ML20053C083

Proposed Tech Spec Changes Re Min Conditions for Criticality,Instrument Operating Conditions & Accident Monitoring Instrumentation

ML20053C083
Person / Time
Site:	Crane
Issue date:	05/20/1982
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20053C069	List: ML20053C068 ML20053C079 ML20053C083
References
NUDOCS 8206010474
Download: ML20053C083 (15)

Text

!

c.

Raccter coolen: system prassure During 4 startup accident from low power or a slow rod'vithdrawal from high power, the ryrtem high pressure trip se: point is reaebed before the nuclear overpower trip set point.

The trip se:tir.g limi: for high reactor coolant system pressure has been established to maintain the s

system pressure below the saf e:y limit (2750 psis) for any design tran-sient.

(6)Due to calibration and instrumen: errors, the safe:y analy-h'N sis, a,ssumed a 45 psi pressure error in' the high rea: tor coolant system

, e ss pressure, trip se::ing.

i

~

Tne high pressure : rip se:poin: was subsequen:1y lowered fro: 2390 psig to 2300 psig.

Tne lowering of the high pressure : rip se:peint and

{-

raising of the se: point for :ne power Operated Relief Valve (pOAV),

s-fro = 2255 psig to 2450 psig, has the effec: of reducing the challenge rate to the pOAV while main:aining ASMZ Code Saf e:y Valve capability.

The low pressure (1800 psig) and variable low pressure'(11.75 TOUT-5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4).

The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis,the low pressure trip setpoint has been raised to the more conservative 1900 psig.

Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trips.

d.

Coolant outlet temperature Tne hign reae:c coolan: outle: temperature trip se :ing limi (61C F) sncun in Figure 2.3-1 has been es:ablished :c prevent excessive core coolan: temperature in the operating range.

'I Tne calibra:ed range of :ne te=perature enannels of the RTD is 520* :e 620*F.

Tne trip setpoin: of the channel is 619'I.

Under the worst case environment, power supply per:urba: ions, and drif t, ne accuracy o f the - trip s tring is l'F.

Inis accuracy was arrived at by su= ing the Oorst case accuracies of each =odule.

Tnis is a conserva:ive me: hod cf error analysis since the nor=al procedure 'is :c use :ne roo: mean square me: nod.

Therefore, i: is assured :na: a trip will occur a: a value no higner than 620*F even under wors: case conditions.

The safety analysis usec a high tempera:ure : rip se: point of 620*F.

l l

l The calibrated range of the channel is :na: por: ion of the span of

~

indica: ion which has been qualified with regard to drift, linearity, repeatability, etc.

Inis does no: imply :na: :neequifment is i

res:ricted tE operation wi:hin the calibrated range.

Additional tes:-

l ing nas demonstrated that in fact, the te=pera:ure channel is fully operational approximately 10% above :ne calibra:ed range.

l 2-7 8206010LOM

\\

~3.1.3

• dINIEJti CONDITION! top. CRITIC 41.IU Ap licabiliev Applies to reactor coolas: syste= conditio:s required phier to criticality.

L Objective a.

To limi

he magni:ude of any power excursions resulting fro: rea::ivi:y Inser: o oue :o moderater pressure and mocera:er :e=pera:ure coe,:..

1-cients.

.c.

.o assure tha: :he rea :c coo;an: sys:e: v:.., A ne: ge s:A:: :.: he even:

of a rod vi:hdrawal er star:up acciden:.

c.

Tc assure sufficies: pressurizer heater capacity :: naintain natura.1 circulation Octdi:1 ens during a loss of effsite power.

Soetifi:atio J.A.;.;

.ne rea::er :oc., an: :e:perature sna.1 s.e aseve 2.,5o:_ ex:ep: :er per: ons c:...ov power pnys::s :es::: une: :ne requiese: s cf Spe:ifica:ie: 3.1. c sha*1 apply.

.e J.

.a.;

nea::or coe.,a:: :e:pera:ure sn.a;. oe soove....

A0.

