Text

Revised Tech Specs Re Limit on RCS Pressure,Operating Status of RCS & Auxiliary Feedwater Pump Periodic Testing
	ML19309G338
Person / Time
Site:	Rancho Seco
Issue date:	04/30/1980
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML19309G337	List: ML19309G336; ML19309G338;
References
	NUDOCS 8005060193
	Download: ML19309G338 (15)
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.

RANCHO SECO UNIT 1 80O 060

/. g 5

TECl!NICAL SPECIFICATIONS O

Surveillance Standards 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feed-water pumps.

Objective To verify that the auxiliary feedwater pump and associated valves are operable.

Soccification 4.8.1 At least every 92 days on a staggered test basis at a time when the avera9e l 64 reactor coolant system temperature is >305 F, the turbine / motor driven and t

motor driven auxiliary feedwater pumps shall be operated on recirculation to the condenser to verify proper operation.

The 92 day test frequency requirement shall be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months af ter the average reactor coolant system temperature is >305 F.

Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at the design flow of 780 gpm.

This flod will be verified using tank level decrease and pump diff erential pressure.

a l

4.8.2 All valves, including those that are locked, sealed, or otherwise secured in position, are to be inspected monthly to verify they are 64 l

In the proper position.

I 4.8.3 Prir r to startup following a refueling shutdown or any cold shutdowa cf longar than 30 days duration, conduct a test to demonstrate that the 64 j-motor-driven AFW pumps can pump water from. the CST to the steam

}.

generator.

Dases 1

1 The quarterly test frequency will be sufficient to verify that the turbine / motor l

driven and motor driven auxiliary feedwater pumps are operable.

Verification of j

correct operation will be made 'both from the control room instrumentation and l

direct visual observation of the pumps.

i P.EFERENC E l

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Proposed Amendment No. 64 Rev. 2 o

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RA'ICHO SECG UNIT 1 TECit!11 CAL SrLCIfICAT10!!S Safety Limits and Limiting Safety System Settings 2.2 SAFF1Y LililTS. REACTOR SYSTEf1 PRESSURE Applicability-Applies to the limit on reactor coolant system pressure.

Obj ec t ive To naintain 't.hc integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.

Specification

?

2.2.1 Tl.c reactor coolant system pressure shall not excced 2750 psig when there are fuel assemblics in the reactor vessel.

2.2.2 T!.a nominal setpoint of the pressurizer code safety valves shall be less than or equal to 2500 psig.

v Itares The reacter co,olant system (1) serves as a barrier to prevent radionuclides in the reactor ccalant from reaching the atmosphere.

In the event of a fuel cladding failure, the reactor coolant system is a barrier against-the relcase of fission products.

Cstablishing a system pressure limit helps to asture the integrity of the reactor ccolant system.

The maximum t ransicht pressure allowable in the reactor coolan'. system pressure (yyssel under the ASME code, Section lit, is 110 percent of design pressure.

The maximum transient pressure allowabic in the reactor coolant system piping, valves, and fittings under ;NSI Section B31.7 is 110' percent of design pressure.

Thus, the safety established.k29sig (110 percent of the 2500 rsig design pressure) has beentrip (2300 psiI limit of 77b The settings for the reactor high and the pressurizer code safety valves (2500 psig) V5yssur have been established to assure that the reactor coolant system pressure safety-limit is not exceeded. The initial hydrostatic ~ test was conducted at 3125 psig (125 per-cent of design pressure) to veri fy the integrity 'of the reactor coolant system.

Additional assurance that the reactor coolant system pressure does not'cxceed the safety limi t. --is provided by sett ing the pressurizer electromatic relief "

valve at 2450_psig.

This setpoint is above normal transients limited by setting the rearter trip at <?300 psig and sufficiently low to. assure limited dependence (4

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on safety. valves operation.

lir rt nrncts -

(l)

FSAR, section:4 (2)

FSAR, paragrarh b.3 8.1 (3)

FSAP., paragraph 4.?.4 h-4 Amendi6ent 64 Rev. 2

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.PANClio SECTO U:3?T 1 TECil:{lCAL SPECIFICAT10:IS

. Safety Limits and Limitlng i

Saf2ty System Settings a

1 i

B.

Pump monitors i

J

}

The pump monitora prevent the. minimum core D:!BR from decreasing below 1.3 j

by tripping tin reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) Joss of one or two reactor coolant pumps during txo-pump operation.

The pump moniturs also restrict the power 1cvel to 55 percent for one reactor coolant pu:n operation in j

cach loap.

A j

C.

Reactor coolant system pressure During a startup accident from low pouer or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuc1 car overpower trip set point.

