scales.

The setpoint for Intermediate Rance channels 3 and 4 is 3 seccnd' while the setpoint for Pouer nange channels 5 and 6 is 60i of full scale. Chanr.e)s 5 and 6 are multi-range instrun,ents which the reactor operator keeps between s 201 and s 65% of full scale by range-changing as reactor power increases during the startup. For transients, such as the continuous rod withdrawal above, initiated from reactor power l'vels at the high end of the range of either nuclear instrument channels 3 (and 4) or 5 (and 6), the short period '

scram or the high reactor power scram will terminate the transient before significant heat generation occurs. For continuous control rod withdrawal transients initiated at the low end of the range, the short period scram or the high reactor power scram will terminate the transient well before the MCPR value decreases to 1.32.

Por continuous control rod withdrawal transients initiated at power levels well below the ranges of Intermediate Range channels 3 and 4 or Power Range channels 5 and 6, the short period or high >

reactor power scrans may not terminate the transient prior to the first power peak. However ..the results of the worst case analysis ,

presented above clearly show that the initial power peak does nct contain sufficient energy to damage the fuel. Damage could occur if subsequent power rise was permitted. However, the short period or high reactor power scrams would terminate the transient

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prior to this , subsequent power I;ise to prevent damage.

Credit was not taken for operation of either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified settings is required by this specification to enhance the overall reliability of the Reactor Safety System.

Power Ranoe, !!eutron Flux The Po.:er Range, I:cutron Plux channel high setpoint.provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

) .

Amendment No. tg- d

1 l - 27kk -

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a s.

i I.J P.ITI! G SAPETY SYSTEM SETTII;GS .

) 1 .

t

_ _ _ BASES F_O._R SECTIO!! 4.0.2.2 Power Rince, Heutron Flu $: - (Continue'i)

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i Below 15 1 21 of indicated power, reactor core protection is provided on a one-of-tuo basii by power Range channels 5 and

6. The active range of Channels 5 and 6 is selectable and provides a scram at 80% of full scale on the selected range.

This provides redundant protection in the louer power range for a power excursion beginning from low power.

Upon reaching 15 d 2t of indicated power, reactor core pro-tection is provided on a 2 of 4 basic by Power Range channels 5, 6, 7 and C.

The calibration of the four po.:er range nuclear instrument channels is verified using heat balance data taken during steady-state conditions. Each channel responds directly to 4 the nortron flux detected by its ion chamber. The amplifier gain for each channel is continuously corrected so that the output corresponds to the power level determined by steam flow

) because chances in the void content in the steam separators, i located in the periphery of the reactor vessel, vary neutron l attentuation, which causes erroneous nuclear instrument indi-cation. The continuous correction of the nuclear instrtutent amplifier gain is limited to a relatively, slow rate; therefore the ion chanbers cause the power , range nuclear instruments to respond directly to rapid changes in neutron flux.

The setpoint for the high reactor power sciam has been set at the lo. test practical value consistent with reactor operations and will prevent MCPR values 'less than 1.32 during operational transients.

Reactor Pre s su re-liich .

Iligh pressure in the reactor co61 ant could cause a rupture to the system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids, thus adding reactivity. -

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Amendment No. f f 9

- 2711 -

LIMIT 1?'G S AFETY SYSTP.M SF.T. T) 'G.S-HASUS FOR STCTIGN 4.0.2.2 -

Rerclor Pressure-Hioh - (Continued) 1.ctuation of the high presrure scram followed by actuation of the safety valves will prevent the pressure Safety Limit from being c.xceeded even during the maximum pressure trannient.

Reliance on the high pressure scram and safety valves is a very conservntive opnroach because of the following considar-ations - the high pressure n, cram is actually a backup for the high flux scran and additional protection is provided by one or more of the folleuine: the turbine stop valve closure partial scram, nain steam bypass valve operation, and shutdown condenser c.perat icn. The nain steam bypass valve is capable of bypassing at ! cast 100? of full load steam flow directly to the main ccndenser and has a desian maxi: sum full stroke opening tine of 3.5 seconds. 'ihe shutdown condenser has a calculated capacity of approxirately 201 of full load utcam flow.

The scrc.m will guickly reduce the neutron flux, counteracting the pressure increase by decreasing heat generation. The trip I setting is higher than the operating. pressure to permit normM operation without spurious trips. The settino provides for a wide cargin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This setpoint is effective at low power / flow condi-tions uhen the uain steam stop valve elesure scram is bypassed.

For a turbine trip under these conditions, the transient analyses show a considerable margin to the thermal hydraulic limit exists.

Reactor Power-to-Porced Ci rculat'ien Flow Abnormal and Reactor Coolant Plcwrate-I.ew The reactor power-to-forced circulation flow setpoint is specified for abnorcal fIcw conditions in rigure 4.0.2.2.1-1 of LSSS Specification

4. 0.2.2.1.

Safe operating values have been estab)1shed over a range of operating conditions for the reactor pow r level and forced circulation flow rate. The I.SSS curve:. establish the minimum allowed forced circulation flow rate at a given reactor power Icvel or, alternatively, the maximum allowed power level at a given forced circulation flow rate. These LSSS curves are a modification of the limit curve presented in Pig. 4.68 of the final Safcguar ds Jieport.

