ML19330C074

Requalification Training Program, Week 1,1980.Provides Required Operator Training in Heat Transfer,Fluid Flow, thermo-dynamics & Plant Transients

ML19330C074
Person / Time
Site:	Sequoyah
Issue date:	08/05/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19330C069	List: ML19330C068 ML19330C074
References
PROC-800805, NUDOCS 8008070515
Download: ML19330C074 (150)

Text

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O e O

SEQUOYAH NUCLEAR PLANT REQUALIFICATION TRAINING PROGRAM  :

4 WEEK 1 1980 h

dn c - so d B h l a ' -

~i 6v  ;#1 POWER OPERATIONS .

TRAINING CENTER 8008070

o e SEQUOYAH NUCLEAR PLANT REQUALIFICATION PROGRAM DAILY OUTLINE Section I. Day 1 A. Classroom

. 1. Accident Analysis

a. Condition I

b. Condition II

c. Condition III

d. Condition IV

2. Discuss proper E0I or AOI applicable to analyzed accident.

B. Simulator Control Room

1. Start unit up from hot standby to full power utilizing proper GOI's and AOI's.

2. Insert malfunctions causing the following analy=ed accidents insure licensees utilize the proper procedures,

a. Small loss of reactor coolant

b. Large loss of coolant (LOCA)

c. Steam line break inside containment

d. Steam generator tube leak

l .

i Section I. Day 2 A. Classroom

1. Operation transient analysis

a. Review response times associated with transients

b. Review major plant parameter responses to transients -

pressurizer pressure, level and reactor coolant system temperature B. Simulator Control

1. Initiate plant shutdown from 100% power

2. Utilize appropriate malfuncitons to require licensee response to plant transient, such as MFPT trip, loss of load, inadvertent opening of turbine control valves 9

t

-3 Section i

I. Day 3 '

j A. Classroom

1. Plant operating characteristics and heat transfer, emphasizing degraded core conditions

a. Discuss basic heat transfer equations and use of the j steam tables

b. Combine the information in the steam tables with basic plant parameters to calculate heat transfer, thermal power, etc. during steady state and transient conditions

c. Calculate heat transfer problems with and without change of phase under st.eady state conditions,

d. Calculate heat transfer problems under transient conditions

3. Simulator Control Room

l. Require each licensee to bring reactor critical from the hot standby condition utilizing control rod malfunctions and nuclear instrumentation malfunctions 1

2. Discuss and demonstrate all means available in control room to determine primary system condition, margin to saturation, and inventory of H 2O k

9 e

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• e l

Section I. Day 4 A. Classroom

1. Introduction and review of E0I Immediate actions and diagnostics

a. Spurious actuation of safety injection

b. Loss of reactor coolant

c. Loss of secondary coolant

d. Steam generator tube rupture

2. Review E0I 1, 2, and 3 with particular emphasis on subsequent actions relating to a degraded core conditions

3. Simulator Control Room

1. Reactor power at 8% bring unit on line and increase power

2. Utilize malfunctions requiring use of E0I-0 1, 2, and 3 G

G l

[

o .

Section I. Day 5 A. Classroom

1. Review Sequoyah Nuclear Plant License for fuel loading and low power testing

2. Review E0I 14 - anticipated transient without scram AOI 6 - excessive primary plant leakage AOI 24 - steam generator tube leak

3. Simulator Control Room

1. Perform a portion of the Sequoyah Nuclear Plant natural circulation tests.

, 2. Utilize applicable malfunctions remaining use of E0I 14 ATk'S, AOI 6, excessive primary plant leakage, and AOI 24 steam generator tube leak t

6 4

l I - . ._ ,

o e Day 1 I. CIASSROCM A. Subjects to be covered

1. Purpose of the accident analysis

2. Definitions and concepts used in accident analysis 3 Accidents

a. Condition II moderate frequent faults (1) Rod withdrawal accident (2) Red assembly sisaidan~ ant (dropped rod)

(3) Uncontrolled boron dilutica (4) Partial loss of forced coolant flow (two PCP's trip)

(5) Startup of an inactive reactor coolant icop (6) loss of external electrical load (and/or turbine trip)

(7) Loss of all offsite pcwer (station blackout)

(8) Feedwater system nalitnctions (9) Excessive load increase incident (10) Accidental depressurication of RCS (lr) Accidental depressurication of m4n steam syste= (S.V.

lifts) ,

t (12) Spurious operation of S.I. at pcwer

b. Condition III - infrequent faults (1) Loss of reacter coolant (n-al' break)

(2) Minor seccedary system pipe break (3) Inadvertent loading of fuel asse=bly into an i= proper i position.

l l

(4) Cenplete loss of forced coolant ficw I l

(5) Waste gas decay tank rupture (6) Single RCC vithdrawal at full power

e .