The trip setting limit shown in figure 2.3-1 for high reactor coolant system pressure (2300 psig) han l 64 been established to maintain the systen pressure below the safety limit I

(2750 psig) for any design transient (1) and minimize the challenges to l

the IZOV and code safetics.

j The low pressure (1900 psig) and variable low pressure (12.96 T

- 5834) out trip set point shown in figure 2.3-1 have been established to 4

maintain the DNS ratio greate.r than or equal to 1.3 for those design 4

accidents that result in a pressure reduction.

(2,,3) j Due f o the calibration and instrumenta-on errors the safety analysis

~

i used t variable low reactor coolant system pressure trip value of l

(12.96 T

- 5884).

out j

D.

Coolant outlet temperature The high reactor coolant outlet temperature trip se'tting limit (619 F)

]

shown in figure 2.3-1 has been established to prevent e::cessive core i

coolant temperatures in the operatint; range. Due to calibra' tion and instrumentation errors, the safety analysis used a trip set point of 620 F.

5 T.

Reactot Luilding pressure i

[

The high Reactor B illding pressure trip setting limit' (4 psig) provides i

positive assurance that a reactor trip will occur in the unlikely i

event of a steam line failure-in the Reactor Building or a loss-of-1-

coolant accident, even in the absence of a low reactor coolant system pressure trip.

4

]

F.

Shutdown bypass l

i In order to provide for control. rod drive tests, zero power physics j_

testing, and startup procedures, there is provision for bypassing j

cert ain segments of t he reactor protection systen. ~- The reactor protection system segments which can be bypasaed' arc shown in l.

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2-T Amendment 64 Rev. 2 s

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Figure 2.3-1 Protective System liaximum Allo.-rable Setpoints, Pressure Vs Temperature l

2600 2100.

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S Acceptabic Operat ion U

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Reactor Outlet Temperature, F s

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l U-l Tl:Cill11 CAL SPEC l Fi CAT I 0lls ~

Amendment 6fi Rev. 2 L

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RA!!Cil0 SECO t!:llT 1 ~

TECll:llCAL SPL,Cir tCATious Limiting Conditions for Operation 3

Lilit T il1G. C0i!DIT 10:15 FOIt OP E P.AT I O!J -

3.1 P.EACTOP. C00LAtlT SYSTEM Applicabi1ity Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations.

3.I.1 OPERAT10!!AL C01 P0tJEi!TS Soccification 3 1.1.1 Reactor Coolant Pu nps Fump combinations permissible for given power levels shall A.

lie as shown in specification table 2.3-1.

' J-B.

The boron concentration in the reactor coolant system shall not be reduced unless tit least one reactor coolant pump or one decay heat removal pump is circulating rcactor coolant.

C.

Operation with two pumps shall be limited to-24 hours in any 30 day period.

3.1.1.2 Steam Generator A.'

One steam generator shall be operabic whenever the reactor coolant average temperature is above 280 F..

3 1.1.3 Pressurizer Safety Valves A.

The reactor shall not remain critical unicss both pressurizer code safety valves are operable.

D.

When the reactor is sul' critical, at least or.e pressurizer code safety valve shall be operable if all rea'etor coolant-system openings are closed, except for hydrostatic tests in-accordance with ASME Boiler and Pressure Vessel Code, Section i

III.

3 1.1.4 Pressurizer Clectromatic Relief Valve.

A.

The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig +_10 psig exceptt ulien required for cold overpressure protection.

42.

j gses A react ar coolant puny -or. decay heat ' removal pump is requi red.to.be in opera-l t lon bdfore the - boran concent rat ion ii.- r' educed by dilut ion wi th makeup wa t er.

lither pump willLprovide-misian which will' prevent sudden positive reactivity

~

' changes caused by dilute' ceolant iriaching t he s cactor..0ne decay heat rcinoval pump will cirtulate the equivalent ol the reactor. coolant system volume-in onc (Ud

RANCII0 SECO UNIT 'l TEC,IlulCAL SpEcirlCATIONS Limiting Conditions for Operation s

1 The decay heat.rcroval systen suction piping is designed for 300 r and 300 psig; thus, the system can remove decay heat when the reactor coolant system is below I

~

(2)

(3) this temperature.

r One pressurizer code safety valve is capable of preventino overpressurization l

uhen the reactor is not crit cal since its relieving capacity is greater than i

that required by the sum of the available heat source which are pun p energy, l

pressurizer heaters, and reactor decay' heat. (4)

Coth pressurizer code safety

)

valves are requi red to t e in service pr ior~ to cri t icali ty to conform to t he l

systen design relief capabilitics.

The code safety valves prevent overpres-sure for rod withdrawal accidents. (5)

The pressurizer code safety valve lif t

}

set point shall be set at 2500 psig + 1 percent allowance for error and each valve shall be capable of relieving 545,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure.