J-i Amendment No. 11 V

- 27m - }

\

bIMITING SAFFTY SYSTD; FETTII;GS DASES FOR SPC_TI,0N 4012 Reactor Pover-to-Torced Circulation Flow Abnornal and Reactor Coolant Flowrate-Low - (Continued)

The bases of the LSSS curve are the burnout and flow instability safety margin. Thus, the LSSS curve establishes a limiting relationship between the fuel element heat flux at a given power level and the forced circulation flow rather than a limiting relationship between actual reactor power and forced circulation flow. Under steady-state operating conditions, the heat flux and reactor power relations would be equivalent. During transients ,

however, heat flux and power level are not equivalent, due to the thermal lags of the oxide fuel. . .

The LSSS curve has not been extended to forced circulation flous of less than 30 percent of full flow, which is the low flow scram set point and which is well above the hydraulic instability flow.

Scram circuitry initiates a reactor scram whenever the relation-

) ship betueen the power level and the forced circulation flow rate corresponds to the LSSS curve. The power-flow scram signal is represented by the equation:

P2-F=3 48.3 Where P = reactor power level, t of RATED THERMAL POWER.

F= recirculation flow rate, t of full flow,.10.75x10 0 lb/r Thus, a scram will occur whenever the value of [P - (2/3)P]

exceeds 48.3.

A low flow scram sicnal is set at 30 percent of full flow.

This low flow scran signal compels the operator to establish a minimum flow before reactor startups and it prevents oper-ation in a low flow range.

t

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4 Amendment fio.i 7

- -32dd-POWER DISTRIBUTION LIMITS

) BASES FOR SECTIONS 4.2.4.2 and 5.2.17 The daily requirement for surveillance of the core APLHGR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core LHGR after power increases > 15% of RATED THERMAL POWER will assure that significant increases in APLHGR are determined.

4.1.4.1.1 and 5.2.17.2 THERMAL POWER-RECIRCULATION FLOW RELATIONSHIP The THERMAL POWER-recirculation flow limiting condition for steady-state operation has been conservatively set below the LSSS curve presented in Figure 4.0.2.2.1-1 of Specification 4.0.2.2.1. This specification also requires that the reactor not be operated in a steady-state condition above the RATED THERMAL POWER, 165 MWth, authorized in the NRC license for the facility. The limiting minimum flow setpoint of 30% of rated recirculation flow is sub-stantially above the natural circulation flow and the flow at which hydraulic instability occurs. A ratio of > 1.6 exists be-tween the low-flow scram setpoints and the instability-natural ,

circulation flow. Therefore, adequate protectio n of the core ]

against flow and core instability exists over the full power range l

^

of anticipated reactor operation as limited by the THERMAL POWER- 1

- 1 recirculaticn flow operating region.

The low flow limitation requires the establishment of a minimum flow of 30% of rated recirculation flow before reactor startup.

Operation in a steady state condition < 68. 3% of RATED THERMAL POWER at the minimum 30% of RATED RECIRCULATION FLOW assures that the CPR remains above the minimum allowable value of 1.32 during J an abnormal reactor transient (recirculation flow speedup is the most limiting). The steady state limiting CPR values corresponding to other operating conditions bound by the Power-Flow LSSS curve reported in Fig. 4.0.2.2.1-1 are defined in Fig. 4.2.4.2.3-1.

The daily requirement for surveillance of the power to recirculation flow relation is sufficient since this relation shifts very slowly when there have not been significant power, flow or control rod

changes. The surveillance of this relation after power increases l > 15% of RATED THERMAL POWER will assure that significant changes In the relation are determined.

4.2.4.2.3 and 5.2.17.3 MINIMUM CRITICAL POWER RATIO

! The required operating limit MINIMUM CRITICAL POWER RATIO (MCPR) at i steady-state operating conditions as established in Specification 4.2.4.2.3 is derived from an analysis of abnormal operational trans-l ients with the transient CRITICAL POWER RATIO > 1.32. The CPR l pg I

1 l Amendment No. JJ ,16 l

1

3

- 35 - -

Safety instrucentation full scran protcetion shall always be through redur. dant channels, except that one channel may be bypassed, ac

_' permitted by Table 1, provided thrt the time during which redundant protection frec any one-out-cf-tuo or one-out-of three system is ,not l )

provided shall not exceec 24 nr in any 30-day interval.

4.2.6.2 The setpoints for the safety instrumentation shall be as

  • 1 specified in Table 4.0.2.2.1-1 and Table 1.

4.2.6.3 Key switches shall permit eparational, maintenance, and test bypass of the safety instru=cntation only with the approv-21 of the Shift Supervisor, and in accordance with the provisicar of Table 1 4ad the following conditions:

. . . (1) The requirement of a full scram cignal to par =it operation of the boron injectica system may be bypasted to permit testin; of that system.

(2) The red withdrau.1 2 prohibit action of both source ran;e nucicar 1

channels may be bypscs,ed only at power levele above the normal range of these channels.

(3) The rod withdrausi prohibit action of both intermediate '

range nuclear channels may be bypassed caly when reactor rouer exceeds 3 Mwt.

(4) The " building locks secured" inter 1cck of the reactor start circuit may be bypassed when thera is no fucl in the recctor or when testing or repairing this intericek function, provided the requiremente.

of Sec. 4.2.1.1.are met. -

(5) The rod withdrausi permit cir;uit c;y be bypassed to permit test withdrawal of a control red in accordanca with Sec. 2.8.6.

4.2.6.4 A control rod shall net be withdrawn unless the conditions of Sec. 2.10.2.3'or Sec. 2.8.6 are met.

4.2.6,5 It shall not be possible to see.1 in the reactor start per=it

circuit unles s the ccnditions of Sac. 2.10.2.4 are met.

1 4.2.6.6 During refeelin; or ot.ker changes in core configuration invoirin; fuel element or contrcl red rer.cvel or intertion, the core shall be monitored by a minimum of two neutron count deteccces.