Dry 1

2. A. 3. c. Condition IV - L1=iting Faults (1) Major Loss of Reactor Coolant Accidents .

(2) Major Secondary Systes Pipe Rupture (3) Stena Generator Tube Rupture (4) Single Reactor Coolant Pu=p Locked Rotor (5) Fuel Handling Accident (6) Rupture of Control Rod Housing

4. Goals

a. Understand the Purpose of the Accident Analysis

b. Understand the Ccucepts Used in the Accident Analysis

c. The Student Should Have a Clear Understanding of INER Concept and the Parameters That Effect the INER

d. Be Able To Discuss All Accidents (1) Describe Accident (2) Discuss Plant Trips and Other Protection To L1=it the

, Consequences of the Accident (3) Se Able To Drav Rough Curves of Selected Parameters

3. Presentation

l. General Information -

a. AUS Classification of Plant Conditions (1) Condition I: Normal Operation and Operational Transients (2) Condition II: Faults of Moderate Frequency (3) Ccudition III: Infrequent Taults (4) Cendition IV: W ting Faults

b. Basis of Safety Analysis (1) Most Frequent Occurance - Little or No Radiolo(.,:al Risk to the Public (10CFP20 is the Limit) Conditica I and II (2) Extreme Accidents -

l

o e Day 1 I. 3. 1. b. (2) (a) N 11 Be Least Likely To Occur (b) Radiological Risk Is By 10CFR100, Condition III And IV aa. Whole Body Dose At Site Boundary For First 2 Hours:

25 Rem bb. Thyroid Dose At Site Boundary Fcr First 2 Ecurs:

3% Rm (3) where Applicable, Reactor Trip System And ISF Functioning Is Assured To The Extent A11cwed 3y Considerations Such As Single Failure Criterion C. General Descriptien of Esch Condition

1. Condition I - Normal Operation And Operational Transients

a. Regular Operaticus of Plant (1) Power Operation (2) Refual ",g (3) Maintenance -

(4) Maneuvering Flant

b. Conditica I - Conditions Are Considered From The Point Of Affecting The Consequences Of Fault Conditions

c. A Setpoint Document Is Generated For Each Site And Is not Analy::ed In The Accident Analysis

d. Typical List Of Condition I Events (1) Steady State And Shutdown Operations (a) Power Operations (15 To 100%)

(b) Startup (Critical, O to 15%)

(c) Hot Shutdown (d) Cold Shutdown (e) Refueling (2) Operation With Per=issible Deviations Linited By Technical Specifications (a) Components Out Of Service (b) Cladding Leakage

e e Dry 1 I. C. 1. d. (2) (c) RCS Activity (d) Steam Generater Tube Leaks (3) Operational Transients (a) Plant Heatup And Cooldown

-(b) Step Load Cher.ges (Up To + 10%)

(c) Ramp Load Changes (Up To 5%/'4"4-~)

(d) Load Rejection Up To And Including Design Load Rejection Transient

2. Conditics II - Faults of Moderate Frequency

a. These Faults At Worst Result In A Reactor Shutdown With The Plant Capable Of Returning To Operations

b. These Faults Do Not Propagate To Cause A More Serious Fault

c. Do Not Cause 4

(1) Fuel Rod Fa4'me a

(2) RCSOverpressurization(2750 psia)

d. Faults In Condition II (1) Uncontrolled Rod Cluster Ccutrol Asse=bly Bank Withdrawal From A Suberitical Condition (2) Uncontrolled Rod Cluster Control Assembly 3ank Withdrawal At Power (3) Rod Cluster Control Assenb2y Misalignment (4) Uncentrolled Boron Dilution (5) Partial Loss of Forced Reactor Coolant 71cv (6) Startup Of An Inactive Reactor Coolant Loop l (7) Loss Of Deternal Electrical Load And/or Turbine Trip (8) Loss of Nornal Feedwater l (9) Loss Of Offsite Power To The Station Auxiliaries (Station Blackout)

(10) Excessive Heat Renoval Due To Feedwater Systen !Ls1 functions i

_1._. .