The electromatic relief valve setpo' int was estabished to prevent operation of the volve during transients.

54 j

)

Two pump operation is limited until further ECCS analysis is performed.

I i

1;EFERENCES

)

(1)

FSAR tabics 9.5-2, 4.1-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6

= - '

(2)

FSAR paragraph 9.5.2.2 and 10.2.2

~

(3)

FSAR paragraph 4.2 5 l

(4)

FSAR paragraph 4.3.8.4 and 4.2.4 1

(5)

FSAR paragraph 4.3.6 and 14.1.2.2 3 1

I 1

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l Amendment 4 Rev. 2 g

RANCl!O SECO UNIT 1 TECIINICAL SPECIFICATIONS l

Lihiting Conditions for Operations 3.1.2 PRESSURIZATION, litATt!P, AND C00LDC'JN LIMITATIONS Spyc l fica t t on 3.1.2.1 Innervice Leak and liydrostatic Tests:

Pre ;sure tenperature linits for the first five EFP years of inservice leak and hydrontatic tests are given in Figure 3.1.2-3.

IIcatup and cooldoua

. rates shall Le restricted according to the rates specified in Figure 3.1.2-3.

3.1.2.2

_1!catup Cooldovn:

For the first five EFP years of power opeations, the reactor coolan' t pressure a:,d the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance uith Fir;ure 3.1.2-1 and Figure 3.1.2-2 respectively. IIcatup and cooldova rates shall not exceed the rates stated on the associated figure.

3.1.2.3 The secondary side of the steam. generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 130 F.

3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100 F in any 1-hour pericd.

3.1.2.5

'Ihe spray shall not be used if the temperature differened between the pressurizer and spray fluid is greater than.4100F.

3.1.2.6 Prior to excecriing five c'fective full power years of operation, Figures 3.1.2-1,

-2, and -3 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.S.

The highest predicted adjusted reference temperature of all the beltline ruterials shall be used to deteruine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.7.

3.1.2.7 The updated proposed technical specifications referred to in 3.1.2.6 shall be submitted for NRC review at least 90 days prior to the end of the service peried. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C.

3.1.2.8 E ercenev/raul red Operat ion:

D**

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In the energency/ faulted condition when there is no' forced or natutaal circulation in the react 6c coolant system and there is high pressure injectien and/or m:Aet.p aidition, the Reactor Coolant Systen ter:perature 64 and pre: sure shall be linited in accordance with the: limit line shown en Figure 3.1.2

's.

Unaer t he above eterr,cacy/f aul t ed conditions, Figure 3.1.2-2 vill not neply.

3 ':

Anendment No. 64 Rev.' 2

.-. ~.

RANCl!0 SECO UNIT 1 i

TECHNICAL SPECIFICATIONS Limiting Conditions'for Operation The naximo.m allouable pressure is ta' fen to be the louest pressure of the three calculated pressures.

The pressure limit is adjusted for the pressure dif fcrential 4

between the point of system. pressure measurcrient and the limiting component for all rcactor ccolant punp combinations.

The limit curves were prepared based upon the cost limiting adjusted reference tenperature of all the beltline region materials at the end of the fif th effective full power year.

The actual shift in RT f the beltline region material will be established NDT periodically during operation by removing and evaluating, in accordance with Appendix ii to 10 CfR 50, reactor vessel natcrial irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Uccause the neutron energy spectra at the specimen location und at the vessel inner wall location arc essentially the same, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The lirit curves must be recalculated when-the AP.T determined from thesurvc111ancecapsuleisdiffercatfromthecalculatedANI for the equiva-I l DT lent capsuic radiation exposure.

I f

The enirradiated ir.rpact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which j

sufficieat amounts of material were available.

The adjusted reference temperatures i

are calculated by adding the radiatien-induced LRT and the unirradiated~RTNDT.

The predicted ART arecalculated'usingtherespNb[iveneutron flur.nce and I

yDT copper and phosphorus contents in accordance wi th Reg. Guide 1.99. -

i The assumed RT f the closure head region is 60 F and the outlet nozzle steel

{

forgingsis60hT b.

The limitations imposed on pressurizer heatup and cooldown and spray water.

I temperature differential are provided to assure that the pressuri2cr is operated within the design criteria assumed for the fatigue analysis perforned in accordance with the ASME code requirements.

The limitations during emergency / faulted operation when all reactor coolant flow and all feedvater flow is lost to the OTSG's are established to take into consideration that HPl gives false cold leg temperatures.

This.transi-64 ent is controlled by figure 3.1.2-4 and the vessel beltline temperature is calculated using incore thermocouples and subtracting 150 F for conserva-tism. When the coolant flow or feedwater flow 'is re-established, a four

~

hour transit ion period will be allowed to progress from f.igure 3.1.2-4: to Figure 3.1.2-2.