1 AmendmentNo./,11 I

- - - - - - ~- - ---- - - "

__ _. _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ . _ - _ _ . _ _ . =_ _ . - - _ _ . - _ _ - - _ _ . _ _ -

, a .

. 4.2.6.7 During operation with the nuclear instrumentation channels in 2 of 4 trip logic, at least three of nuclear channels 5, 6, 7 and 8, including their

}

automatic gain control subsystem channels, shall be OPERABLE. If nuclear ,,

channel 5 or 6 is inoperable, its scram contacts shall be placed in the trip

  • j position. If power is de-escalated, the tripped channel's output shall be 1 bypassed prior to entering 1 of 2 logic, subject to Section 4.2.6.1 require-l- sents. If nuclear channel 7 or 8 is inoperable in a manner af fecting the . -

!- operability of its corresponding power-flow chanr.el, the power-flow channel - -

I shall be bypassed, pursuant to the time limitations of Section 4.2.6.1, and . -

the scras contacts of the nuclear channel shall be placed in the trip position.

4.2.6.8 Safety channels directly backed up by an identical channel or- -

.::r - - channels may be bypassed for maintenance or testing. Safety channels in the, - ---

partial scram circuit may be bypassed for maintenance or testing for up to 24 4

hours.

i

'4.2.6.9 Both reactor forced circulation pumps shall be automatically shut down by a high reactor pressure signal or by a low reactor water level signal.

i 1 Bases - The RPTS is a diverse and independent backup except for common current

sensing loops to the normal scram system for rapid shutdown of the reactor.

To protect the primary system from an ATWS event in which either MSIV closes at power, thus eliminating the main condenser as a heat sink, the recirculation pumps must be shut down to prevent damage to the primary system due to high I pressure. A rapid shut down of the recirculation pumps has the effect of

) causing an increase in the moderator voids in the reactor core. A substantial negative reactivity results and the power and pressure surges that might j otherwise occur in the most limiting transient (MSIV closure) are substantially

) reduced. With the recirculation pumps shut down, the reactor power will be reduced to a steady state power level of less than 20% (based on natural j circulation through the core).

4.2.7 - peteted -

4 I

I I

4 d

j i

36 ChangeNo./,/,

Amendment No. 4 I ,36 i

- . _ - . _ _ - , ~ _ _ _ , _ _ . . . - - - - _ _ . _ _ _ _ , _ , , _ . , _ . _ _ _ . - . - _ _ . . _ . . _ _ _ _ . _ _ _ _ - _ - _ .

E- . . ,

MINIMUM FREQUENCIES FOR TESTING, CALIHRATING, AND/OR CHECKINC OF INSTRUMENTATION 3

Channels Action Minimum Frequency

1. Reactor Water Level Calibration At each refueling shutdown.

Test

  • Monthly when in service j and prior to each reactor startup if test has not been performed within 30 days.

Check Daily. -

l 1

2. Reactor Pressure Calibration At each refueling shutdown.

Test

  • Monthly when in service and prior to each reactor j startup if test has not been performed within 30 days.

Check Daily.

3. Reactor Power - Flow Calibration At each refueling shutdown.

Test

  • Monthly when in service and prior to each reactor startup if test has not been performed within 30 days.

Check Daily. j 1

4. Reactor Coolant Flow Calibration At each refueling shutdown.

Rate Low Test

  • Monthly when in service and prior to each reactor startup if test has not been performed within 30 days. .

l Check Daily. I

5. Source Range Test Prior to each reactor (Channels 1 and 2) (60 cycles startup if test has-not been per sec.) performed within 30 days.

Check Once per shift when in service.

)

5-7 ChanbeNo. Amen ment N

MINIMUM FREQUENCIES FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd)

Minimum Frequency

) Channels Action  :

Intermediate Range Test (10-10 and Prior to each reactor

6. 10-5 amps; start up if test has not Channels 3 and 4) been performed within period *)

30 days.

Check Once per shift when in service.

Check by heat Monthly when in service.

7. Wide Range and Power Range (Channels 5, balance 6, 7, and 8) Monthly when in service and Test
  • prior to each reactor A. Nuclear Instru-
  • startup if test has not been mentation & Auto- performed within 30 days.

matic Cain' Control 'l Sub-System.

Check Once per shift when in B. ITuclear Instru- service.

mentation & Auto-matic Gain Control ,

Sub-System.

Calibration At each refueling outage.

) C. Automatic Gain Control Sub- l i

' System NOTE: Testing of the Nuclear Instrumentation and Automatic Gain Control Sub-System shall be done concurrently.

] once a month,

8. Full Scram Circuits Test for hot short by means of built-in test switch.

Calibration At least once per 18 months.

l

9. Area Radiation Monitors Quarterly Test Check Daily il l

Amendment No. 36,49 5-8 1

i

v s v. . -. .

s TABLE 1  :

OPERATING LIMITS S

KEYSWITCH BYPASS

' CONDITION CllANNEL OR SENSOR SET POINT ACTION PROVISION

1. Reactor Power Two of four nuclear Table 4.0.2.2.1-1 Full scram None liigh channels 5, 6, 7 &

8 if power level is

)5% of full power Either nuclear Table 4.0.2.2.1-1 Full scram One channet may be by-channel 5 or 6. if passed for calibration

. power level is <5% and testing.

of full power Y 1) Both channels may be lI 2. Reactor Period Nuclear channel Table 4.0.2.2.1-1 Full scram Short 3 or 4 bypassed only when reactor power exceeds 3 Mwt.

2) One channel may be bypassed for cali-bration and testing.