Dny 1

. I. C. 2. d.(11) Excessive Load Increase (12) Accidental Depressurizatica Of The Reactor Coolant System (13) Accident Depressuri stien Of The Main Stes: System (14) Inadvertent Operation Of ECCS (Or ESS) Power Operation 3 Condition III - Infrequent raults

a. Occur Under Infrequent Conditions

b. Fuel Rod Danage May Occur, But only A Snall Fraction At Worst Case

c. Release of Radioactivity Will Not Eastrict Use Of Land Beyond Exclusica Radius

d. Will Not Cause A Condition U Nor Ccusequential Loss of Function Of RCS Cr Centainnent

e. Faults In Condition III (1) Loss of Coolant From Snall Breaks (2) Minor Secendary System Pipe Breaks (3) Inadvertent Leading Of A Fuel Assenbly Into An I= proper Position (4) Couplete Loss Of Forced Reactor Coolant Flow (5) Waste Gas Decay Tank Rupture (6) Single RCCA Withdrawal At Full Tower

4. Condition N

a. Faults Which Are Not Expected To Occur

b. Most Drastic Accidents Which .T.tst 3e Oesigned Against

c. Release Of Radioactivity Li=ited By 10CFRICO

d. A Single Accident Will Not Cause A Consequential Loss Of Required Functions of Systens Need Fcr ESF

e. Faults In Condition 17 '

! (1) Major Ruptures of RCS l

(2) Major Secondary System Pipe Rupture (3) Steas Generator Tube Rupture

I' Dcy 1 I. C. 4. e. (4) Single Reactor Coolant Punp Locked Roter (5) Fuel Handling Accident (6) Rupture Of A Control Rod Rousing D. Definitions And Concepts

1. Definition Of Reacter Trip - Insertion Of All Full length Reds Except The Most Reactive. It Is Considered Stuck At The Top.

2. Steady State Instrument Errors

a. Chosen For Each Accident To Make The Accident As Bad As Possible

b. Errors Used (1) ?cwer - + 2% (Calorinetric Error)

(2) T - + k 7 (Dead 3and And Instrument Error)

(3) P - + 30 psi (3ank of Operatica And Instre.nent Error)

c. Errors Normally Chosen To 144nd-4 ce Initial DIER 3 Review The Definition Of UGR g ,

Heat Flux ITecessary To Cause DNB Local Heat Flux 9

9

9.

• Dcy 2 I. CIASSRCCM

A. Operational transient analysis

1. Basic plant parameters

2. Heat transfer in a dynamic plant

a. Q = WCp a T (heat input)

b. Q = UA 6 T (S/G heat output) 3 calculation of RCS temperature acceleration with 100% reactor power nnd0%turbinepower.

k. Pressurizer percent level change per F change in RCS temperature.

5 Pressurizer pressure change per percent pressurizer level change.

6. Cceparisen of constant Ts program and constant Tave program and sliding Tave program.

7 ?cwer mismatch effects

3. Design function and capabd2.ity of rod control system and steam du=p system C. Operational transients

1. 50% ramp load increase frca 50% to 100% load at 5%/ min.

rate. (BOL) All systems in automatic.

2. 10% step load increase from 90% load to 100% load. (30L)

All systems in automatic.

3 10% step load decrease from 100% to 90% load. (30L) All systems in automatic.

4. 50% load rejection (BOL) 100% to 50% load. (BOL) with control rods in automatic and in manual.

5 50% step load decrease (30L) frca 100% to 50% load with auto =atic rod control but with no steam dump.

6. 75% step load rejection (30L) 100% to 25% with all systems in automatic.

.e s Day 2 3ASIC FLAUT PARAMEIRS (Tave, Tstream, RCS press, steam press, level in' p::r, te p in pzr)(RCS flow)

Heat transfer in a dynamic plant.