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3-4 Amendaent No. - 64 fiev. 2 t

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a I;f ACTOR C00l AtlT SYSTEli, EltERGE!!CY/ Faut i ED C0llDITI0ff-COOL DOWil Lilill ATIO:lS, APPLICAULE 10R 5.0 EFFLCTIVE FULL P01/ER YEARS

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2800 2400 LOCl Temp *F Press. (PSIG) 220" 350 21 0' Ut6 4

2000 270*

652 300*

970 o

330*

1460 5

360*

2214 E

368*

2500 O

I 1600 4

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Region Operating E

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- Pressure

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200 250 300 350 400 incore Thermocouple Temperaturc i

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l'i gurt 3.1. 2 4,'

Amendment 64 Rev. 2

i

RA!!Clio SECO UlitT 1 Technical Specifications Limiting Conditions for Operation 3.h.2.2 When two independent 100% capacity auxiliary feeduater flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant 64 shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.4.2.3 When at least one 100% capacity auxiliary [cedwater flow path is not available, the reactor shall be made subcritical within four hours and the facility placed in a shutdown cooling mode which does not rely on steam generators for cooling within next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Ila ses The feedaater systen and the turbine bypass system are normally used for decay heat removal and cooldown above 280 F.

Feedwater ma:

is supplied by operation of a condensate pump and main feedwater pump.

In the event of complete loss of

)

electricti power, feedwater is supplied by a turbine driven auxiliary feedwater pump whi i takes suction from the cordensate storage tank.

Steam relief woeld be through the system's atmospheric relief valves.

f if neither main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the turbine bypass system to the condenser.

{

l in order to heat the reactor coolant system above 280 F the maximum steam removal capability required is 4-1/2 percent of rated power.

This is the maximum decay heat rate at 30 seconds after a reactor trip.

The requirement for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h).

In addition, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity.

The 250,000 gallons of water in the condensate storage tank is the amount needed for cooling water to the steam generators yraperiodinexcessofonedayfollow-ing a complete loss of all unit ac power.

The mi inum relief capacity of seventeen steam system safety nives is 13,329,163 lb/hr. 2) This is sufficient capacity to p ect the steam system under the design overpower condition of 112 percent.

REFEREi1CES

+

(1)

FSAR paragraph 14.1.2.8.4 (2)

FSAR paragraph 10.3.4 s

(3)

FSAR Appendix 3A, Answer to Question 3A.5 J

]

i i

3-24 Amendment No. 64 Rev. 2 l

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TABLE 4.1-1 (Centinued)

INSTRUMENT SURVEILLD ~E REQUIRE'4ENTS Channel Description j

Check

]

Test Calibrate Kerarks l

3

'B.

Reactor coolent precsure/ -

S M

R

{

temperature conparator 9.

Power /f r.balanec/ flow S

M R

~

conparator 10._-Purp/ flux comparator S

M E

11.

High neactor Building D

M R

Precsure Channels 12.

Protection channel NA M

NA coincidence logic 13a. CRD Trip Breaker 1)

EPS Undervoltage trip NA M

NA 2)

Turbine / Gene rator, Lons of Fcedwater Trip NA M

NA 64 ;

133. Turbine Generator Trip

~

Functional' Test NA SY(1)

R (1) T st at next cold shutdown I

I Safety Features System l

14 E=crgency core cooling 3

injection, c=ergency building cooling and building isolation

.a6alog channels a.

Reactor coolant S

M R

pressure channel b.

Reactor Building S

M R

4 psis channel 4

e y

TAOLE 4.1-1 (Cont *nved)

INSTRUMENT SURVE1LLANCE REQUIREMENTS Chan ci Description Check Test Calibrate Remarks i

42.

Reactor Building drain accumulation tank level NA NA R

43 Incore neutron detectors M(1)

NA NA (1) C5cck functioning, including functiening of computer read-out and/or rccorder rcadcut 44.

Process and area radiation cenitoring systems V

M Q

45 Eccrgency plant radiction Instre ents M(1)

NA R

(1)

Battery check

46. Environmental air ronitors M(1) 14A R

(1)

Check functioning

$7.

Strcng rotion accelerometer Q(1)

NA R

(1)

Battery check I

LB.

Auxiliary Feedwa ter Ficw M

NA R

')

'43 Power Operating Relief Valve NA NA R

I a

S = Each Shift M = Monthly P = Prior to cach startup if not donc previcus week

,0 = Caily Q = Quarterly.

R = Once during the refueling interval V = Veckly SY = 9emiannual i

Amendment 64 Rev. 2 5

J

ML19309G338

Text

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