Pressure safety < 1325 psig 1) Full scram One channel may be by-

3. Reactor channel 1 or 2 2) Shutdown condenser passed for calibration Pressure ope ra t es . and testing.

l[

m Iligh

3) Closure of venti-lation inlet and s

lf outlet dampers

4) Closure of con-tainment off gas

$E vent header valve. .

31

.s .

4 4 ..

OS

';?o e

. ~e 9 g

.. I TABLE 1 - OPERATING LIMITS - (cont'd) ,

i KEYSWITCH BYPASS '

CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION

4. Reactor Power- Power-flow safety Table 4.0.2.2.1-1 Full scram One channel may be by-Flow Rate channel 1 or 2 passed for calibration .

Abnormal and testing.

4

5. Reactor Power-flow safety Table 4.0.2.2.1-1 Full scram One channel may be by-Coolant Fiou channel 1 or 2 passed for calibration Rate Low and testing.
6. Reactor Water Water level safety Table 4.0.2.2.1-1 Full scram One channel may be by-Level High channel 1 or 2 passed for calibration and testing.

(Nominal indicated unvoided saturated water level shall be permitted to vary Y' f ruta 2'9" above the fuel to up 4'6" above the fuel during reactor heatup and operation.

G

7. Reactor Water Water level safety <l2" below nominal 1) Full scram One channel of Item 7 Level Low channel 1 or 2 indicated level 2) Initiation of high or channel 3 of Item 7A pressure core may be bypassed for spray pumps calibration and testing.
3) closure of reactor building steam isolation valve jf and its bypass

@ 4) Closure of reactor

@f binwdown through g decay heat removal r' valve ll

5) Start IA and la diesel generators 3%- 6) Closure of shutdown condenser conden-

)k sate drain valve .-

m Ws

December 19, 1986 In reply, please refer to LAC-12019 DOCKET NO. 50-409 ENCLOSURE B Proposed Technical Specifications 2.8.4 The drive mechanisms shall have individual scram accumulators which drive the hydraulic motors and which shall be connected to the gas charging system and to the hydraulic charging system through individual check valves.

2.8.5 Each control rod shall be decelerated near the end of its scram stroke by an internal hydraulic dashpot action.

2.8.6 Deleted.

2.8.7 Sustained loss of voltage to Reactor Building Motor Cont rol Center lA (which supplies the hydraulic charging pumps and the control rod drive motors), or low gas pressure, or a low level of hydraulic fluid in any scram accumulator shall initiate safety action as specified in Table 1.

2.9 C0NTROL SYSTEMS 2.9.1 The power level of the reactor shall be adjusted manually by the positioning of control rods or by changes in forced circulation pu.np speed, or automatically by pressure controlled changes in the forced circulation pump speed.

2.9.2 The position of the scoop tubes which control the forced circulation pump speed shall be controlled maually or automatically by the pressure control system.

2.9.3 Deleted.

2.9.4 An Initial Pressure Regulating System shall assume control of turbine inlet valves at low pressures so as to maintain turbine inlet pressure within a controlled band. The high pressure set point of the control band shall normally be set below the operating set point of the forced circulation pump automatic control system. To maintain rated turbine inlel pressure, the Initial Pressure Regulating System may be used without use of the forced circulation pump automatic control systen.

Change No.pyf;/f 4 I

s

l 2.9.5 Reactor water level shall normally be controlled automatically by a three-element control system which senses unvoided reactor water level, feedwater flow, and main steam flow, and which adjusts feedwater pump speed by positioning scoop tubes within fluid couplings between the feedwater pumps and their motors.

2.10 SAFETY AND MONITORING SYSTEMS 2.10.1 General ,

2.10.1.1 Deleted.

2.10.1.2 Deleted.

2.10.1.3 Deleted.

2.10.1.4 Radiation monitors shall be provided to detect and indicate radiation levels and to cause reactor building ventilation system isolation at excessive radiation levels.

2.10.2 Reactor Scram and Reactor Start and Withdraw Permits 2.10.2.1 A full scram signal shall cause the electric and hydraulic motors for all control rods to be actuated for rod insertion except as provided by Sec. 4.2.4.7.

2.10.2.2 A partial scram signal shall cause the electric and hydraulic scram motors for 13 preselected control rods to be actuated for rod insertion except as provided by Sec. 4,.2.4.7. Full insertion of operable partial scram rods during power operation shall render the reactor suberitical. After a partial scram, the reactor shall either be restarted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or all other operable rods shall be fully inserted.

2.10.2.3 Deleted, l

2.10.2.4 Deleted.

-2.10.2.5 Deleted.

2.10.2.6 Deleted.

2.10.3 Deleted.

/

l 2.10.4 Process Instrumentation 2.10.4.1 The process safety instrumentation shall be capable of detecting and indicating safety parameters and shall provide safety actions under abnormal conditions as specified in Table 1.

2.10.4.2 Additional process instrumentation shall be provided to accomplish control functions and to furnish information necessary for safe plant operation.

2.10.5 Hadiation Instrumentation 2.10.5.1 Radiation monitoring instrumentation shall be provided which is capable of detecting and indicating radiation levels within the plant and in the effluent streams from the restricted area.

Change No.,d/

Amendment No. ___

4

- 27bb -

TABLE 4.0.2.2.1-1 REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS E

FUNCTIONAL UNIT TRIP SETPOINT

1. Manual Reactor Trip Not applicable j- 2. Source Range, Neutron Flux Not applicable (Nuclear Channels 1 and 2)-

l

3. Source Range, Reactor > 7 seconds Period-Short (Nuclear channels 1 and 2)
4. Wide Range, Neutron- 1 120% of Rated Thermal Power 4

Flux (Nuclear Channels 3, 4,.5, and 6)

5. Wide Range, Reactor Margin > 3 seconds (Period) (Nuclear channels 3, 4, 5, and 6)
6. Reactor Power-to-Forced Circulation - P-(2/3)F 148.3
  • Flow Abnormal, .