EY Apply for=ala Q = WCp A T where:

Q = heat input to RCS ( )

1 lb w = aCS riov (g) = 135 x co,

' coo gal

=in (4 % = 360, gal su=1 10 1 1D (3.6 x p,. M )(8.33 p lb)(60hrW )(.75) = 1 35 x ur Cp = specific heat - F It takes 1 Stu to raise 1 lb of water 1 F at 70 0F but at 578 F, it takes a little = ore than 1 Etu to raise one Ib 1 F (1.3 Stu , oy )

lb AT = temp rise across vessel = 65 F (Rx iAmt) 1 10 3*"

Q = (1 35 x )(l' h - /)(65 M = 11.h x

, 4

^

S/GheatcutputQ=UA oT=

2 U = heat transfer coefficient - ft - F A = heat transfer area = ft AT = Tave - Tsteam = F = 55 (578 - 523 )

Since U and A are ccustant, the only way to increase heat output from S/Gistoincrease o T between Tave and Tsteam.

You can ei'.ner hold Tave constant and let Tsteam drop (which would drop steam I.,ressure a lot og Increase Tave encugh to keep Tsteam constant (which would cause fuel clad temp. to get close to DNS)

CE L .

s . -

Day 2 Increase Tave some and let Tsteam drop sc=e to achieve the d T.

'At SNP , we use this third gthod whgeh in:reases Tave by 31 F and let Tsteam drop 24 F (from 547 to 523 F).

Calculate how fast RCS temperature will change with a 100% power mismatch between turbine and reacter. Reactor heat output =

1

• and S/G heat cutput is 0*

115x 3 First--Must calculate how many 3TU's required to raise RCS tesperature 17 (heat capacity of the H O). 2 Total Cp = (1bs of water in RCS)(specific heat -

F) 1 63 **

cp = (500,000 M (1 3 "7 - F) = .65 x (Takes 650,0C0 Etu's to raise the RCS 1 degree F) o Second--Find out how =ary , the RCS will heat up if reacter power is 10C1,and steam flow goes to zero.

(11 5 x 10 * # )=32x 10

• g )(3.6 x 103 see see 32x 10 M Rate of rise = Cp total 10

.65 x 9

= 5% (n t counting stored heat in =etal) see L

I

• s Dy2 RULES OF T?2G

1. 1CC% =isnatch = 5F change h Tave

,,g

2. As Tave changes, the density of the water will change.

As Tave increases, water will expand and pzr level will rise at what rate?

It has been established that the pzr program level is to take care of go expansion due to the slidi g Tave program.

.!!O.

60% - 24.7% C: 35% change for a Tave change of 578 F - 547 F = 31 ?

354 , 1.13% change .

31 F 1 F Tave change y lI

~

1F 3 As pzr level changes, pressure will change. A good ballpark figure is

~ 10 SM chera l#g level change g* gals

% level I

• (1700 (

JE ) = 12750 gal for entire range so

~ 125 ga1

1. p level

• due to taps not covering entire range.

1 1

l l

o s Dzy 2 Constant Ts Program

? t

, .Q = UA (Tave ~ Ts)

Ts is constant Ps is constant Advantage - Simplified secondary plant design.

Disadvantages - a. Large volume char.ges with power; i.e., large pzr required.

b. Large change in Tave with power; i.e., large reactivity changes required. More control rods.

c. Very high TH - possible INE design problems.

Constant Tave Program 7 ** 4 Q = UA.(Tave - Ts)

Advantages - a. Less rod metica during transients.

b. Less reactivity change during transient.

c. Less volume change, smaller pzr.

Dicadvantages - Increased complexity of secondary plant design.

Increasing Tave With Pcwer Program 1 +-* f  ?

Q = UA (Tave - Ts)

Reason for Choice:

a. Stean pressure doesn't drop as much as con:Sant Tave progrs=.

b. Pzr not as big as ccustant Ts program,

c. Reactivitf changes not as large as censtant Tr program.

d. Do not as much INB design problems as constant Ts program.

PowerMigmatch 6pm Change 6 Tave (20% pwp) assume

. 30L e/ m = -5 eem4 <ed = 1 eem 4-a

1. =(54")(6F) = 30 pem j d (1,)(6 r) = 6 ee=

36 pcm = 6 steps EoL /s = -35 cm /n = 1 #9 "

(+35fC") (6 ?) = 190 p m (1 3F )(6 F) = 6 pcm 196 pcm 106 ces i 6 cem = 32 steps step -

o .