.' ***J.;

.,ner. the rea**or coOlat.:

te=pera:ure is 'elov Ohe MinitrJ:

e:-

m o

perature spe:ified in 2.1.2.1 aeove, excep: for per: ions cf lov Power physics testi=; whe: the receireme :s c:. c pe::....: :.ca ::.c

.~. 1. C shal! apply, ne rea::c: shall se sub:ri:ical ey an a=out; ecual :: er grea:er nz: :he calcula:ed rea::ivi:y i:-

ser:io: cue :c depressurica:ics.

3.1.3.4 Pressuri:er 3.1.3.

1

• ne re a::e sna.1 oe main ained superitica". by a: leas: one percen: lak/k us:i; a s:ca oub:le is fe =ed and an inci:a:ed va:er level be: vee: 5: and 255 in:nes is eszahiisned in :ne presser::er.

(a) With the pressurizer level outside the required band, be in at least HOT SHUTDOWN vith the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN vithin an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.1.3.4.2 A c:.n:cu: or 107 xv cf pressurizer heaters, fro: each of two l

i pressurizer hea:er groups shall be 07I742.

Each OPIFM 2 107 kw of pressurizer hea:ers shall be capable of re.ceivf.ng power fro: a 450 vol: IS bus via the established manus.1 transfer senene.

3-6

3.4.3 With tb3 reccter ecolen: sis:cc tempere:ure gracter ths: 250*y, all eighteer (18) main steam safety valves shall oe operable cr, if cny are not operable, the =axi=u= overpower crip se: point' (see Table 2.3-1) shall be rese: as follows:

Maximu= Number of Mav 4 u= Over ower r

Saf ety Valves Disabled on Irip Se:poin:

'Anv Stea=~ Generator

(' ' of ' ?.s t ed ' Powe r) 1 92.4 2

75.4 3

66.3 With more than 3 main steam safety valves inoperable, restore at least fifteen (15) =ain steam safety valves to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

j Eases A reae:or shutdown iciloving power opera: ion re:uires re=ovtl of :cre ce:4y heat.

Nor=Al decay hea: re= ova". is my tne s:ca: genera: Ors vi:h ne s:ca:

c u=; :: :ne condenser whe: REE

e=pera:ure is aoove 250*y and ey :ne decay hea: removal sys:er.belov 250*y.

Core cacay hea: can be cen:itueusly dissipa:ed up o 15 perc e:: of f ull power via :he s:et: cypass :: :he con-denser as f evedva:er in the s:ca: genera:or is :ever:ed :c s:can by nea:

Obserp; ion.

Ner=tily, :ne toa:. :v :e re:urt :eeces:er flov :: :ne stea=

genert:crs is erovihed bv :he = tit feedva:er s.ys:e.

saf e:y valves vil; be able te relieve := a =ospnere One :::al T e =ti s:ca:

s:c a: flov if ne:essary.

If Main S:ea= Safe:y Valves are inopers:le, the power level =us: se reduced, as s a:ed i: Te:hnicti 5pecificatie: 1.s.3, such tha: :he resti=ing saf e:y valves can ac==oca e :he de:ay nea.

1 :ne unlikely event cf co=ple e loss of fif-si:e ele::rical power :: :he s:a:ict, cecay hes: re= oval is by eitner :ne s:ez=-criven ecergen:y f eed-vtte.

or :ve half-s ized = ::r-driven pu=ps.

S:ea: cischarge is c :he r._r, a ::spnere via :ne =2in s:et= saf e:y valves tad

'. led a:=ospneri:

re'.ief valves, and in :ne case f :ne :ur:ine drive pu=p, frc= :ne :urbine

(.,2/

er.saus:.

Le r.==:er-criver pu=ps are receired ini:itlly te re=cve ce:ay nea: vi-h : e eve:.:utlly suf fici ;.

Tne 4 '-"- a=ous: cf va:e-i: :he condensa e s:erage tanks, contained it ~e:n=ical 3pe:ifica ic 3.4.1, val'. allov cocidov: ::

250*y vi:h s:can being dis:narged c the a:=osphere.

After cooling c 230*y, tne decay hea: re= oval sys:e is used :c a::ieve f urther cooling.

An unli=i:ed e=ergen:y f eedva er supply is available fro: -he river via either cf :ne Ivo =o:cr-criven ret::or building e=ergen:y cooling va:er pu=ps for an indefini:e period :e :i=e.