(Neutron Flux Channels 3, 4, 5, and 6) (Fig. 4.0.2.2.1-1) 4 4

i

  • P =
  • of RATED THERMAL POWER, F = % of rated forced circulation flow.

(See Fig. 4.0.2.2.1-1) 4 i

j. i 2

(

Amendment No. %

,- .ve,-,,5e ye---m,..--,-----, n y .. ...--,.,w-y,nw.,.-,.-- w .c . ,- g r e -e. ww, , .-,,t., ,,,,,-~.,+,--y,,.,-,,e%-e,.m.,.- . .- - . - - - - , - . , ~ - , - ~ - - - ,

- 27cc -

TABLE 4.0.2.2.1-1 (cont'd)

REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT

7. Reactor Pressure-High (Pressure I 1325 psig

-Safety Channels 1 and 2)

8. Intentionally Blank-
9. Reactor Coolant Flowrate-Low 1 30% of rated forced (Eecirculation Flow Safety Channels I circulation flow and 2) (Fig. 4.0.2.2.1-1)
10. Reactor Water-Level-High (Water High: I 19 inches above level safety channels 1 and 2) "0" inches indicated and Low (Water level safety water level channels 1, 2 and 3) Low: 1 12 inches below "0" inches indicated water level
11. Main condenser Vacuum-Low 2 19 inches'Hg vacuum (Main Condenser Vacuum Switches 1 and 2) 12.. Main Steam Isolation Valves
a. Containment Bldg. MSIV Not Fully Open (Valve Closure Relays 1 & 2) 1 90% full open travel
b. Turbine Bldg..MSIV Not Fully Open 1 90% full open travel (Valve closure Relays 1 and 2)
c. Turbine Stop Valve Not Fully Open > 90% full open travel

. (Valve Closure Limit Switch) l l

i-i-

l. '

t i

f AmendmentNo.,LPI L

- 27ii -

4.0.2.2 LIMITING SAFETY SYSTEM SETTINGS BASES FOR SECTION 4.0.2.2 4.0.2.2.1 REACTOR SAFETY SYSTEM REACTOR SRUTDOWN INSTRUMENTATION SETPOINTS The Reactor Safety System Reactor Shutdown Instrumentation Setpoints specified in Table 4.0.2.2.1-1 are the values at which the reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant Systems are prevented from exceeding their Safety Limits.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic instrumentation channels and provides manual reactor trip capability.

Neutron Flux channels The Source Range Reactor Period Channels provide reactor core protection during reactor startup.

The Source Range Reactor Period channel provides a prohibit signal to the Reactor Start Interlock when the reactor flux level is not clearly discernible above the channel noise level. The channel provides a prohibit signal to the Rod Withdraw Interlock when the reactor period is less than 7 seconds. It also provides a scram signal to the Control Rod Test Interlock when the reactor period is less than 7 seconds when in Rod Test Mode.

The Margin scram provides protection in the low ranges of the Wide Range Neutron Flux Channels 3, 4, 5, and 6 in 2-of-4 logic in redundant scram strings. The Margin scram is effective on either high rate of change of flux 1 or excessively high flux level transients. This trip is bypassed when reactor nower is above 15%.

A worst possible cause of a significant power rise during startup is continuous control rod withdrawal. This accident has been analyzed for an extremely conservative case: reactor initially subcritical and at ambient temperature, all reactor period and reactor power level scrams inoperative, and two control rods being withdrawn, with a reactivity insertion rate of 18 cents /sec. The analysis shows that a minimum reactor period of 0.045 seconds occurs G.2 seconds after criticality is achieved and that reactor power will peak at 443 Mw, ~270% of RATED THERMAL POWER, shortly thereafter due to the Doppler effect and then decrease further as voids are created.

The analysis concluded that even without scrams, fuel damage would not occur since the maximum peak fuel temperature would be about 16850F.

s Amendment No. ,rl, ___

- 27jj -

LIMITING SAFETY SYSTEM SETTINGS BASES FOR SECTION 4.0.2.2 Neutron Flux Channels - (cont'd)

Protection against this type of startup accident is provided by the Wide Range Reactor margin trip and by the high reactor power scram.

The setpoint for the margin scram corresponds to a 3 second period while the setpoint for Wide Range flux scram is 120% of indicated power. For transients, such as the continuous rod withdrawal above, initiated from reactor power levels at any point in the range of the wide range nuclear instrument, the margin scram or the high reactor power scram will terminate the transient before significant heat generation occurs. For continuous control rod withdrawal transients initiated at the low end of the range, the margin scram or the high reactor power scram will terminate the transient well before the MCPR value decreases to 1.32.

For continuous control rod withdrawal transients initiated at power levels well below the ranges of Wide Range channels 3, 4, 5, and 6, the margin or high reactor power scrams may not terminate the transient prior to the first power peak, llowever, the results of the worst case analysis presented above clearly show that the initial power peak does not contain sufficient energy to damage the fuel. Damage could occur if subsequent power rise was permitted. However, the margin or high reactor power scrams would terminate the transient prior to this subsequent power rise to prevent damage.

Credit was not taken for operation of the Source Range Channels or a 3 second period scram in the accident analyses; however, their functional capability at the specified settings is required by this specification to enhance the overall reliability of the Reactor Safety System.