PLANT OPERATING CHARACTERISTICS AND HEAT TRANSFER INTRODUCTION The post TMI discussions have placed a lot of emphasis on upgrading the operator's knowledge of the plant. Much of the emphasis had been directed to the understanding and use of the saturation curve and the steam tables. Saturation curves have consequently been prominently displayed in many control rooms, including ours at Sequoyah. It has been recommended that operator's be more familiar with heat transfer fundamentals and be able to make calculations using basic heat transfer formulas and the steam tables. We will perform several heat transfer calculations.

1

• s -

l Dry 3 OlWECTIVES At the .end of this r esentation, 'the student should be able to:

1. Discuss basic ' cat transfer equations and use of the steam tables. '

2. Combine the iriformation in the steam tables with basic plant parameters to calculate heat. transfer, thermal power, etc. during steady state and transient conditions.

3. Calculate heat transfer problems with n_oo change of phase under steady state conditions.

4. Calculate heat transfer problems with a change of phase under steady state conditions. ,

9

5. Calculate heat transfer problems under transient conditions.

A. Basic Plant Parameters

1. Basic concept is to use plant parameters such as flow and

• temperature and the " Steam Tables" to calculate and explain the values of the plant parameters. '

2. Enthalpy

a. Enthalpy is used in the concepts discussed above.

b. Enthalpy could be called inherant energy. It is equal to the internal energy of the fluid plus the flou energy of the fluid.

c. The enthalpy of water has been experimentally determined and the results have been tabul ted in such publications as the steam tables.

d.' Enthalpy (h) has units of Btu per lb.

Page 2

5 L Day 3 c.

Exampics of enthalpy and use of the table:.

(Students look up^ answers) '

BTU (1) Enthalpy of saturated watnr (hf ) at 70'F = 38.052 LB .

(2) , Enthalpy of saturated steam (hf) at 70*F = 1092.1

. BTU /LB. . ,

= -

(3) The difference in enthalpy is the energy required to convert one pound of water into steam. This is called h ygt . the Latent Heat of Vapo ~rization. ife'll

, , , use this phenomenon in the study of the steam generator.

~

(4) Enthalpy also increases when the water is heated. .

If we raise the temperature of water from 424*F to 525*F (both saturated temperaturcs) we have increased the heat content frcm 401.3 BTU /LB to 518.1 BTU /;B.

,' lt required 116.8 BUT/LB. to increase the temperature.

(5) The steam. tables have a list of values for stca;a and water at saturation. It does not list values for '

compressed liquid. However, the BTU content of compressed liquid does not vary appreciably from liquid at saturated temperature, so we'll use thesc steam tables. -

3.

Heat Generation In a Reactor Plant '

TP-1 a. Q) is the heat gained by RCS in core.

i

b. Q2 is the heat lost by.RCS in S.G.

4 c

Q3 is the heat transferred from the primary side of the SG. to the secondary side of the S.G, -

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Page 3

- h ;, _ _ _ __ . .

Dy 3

. d. 04 is the heat gained by water in 5.G, which is

• then converted to steam and supplied to the l lain Steam System.

c. QS is the heat demanded by the various steam loads.

f. At. steady state conditions:

Q) = Q2"03"04"05

4. Heat Transfer Equations _

a. The basic equations used are:

Q=mCp at TP-2 Q = UA at -

Q=m ah Where: .

i Q = heat added or subtracted - Stu Th - Rx outlet temperature *F .

TP-3 Tc - Rx inlet temperature *F

' Tstm - Steam temperature, steam generator *F .

F

' fu = Feed. tater temperature to steam generator *F Pstm - Pressure in S.G. - PSIA m pri - Primary Coolant mass - lb '

m sec Secondary mass - lb -

h stm - Enthalpy of steam leaving the S.G. - Btu /lb h

fu Enthalpy of feedwater entering the S.G. - Btu /lb Cp Specific heat of water - Btu /lb/*F ,

U - Overall heat transfer coefficient - Btu /f t *F - Ib .

A - !! cat transfer area '- Ft2 .