3-26

TABLE 3.5-1 (con't.)

INSTRUMENTS OPERATING CONDITIONS Functional Unit (A)

(B)

(C)

Minimum Operable Minimum Degree of Operator Action if Conditions C.

Engineered Safety Features con't.

Analog Channels Redundancy of Column A and B Cannot be Met (a) 3.

Reactor Building Isolation and Reactor Building Cooling System a.

Reactor Bldg. 4 psig (g) 2 1

llot Shutdown Instrument Channel b.

Manual Pushbutton 2

1 Ilot Shutdown c.

RPS Trip 2

1 Ilot Shutdown d.

Reactor Building 30 psig 2

1 Ilot Shutdown e.

RCS Pressure less than 1600 2

1 Ilot Shutdown Psig f.

Reactor Bldg. Purge line 1

0 (f) a 6

Isolation ( AllV-1 A and AllV-lD) liigh Radiation 4.

Reactor Building Spray System a.

Reactor Building 30 2(d) 1 Hot Shutdown psig Instrument Channel b.

Spray Pump Manual Switches (c) 2 1

Ilot Shutdown 5.

4.16KV ES Bus Undervoltage Relays a.

Degraded Grid Voltage Relays 2

1 (e) b.

Loss of Voltage Relay 2

1 (e) 6.

Emergency Feedwater System (all pumps auto start) n.

Loss of both Feedwater Pumps 2

1 Ilot Shutdown b.

Loss of all RC Pumps 2

1 llot Shutdown

w 33 TABLE 3.5-1 Continued INSTRUMENTS OPERATING CONDITIONS Functional Unit C.

Engineered Safety Features (con't.)

(a)

If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the unit shall then be placed in a cold shutdown condition.

(b) Also initiates Low Pressure Injection i

(c) Spray valves opened by manual pushbutton listed in Item 3 above.

(d)

Two out of three switches in each actuation channel operable.

l (c)

If a relay fails in the untripped state, it shall be placed in a tripped state within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to u,

d, obtain a degree of redundance of 1.

The relay may be removed from the tripped ' state for up to

(*

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for functional testing pursuant to Table 4.1-1.

(f)

Discontinue Reactor Building purging and close AllV-1A, IB, 1C, and ID.

Note:

(a) above.does not j

apply if AllV-1A, IB,1C and ID are closed.

l (g)

Fo r ho t functional testing, prior to Cycle 5 criticality the 4 psig signalis not required for Nuclear Service Closed Cycle Cooling water, Intermediate Cooling and Reactor Coolant Pump Seal Injection (return line only). Two operable channels of a 30 psig Reactor Building isolation signal with a minimum degree

[

of redundancy of 1 are required if the 4 psig signal is not operable for these lines, i

~.

3.5.5 ACCIDENT MONITORING INSTRUMENTATION Applicability Applies to the operability requirements for the instrument identified in Table 3.5-2 during START UP, POWER OPERATION and HOT STANDBY.

Obj ective To assure operability of key instrumentation useful in diagnosing situations which could lead to inadequate core cooling.

Specification 3.5.5.1 The minimum number of channels identified for the instruments in Table 3.5-2, shall be OPERABLE. With the number of instrumentation channels less than the minimum required, restore the inoperable channel (s) to OPERABLE status within seven (7) days (43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> for pressurizer level) or'be in at least HOT SHUTDOWN within the next six (6) hours and in COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Prior to start-up following a COLD SHUTDOWN, the minimum number of channels shown in Table 3.5-2 shall be OPERABLE.

Bases The saturation Margin Monitor provides a quick and reliable means for determination of saturation temperature margins.

Hand calcu-lation of saturation pressure and saturation te=perature margins can be easily and quickly performed as an alternate indication for the Saturation Margin Monitors.

Discharge flow from the two (2) pressurizer code safety valves and the PORV is measured by differential pressure transmitters connected across elbow taps downstream of each valve. A delta-pressure indication from each pressure transmitter is available in the control room to indicate code safety or relief valve line low.

An alarm is also provided in the control room to indicate that discharge from a pressurizer code safety or relief valve is occuring.