The Wide Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid for negative reactivity feedback from increases in moderator temperature and voids to be effective.

AmendmenL No. )(~ __

- 27kk -

LIMITING SAFETY SYSTEM SETTINGS BASES FOR SECTION 4.0.2.2 Reactor core protection is provided on a 2-of-4 basis by Wide Range channels 3, 4, 5, and 6.

The calibration of the four wide range nuclear instrument channels is verified using heat balance data taken during steady-state conditions. Each channel respondo directly to the neutron flux detected by its ion chamber.

An amplifier in each channel continuously corrects the neutron flux signal so that the output corresponds to the power level determined by steam flow.

This is necessary due to continuously changing void content in the steam separators located in the periphery of the reactor vessel which affects the neutron flux to the detectors and gives an erroneous flux signal. The continuous correction of the nuclear instrument amplifier gain is limited to a relatively slow rate; therefore, the ion chambers cause the power range nuclear instruments to respond directly to rapid changes in neutron flux.

The setpoint for the high reactor power scram has been set at the lowest practical value consistent with reactor operations and will prevent MCPR values less than 1.32 during operational transients.

Reactor pressure-High High pressure in the reactor coolant could cause a rupture to the system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids, thus adding reactivity.

Amendment No. g ___

4

- 2711 -

LIMITING SAFETY SYSTEMS SETTINGS l BASES FOR SECTION 4.0.2.2 Reactor pressure-High - (cont'd)

Actuation of the'high pressure scram followed by actuation of the safety valves will prevent the pressure Safety Limit from being exceeded even during

.the maximum pressure transient. Reliance on the high pressure scram and safety valves'is a very conservative approach because of the following considerations -.the high pressure scram is actually a backup for the high flux scram and additional protection is provided by one or more of the following: the turbine stop valve closure partial scram, main steam bypass valve operation, and shutdown condenser operation. The main steam bypass valve is capable of bypassing at least 100% of full load steam flow directly to the main condenser and has a design maximum full stroke opening time of 1.5 seconds. The shutdown condenser has a calculated capacity of approximately 20% of full load steam flow.

The scram will quickly reduce the neutron flux, counteracting the pressure increase by decreasing heat' generation. The trip setting is higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This setpoint is effective at low power / flow conditions when the main steam stop valve closure scram is bypassed. For a turbine trip under these conditions,

-the transient analyses show a considerable margin to the thermal hydraulic

' limit exists.

Reactor power-to-Forced Circulation Flow Abnormal and Reactor Coolant Flowrate-Low The reactor power-to-forced circulation flow setpoint is specified for abnormal flow conditions in Figure 4.0.2.2.1-1 of LSSS Specification 4.0.2.2.1.

Safe operating values have been established over a range of operating j conditions for the reactor power level and forced circulation flow rate. The LSSS curves establish the minimum allowed forced circulation flow rate at a j

given reactor power level or, alternatively, the maximum allowed power level

~

nt n given forced circulation flow rate. These LSSS curves are a modification of the limit curve presented in Fig. 4.60 of the Final Safety Analysis Report.

1 j' i 1-AmendmentNo.,}YI___

L -

- 27mm -

LIMITING SAFETY SYSTEM SETTINGS BASES FOR SECTION 4.0.2.2 Reactor power-to-Forced Circulation Flow Abnormal and Reactor Coolant Flowrate-Low - (cont'd)

The bases of the LSSS curve are the burnout and flow instability safety margin. Thus, the LSSS curve establishes a limiting relationship between the fuel element heat flux at a given power level and the forced circulation flow rather than a limiting relationship between actual reactor power and forced circulation flow. Under steady-state operating conditions, the heat flux and reactor power relations would be equivalent. During transients, however, heat flux and power level are not equivalent, due to the thermal lags of the oxide fuel.

The LSSS curve has not been extended to forced circulation flows of less than 30 percent of full flow, which is the low flow scram set point and which is well above the hydraulic instability flow.

The Power to Flow comparison is performed in the 4 Wide Range Flux Monitors and the scram circuitry in a 2-of-4 logic initiates a reactor scram whenever the relationship between the power level and the forced circulation flow rate corresponds to the LSSS curve. The power-flow scram signal is represented by the equation:

P - 2 F I 48.3 3

Where P = reactor power level, % of RATED THERMAL POWER F = recirculation flow rate, % of full flow, 10.75 x 106 lb/hr Thus, a scram will occur whenever the value of (P - (2/3)F] exceeds 48.3.

A low flow scram signal is set at 30 percent of full flow. This low flow scram signal compels the operator to establish a minimum flow before reactor startups and it prevents operation in a low flow range.

Amendment No. % ___

- 27rr -

S AFETY LT_M,ITS AND LIMITING SAFETY SYSTEM SETTINGS 4/5.0.2.3. NUCLEAR INSTRUMENTATION LIMITING CON _DI_TIONS FOR OPERATION 4.0.2.3.1 The source range nuclear instrumentation system shall be OPERABLE with a minimum of 2 channels. Setpoints are given in Table 4.0.2.2.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 2*, 3, 4 and 5.

  • Required OPERABLE only until 1/2 decade overlap is achieved on at least 2 Wide Range Monitors.