. Page 4

_n -- - --

. D:y 3 B. Ileat Addition to a Material With 1:o Change in Phase

~

1. The fundamental equation which describes this effect.

Q = m Cp AT *

(write on chalkboard, explain terms again)

Where:  ;

Q = total heat added on subtracted from body (8tu) m = mass of body (Ib) ~

Cp = heat capacity of a material (Btu /lb *F) '

aT='(Tfinal - Tinitial) = temperature change of body (*F)

' In most power plant applications heat is added to flowing

' fluids and in these cases it is convenient to consider heat and mass flow rates and rewrite the equation: -

Q = m Cp AT '

Where: '

= heat adoition or subtraction rate (Btu /hr) m = mass flow rate (lb/hr) '

$1

2. Another fundamental equation we can use. ~

Q =, m oh (itrite on chalkboard explain terms again)

Where: - - -

, . . ~

Ah=hinlet - houtlet = enthalpy in Btu /lb. ,

3.

Ex, ample : In a certain PWP. the coolant enters the Steam Generator at

_576 *F and leaves the S.G. at F34 _* F.

The reactor coolant flou {

rate through the S.G. is 114.5*[/hr. At what power is th'e reactor operating.

. , Page 5

m .

~

D::y 3' Solution di -

Q2= n Cp aT .

In = 114.5 x 100 lb/hr Cp = 1.28 (550*F) Btu /lb - F AT = 576-534 Th - Tc *F .

6 lba 1.28 -

Q22 114.5 x 10 hr x lb - F x (57G-524) F

. t

• I4W 6

Q2= 6155 x 10 Btu /hr x 6 3.413 x 10 Stu-he .

.

Q = 1803 nt;

, 2 Solution !2 * . .

.Q m ah ' - -

2= 6

, m =,114.5 x 10 lb/hr -

Ah = h

• 576*F = H7 583.7 Btu /lb inlet h

=h outlet =T 534 = M 529.3 Btu /lb c f ,

= 54.4 Btu /lb

- . 1b 6 Btu - .

, Q2 = 114.5 x 10 ITi x 54.4 T5'-

Q2 = 6228.8 Stu/hr Using the conversion IGl = 3.413 x 106 Btu /hr, we can calculate

• reactor power (thermal) .

. o 6

6tu to! '

Q:h = 6228.80 x 10 nr x 6 3.413 x 10 Btu - hr

• 1825 megawatts C.

Heat Addition to a liatorial With'a Change of Phase

1.

To heat a flaid which is initially below the boiling point we can use the formula: ..r -

Q . = r.i Cp a t +xa .Wap ' '

Page G

, _a 6

-.. - ~ ~" ~

unerc: D2y 3

'4 steam -

X = % quality (liquid + stea ) note: 0 = pure liquid Avap = Btu /lb (heat of vaporization) -

AT = T sat

-T inlet -

o_r, we can use the formula: -

Q = m, ah .

. Mhere: ,

^

t

~5:=steamflow(lb/hr) -

A h = hg of steam - hf of feedwater

2. Problem:

Calculate the transfer of heat from the 5.G. and determine the power being delivered to M.S. system. 6

• The steam flow is7.8x1011Whr .

.the feedwater temperature is 425 *F, and S.G. pressure is705.18 pstA.

3. Solution:

Q4= m ah - -

6 lb Btu Q4 = 7.8 x 10 lir x 799.27 W .

Q

<4 = 6234.305 Btu /hr - steam flow g . .

m = 7.8 x 10 lbm/hr hf = 425 F = 402.5 Btu /lb hg = 504 F = (705.78 PSIA) = 1201.7 atu/lb ah = 799.27 Btu /lb -

Btu

4. Q4 = 6234.306 Btu /hr x 3.413 x 106g .g -hr .

Q4 = 1826.64 H.I

. thermal' D. Primary &. Secondary Flow Cal.culations Probicm: Explain why the primary flo : of 114 x 106 lb/hr at 100% is 6

equal to; 7.8 x 10 lb/hr steam flcu.

Page 7

6 6 Dsy 3 This is explained by the difference of the clergy content of the heat transfer mediu:as; RCS fluid and I:cin steam. .

Solution:

Primary Q2= m Ah .

Secondary Q4= m oh g lb 54.4 Btu

. .Q 2 =;114.5 x 10 h7 x ( lb) g4 7,3 x 100 lb/hr '

x 799.2 Stu W

- 6

. Q2 ",6228.8 x 10 Btu /hr - -

~

Q4 = G233.S x 106

.ea e.