In addition, an acoustic monitor is provided to detect flow in the PORV discharge line. An alarm is provided in the control room for the acoustic monitor.

1 i

i j

3-40a l

l i

The Emergency Feedwater System is provided with two channels of flow instrumentation on each of the two discharge lines.

Local flow indication is also available for the emergency feedwater system through the Non-Nuclear Instrumentation (NNI).

Although the pressurizer has multiple level indications, the separation indications are selectable via a switch for display on a single display.

Pressurizer level, however, can also be determined via the patch panel and the computer log.

In addition, a second channel of pressurizer level indication is available independent of the NNI.

Although the instruments identified in Table 3.5-2 are significant in diagnosing situation which could lead to inadequate core cooling, r

loss of any one of the instruments in Table 3.5-2 would not prevent continued, safe, reactor operation. Therefore, operation is justified i

for up to 7 days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for pressurizer level).

Alternate indications are available for Saturation Margin Monitors using hand calculations, the PORV/ Safety Valve position monitors using discharge line thermocouple and Reactor Coolant Drain Tank indications, and for EFW flow using Steam Generator level and EFW pump discharge pressure. Pressurizer level has two channels, one channel from NNI (3 D/P instrument strings through a single indicator) and one channel independent of the NNI.

Operation with the above channels out of service is permitted for up r

to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Alternate indication would be available through the plant computer.

t I

I r

i l

L r

l 3-40b

9 f

0 t

INTENTIONALLY DELETED l

3-95a

e e

a 1

INTENTIONALLY DELETED 1

3-95b

TAT'l 4.1-1 (Continued)

INSTRUMENTS OPERATING CONDITIONS CllANNEL DESCRIPTIONS CHECK TEST CALIBRATE REMARKS 19.

Reactor Building Emergency Cooling'and Isolation System Channels a.

Reactor Building S(l)

M(1)

R (1) When CONTAINMENT INTEGRITY is required 4 psig Channels b.

RCS Pressure 1600 psig S(1)

M(1)

NA (1) When RCS Pressure >1800 psig c.

RPS Trip S(1)

M(1)

NA (1) When CONTAINMENT INTEGRITY is required d.

Reactor Bldg. 30 psig S(l)

M(1)

R (1) When CONTAINMENT INTEGRITY is required e.

Reactor Bldg. Purge W(1)

M(1)

R (1) When CONTAINMENT INTEGRITY is required Line liigh Radiation (All-V-A/D) f.

Line Break Isolation W(1)

M(1)

R (1) When CONTAINMENT INTEGRITY is required Signal (ICCW &

NSCCW)

20. Reactor Building Spray NA Q

NA

7 System Logic Channel

.u

21. Reactor Building Spray System Analog Channels a.

Reactor Building NA M

R 30 psig Channels

22. Pressurizer Temperature S

NA R

Channels

23. Control Rod Absolute Position S(l)

NA R

(1) Check with Relative Position Indicator

24. Control Rod Absolute Position S(l)

NA R

(1) Check with Absolute Position Indicator I

25. Core Flooding Tanks a.

Pressure Channels S(l)

NA R

(1) When Reactor Coolant system pressure is greater than 700 psig b.

Level Channels S(1)

NA R

26. Pressurizer Level Channels S

NA R

27. Makup Tank Level Channels D(1)

NA R

(1) When Makeup and Purification Syetem Is In operation

- -. =

1 TABLE 4.1-1 (Continu d)

CilANNEL DESCRIPTION C11ECK TEST CALIBRATE REMARKS

~49. Saturation Margin Monitor S(1)

M(1)

R (1) When ave is greater than 525 F.

l

50. Emergency Feedwater Flow NA M(1)

R (1) When ave is greater than 250 F.

Instrumentation

51. Emergency Feedwater Initiation a.

Loss of RCP's NA A(1)(2)

R (1) When ave is greater than 250 F.

b.

Loss of both Feedwater NA Q(1)(2)

R (2) Includes logic test only l

Pumps 4

l i'

s' S - Each Shift T/W - Twice per week R - Each Refueling Period D - Daily B/M - Every 2 months NA - Not applicable W - Weekly Q - Quarterly B/W - Every two weeks M

.;onthly P - Prior to each startup if not done previous week

(

r 4.6.3 Pressurizer Heaters 4.6.3.1 Once Each Refueling a.

Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergency power bus and energized. Upon completion of this test, the heaters shall be returned to their normal power bus, b.

Determine that the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.

c.

Determine that following input of the Engineered Safeguards Signal, it shall be verified that the circuit breakers, supplying power to the manually transferred loads for pressurizer heater groups 8 and 9, have been tripped.

4-46a

4.9 EMERGENCY FEEDWATER SYSTEM PERIODIC TESTING Applicability Applies to the periodic testing of the turbine driven and two motor-driven Emergency feedwater pumps, associated actuation signal, and valves.

Obj ective To verify that the Emergency Feedwater (EFW) System is capable of performing its design function.

Specification 4.9.1 TEST 4.9.1.1 Whenever the Reactor Coolant Sy' stem temperature is greater than 250 F, the EFW pumps shall be tested in the recirculation mode in accordance with the requirements.and acceptance criteria of ASME Section XI Article IWP-3210. The test frequency _shall be at least every 31 days of plant operation at Reactor Coolant Temperature above 250 F.

~

4.9.1.2 During testing of the EFW System when the reactor is in STARTUP, HOT STANDBY or POWER OPERATION, if one steam generator flow path is made inoperable, a dedicated qualified individual who is in communication with the control room shall be continuously stationed at the EFW local manual valves (See Table 4.9-1).

On instruction from the Control Room Operator, the individual shall realign the valves from the test mode to their operational alignment.

4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be verified to be in the status specified in Table 4.9-1, when required to be operable.

4.9.1.4 on a quarterly basis, verify that the manual control (HIC-849/850) valve station functions properly.

4.9.1.5 on a quarterly basis, EFV-30A dn 30B shall be checked for proper operation by cycling each valve over. its full stroke.

4.9.1.6 Prior to start-up, following a refueling shutdown or a cold shutdown greater than 30 days, conduct a test to demonstrate that the motor driven EFW pumps can pump water from the condensate tanks to the Steam Generators.

1

• For the purpose of this requirement, an OPERABLE flow path shall mean an unobstructed path from the water source to the pump and from the pump to a Steam Generator 4-52

4.9.2 ACCEPTANCE CRITERIA These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9.1.1.

Bases I

1 The 31 day testing frequency will be sufficient to verify that the turbine l

driven and two motor-driven EFW pu=ps are operable and that the associated valves are in the correct alignment.

ASME Section XI Article IWP-3210 specifies requirements and acceptance standards for the testing of nuclear safety related pumps.

Compliance with the normal acceptance criteria I

assures that the EFW pumps are operating as expected. The test frequency of 31 days (nominal) has been demonstrated by the B&W Emergency Feedwater Reliability Study to assure an appropriate level of reliability.

If testing indicates that the flow and/or pump head for a particular pump is not within the normal acceptance standard an evaluation of the pump performance shall be completed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or the pump declared inoperable.

In the case of the EFW System flow, the flow shall be considered acceptable if under the worst case single pump failure, a. minimum of 500 gpm can be delivered 3

when steam generator pressure is 1050 psig and one steam generator is isolated.

A flow of 500 gpm, at 1050 psig head, ensures that sufficient flow can be delivered to either Steam Generator. The surveillance requirements ensure that the overall EFW System functional capability is maintained.

1 i

4 4-52a

3 Table 4.9-1 l

Status of EFW Valves i

4 Valve No.

Status

]

CO-V-10A Open I

CO-V-10B Open EF-V-1A Open I

EF-V-1B Open i

i EF-V-2A Open i

EF-V-2B Open MSV-2A Open i

e MSV-2B Open j

EF-V4 Locked Closed i

EF-V5 Locked Closed EF-V6*

Locked Open EF-VIOA*

Locked Open EF-V10B*

Locked Open EF-V-16A*

Locked Open j

EF-V-16B*

Locked Open i

l EF-V-20A*

Locked Open t

l EF-V-20B*

Locked Open i

~

EF-V-22 Locked open CO-V-176 Locked Open i

i

• Manual valve to which Specification 4 9 1 2 applies 4

a E

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s y--

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y,

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