ACTION:

3 1. With one channel inoperable:

a. In OPERATIONAL CONDITION 2, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER.
b. In OPERATIONAL CONDITION 3 or 4, place Control Power key in off within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
c. In OPERATIONAL CONDITION 5, suspend all operations involving i

CORE ALTERATIONS within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

2. With more than one channel inoperable:
a. In OPERATIONAL CONDITION 2, be in at least Hot Shutdown within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, with Control Power off.
b. In OPERATIONAL CONDITION 3 or 4, immediately place Control Power off.

i'

c. In OPERATIONAL CONDITION 5, immediately suspend all operations involving CORE ALTERATIONS.

l SURVEILLANCE _RE0UIREMENTS ,_

I 5.0.2.3.1 The Source Range nuclear instrumentation shall be demonstrated OPERABLE:

l

  • a. By performing a CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
  • b. By performing a CHANNEL FUNCTIONAL TEST at least once per 31 days.
c. By performing a CHANNEL CALIBRATION at least once per 18 months.

1

  • Neutron detectors may be excluded.

l WPl.3.7 i

1

- 27ss -

SAFETY _ _ LIMITS AND LIMITING SAF1U SY_ STEM SETTIN,C_S 4/5.0.2.3 NUCLEAR INSTRWHENTATION (cont.)

LIMITING CONDITIONS FOR OPERATION 4.0.2.3.2 The wide range nuclear instrumentation system

  • shall be OPERABLE with a minimum of 4 channels. Setpoints are given in Table 4.0.2.2.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: In OPERATIONAL CONDITION 1 or 2:

1. With one channel inoperable; power operation may continue.
2. With two channels inoperable; be in at least Hot Shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
  • Includes automatic gain correction, power to flow comparison, and margin computation sub systems.

SURVEILLANCE REQUIRE.MENTS _,

5.0.2.3.2 The wide range nuclear instrumentation

  • shall be demonstrated OPERABLE:
a. By performing a CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. By performing a CilANNEL FUNCTIONAL TEST:
1. Prior to each reactor startup if not performed within the past 31 days.
2. At least once per 31 days.
c. By performing a CilANNEL CALIBRATION at least once per 18 months.
d. By performing a heat balance at least once per 31 days.
  • Neutron detectors may be excluded.

WPl.3.7

- 32dd -

POWER DISTRIBUTION LIMITS BASES F0D SECTIONS 4.2.4.2 AND 5.2.17 The daily requirement for surveillance of the core APLilGR above 25% of RATED TIIERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core LIIGR after power increases > 15% of RATED TIIERMAL POWER will assure that significant increases in APLIIGR are determined.

4.2.4.2.2 and 5.2.17.2 TIIERMAL POWER-RECIRCULATION FLOW RELATIONSilIP The TIIEINAL POWER-recirculation flow limiting condition for steady-state operation has been conservatively set below the LSSS curve presented in Figure 4.0.2.2.1-1 of Specification 4.0.2.2.1. This specification also requires that the reactor not be operated in a steady-state condition above the RATED TREINAL POWER,165 MWth, authorized in the NRC license for the facility. The limiting minimum flow setpoint of 30% of rated recirculation flow is substantially above the natural circulation flow and the flow at which hydraulic instability occurs. A ratio of > 1.6 exists between the low-flow scrtun setpoints and the instability-natural circulation flow.

Therefore, adequate protection of the core against flow and core instability exists over the full power range of anticipated reactor operation as limited by the TIIEINAL POWER-recirculation flow operating region.

The low flow limitation requires the establishment of a minimma flow of 30%

of rated recirculation flow before reactor startup. Operation in a steady state condition i G8.3% of RATED TIIEINAL POWER at the minimum 30% of RATED RECIRCULATION FLOW assures that the CPR remains above the minimum allowable value of 1.32 during an abnormal reactor transient (recirculation flow speedup is the most limiting). The steady state limiting CPR values corresponding to other operating conditions bound by the Power-Flow LSSS curve reported in Fig. 4.0.2.2.1-1 are defined in Fig. 4.2.4.2.3-1.

The daily requirement for surveillnnce of the power to recirculation flow relation is sufficient since this relation shifts very slowly when there have not been significant power, flow or control rod changes. The surveillance of this relation af ter power increases 1 15% of RATED TIIEINAL POWER will nasure thnt significant changes in the relation are determined.

4.2.4.2.3 und 5.2.17.3 MINIMUM CRITICAL POWER RATIO The required operating limit MINIMlN CRITICAL POWER RATIO (MCPR) at steady-state operating conditions as established in specification 4.2.4.2.3 is derived from nn unalysis of abnormal operational transients with the transient CRITICAL POWER RATIO 2 1.32. The CPR Amendment No. % g _

Safety instrumentation full scram prctection shall always be through redundant channels, except that one channel may be bypassed, as permitted by Table 1, provided that the time during which redundant protection from any one-out-of-two or one-out-of-three system is not provided shall not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day interval, or as otherwise specified in these Technical Specifications.

4.2.G.2 The setpoints for the safety instrumentation shall be as specified in Table 4.0.2.2.1-1 and Table 1.

4.2.G.3 Key switches shall permit operational, maintenance, and test bypass of the safety instrumentation only with the approval of the Shift Supervisor, and in accordance with the provisions of Table 1 and the following conditions:

(1) The requirement of a full scram signal to permit operation of the boron injection system may be bypassed to permit testing of that system.

(2) The rod withdrawal prohibit action of both source range nuclear channels may be bypassed only at power levels above the normal range of these channels. l (3) The " building locks secured" interlock of the reactor start circuit may be bypassed when there is no fuel in the reactor or when testing or repairing this interlock function, provided the requirements of Sec.

4.2.1.1 are met.

(4) The rod withdrawal permit circuit may be bypassed to permit test withdrawal of a control rod.

4.2.G.4 A control rod shall not be withdrawn unless the conditions of the rod withdrawal interlock or control rod test are met.

4.2.6.5 It shall not be possible to seal in the reactor start permit circuit unless the conditions of reactor start interlock are met.

4.2.G.G Deleted.

Amendment No. f( f ___

4.2.6.7 Deleted.

4.2.6.8 Safety channels directly backed up by an 2dentical channel or channels may be bypassed for maintenance or testing. Safety channels in the partial scram circuit may be bypassed for maintenance or testing for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.2.6.9 Both reactor forced circulation pumps shall be automatically shut down by a high reactor pressure signal or by a low reactor water level signal.

Bases - The RpTS is a diverse and independent backup except for common current sensing loops to the normal scram system for rapid shutdown of the reactor. To protect the primary system from an AThS event in which either MSIV closes at power, thus eliminating the main condenser as a heat sink, the recirculation pumps must be shut down to prevent damage to the primary system due to high pressure. A rapid shutdown of the recirculation pumps has the effect of causing an increase in the moderator voids in the reactor core. A substantial negative reactivity results and the power and pressure surges that might otherwise occur in the most limiting transient (MSIV closure) are substantially reduced. With the recirculation pumps shut down, the reactor power will be reduced to a steady state power level of less than 20% (bar,ed on natural circulation through the core).

4.2.7 - Deleted -

l ChangeNo.<t{,21___( ___

AmendmentNo.J2$JM

MINIMUM FREQUENCIES FOR TESTIN(., CALIBRATING,'AND/OR CHECKING OF INSTRUMENTATION

CHANNELS  : ACTION  : MINIMIN FHEQUENCY  :

l l l l1. Reactor Water Level : Calibration  : At each refueling shutdown. l

: Test
  • l Monthly when in service and prior l to each reactor startup if test  :
:  : has not been performed within 30 .:
:  : days. l l l l l l l Check l Daily.  :

!  ! l l l l l l l2. Reactor Pressure l Calibration  : At each refueling shutdown.  :

:  : l-l l Test
  • l Monthly when in service and prior :
.  : to each reactor startup if test  :
: has not been performed within 30 l l l  ! days. l l l l l
.  : Check  : Daily. l
l l l l3. Deleted  :  :

l l  :  :

4. Reactor Coolant Flow: Calibration l At each refueling shutdown. l l Rate Low  :  :

l l Test

  • l Monthly when in service and prior l l l l to each reactor startup if test  :
:  : has not been performed within 30 :

l l days. l l l l l l l Check l Daily. l l l l l l l l l

5. Deleted  : l- l l l l l l l l  !

Change No. ,36*,

Amendment No. k [____

S-7

b

CIIANNELS  : ACTION  : MINIMUM FREQUENCY  :

l l l l l6. Deleted  : l l l l l l

17. Deleted l l l l l l l l l l l l8. Full Scram Circuits : Test for hot : Once a month.  :
short by means!  !

l l of built-in l l l l test switch. l l l l l l

9. Area Radiation l Calibration At least once per 18 months.

Monitors l l l l l Test  : Quarterly.

l l l l l l Check Daily. l l l l l ChangeNo.f){__[

Amendment No. % A*

S-8

TABLE l~

OPERATING LIMITS KEYSWITCH CONDITION CHANNEL OR SENSOR SET POINT ACTION BYPASS PROVISION

1. Reactor Power Nuclear Channels Table 4.0.2.2.1-1 Full Scram May render channel High 3, 4, 5, and 6 from 2 of 4 inoperable per LCO Channels 4.0.2.3.2
2. Reactor Margin Nuclear Channels 3, Table 4.0.2.2.1-1 Full Scram
  • May render channel (Period) 4, 5 and 6 from 2 of 4 inoperable per LCO Channels 4.0.2.3.2 T
3. Reactor Pressure Pressure Safety < 1325 psig 1) Full Scram One channel may be High Channel 1 or 2 2) Shutdown Condenser bypassed for Operates calibration and
3) Closure of testing.

Ventilation Inlet and Outlet Dampers g 4) Closure of g Containment Off-Gas

g. Vent Hender Valve o

5 4. Reactor Power-Flow Nuclear Channels Table 4.0.2.2.1-1 Full Scram

  • May render channel n: Abnormal 3, 4, 5 and 6 from 2 of 4 inoperable per LCO Channels 4.0.2.3.2 to
  • Reactor Margin scram is in effect at low powers and Power / Flow scram is in effect at high powers with automatic

-\ switching of the mode at a user selected value in the range of 1 to 15% rated power.

KEYSWITCH CONDITION CHANNEL OR SENSOR SET POINT ACTION BYPASS PROVISION

5. Reactor Coolant Recire. Flow Safety Table 4.0.2.2.1-1 Full Scram One channel may be Flow Hate Low Channel 1 or 2 from 1 of 2 bypassed for channels calibration and testing.

G. Reactor Water Water Level Safety Table 4.0.2.2.1-1 Full Scram One channel may be Level High Channel 1 or 2 bypassed for calibration and testing.

(Nominal indicated unvoided saturated water level shall be permitted to vary from 2'9" above the fuel to 4'6" above the fuel during reactor heatup and operation. )

7. Reactor Water Water Level Safety <12" below nominal 1) Full Scram One channel of Item 7 Level Low Channel 1 or 2 indicated level 2) Initiation of High or Channel 3 of Y' Pressure Core Item 7A may be G Spray Pumps bypassed for
3) Closure of Reactor calibration and Building Steam testing.

Isolation Valve and Its Bypass y 4) Closure of Reactor

@ Blowdown Through i Decay Ileat Removal

@ Valve S) Start lA and IB Ef Diesel Generators

6) Closure of Shutdown J^ Condenser Condensate

- Drain Valve

.

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