, . . Stu/hr O

a L

l 0

7 Dsy 3

• ~

F. Effect of Tu'rbine Power - Reactor Power Mismatch

.  ; -

.s .

Example: Reactor power is 1% higher than turbine power which we can

. simulate by closing the turbine control valves by 1%. Calcula te the heat up rate.

. 5 , .. .

Solution:

J From previcus calculation,100% power = 6233.76 x 106 ,

9 '

or s 6.2 x 10 Btu /hr. (q .

9

,, therefore 1%,= 6.2 x 10 x .01 = 6.2 x 107 Btu /hr. ,

Example: -

To calculate the Btu's in the RCS at a T ave of 555 F, we use RCS vol = 7480 f t Spec. Volume of RCS at 555 F =f V of .02191 f 3t /lb (Note: Specific Volume is the reciprocal of density and is found in the stcac tables) hfof RCS at 555=555.9 Btu /lb Q=Vol x vfxht 3 '

'

• 3

_.02191 ft 555.9 Btu ,

Btu in RCS = 7480 ft x lb

• lb 6 ' '

= 189.78 x 10 ' Btu originally in core .

~

y.Blu Btu added = 6.2 x 10 hr x 0.1 hr (6 min.)

= 6.2 x 10,6 Btu .

Original RCS Btu's = 189.78 x 10 6 '. .

6.2 x 10 6 Added RCS Btu's = .

• • Total RCS Btu's '195.98 .

(after 6 minutes)

Page 9

l Dry 3

~

, To calculate the enthalpy ot the system, uc need Iltu/lb.

195.98 x 106 Btu lb

,34), 39 {x 10 G = 574 Stu/lb (7480 ft3 x

= .341, 396 x 1c'

.02191 ft 3 From the stcom tables: -

n 5,74 Btu /lb = 569 F ,

New RCS Temperature = 569 F

• Original RCS Temperat ire = 555 14*F change in 6 minutes .

Therefore a mismatch of 1:: will cause temperature to rise 14*F/6 minutes or 23*F/ minute.

t The calculated heat up rate is. higher than expected because we, .

1.-

.  !!cgiccted reactivity change due to increasing temperature du'e to MTC.

2.

Neglected temperature changes in metal which would more slowly absorb the heat input to the RCS.

. e i .

. .

e.

., .: , .r O

Page 10 v.

HEAT TRANSFER EQUATIONS . .

Q 1=s p ri cp (T h -Tc ) - REACTOR HEAT OUTPUT (MINUS

. PUMP HEAT) ,

02*b p ri cp (Th -Tc) - HEAT TRANSFER TO STEAM GENER ATOR (S. G. ) .

03 = UA (T ave -Tstm) - HEAT TRANSFER FROM PRIMARY SIDE OF S.G. TO SECONDARY SIDE OF S.G.

04=i sec (h stm-h fw) - HEAT TRANSFERRED FROM STEAM GENERATOR - HEAT GAINED BY FEEDWATER PASSING THROUGH S.G.

E

, 05= ,

- HEAT DEMANDED BY STEAM LOADS 2

HEAT TRANS'FER - FLUID FLOW TERMINOLOGY -

d ~ = HEAT ADDED (OR SUBTRACTED)- Btulhr .

T = REACTOR OUTLET TEMPERATURE - F i 1 h

~

T = REACTOR INLET TEMPERATURE UF c

T = STEA.M TEMPERATURE, STEAM GENERATOR ,UF sim T = FEEDWATER TEMPERATURE TO STEAM GENERATOR UF f

Pstm = PRESSURE, STEAM GENERATOR - PSIA mi pr = PRIMARY COOLANT FLOW RATE -Ibm /hr .

.. m sec = SECONDARD FLOW RATE, STEAM OR FEED -Ibm /hr .

hstm = ENTHALPY OF STEAM LEAVING S.G. - B!ullb ,

hg = ENTHALPY OF FEEDWATER ENTERING S.G. - Btullb C = SPECIFIC HEAT OF WALER - Stullb - F (~1.0 AT 70 F) (~'1.28 AT 550 F) .

P . g

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s s SequoJah Nuclear Plant .

DISTRIBUTION l 1C Plant Master File E}ERGENCY OPERATING INSTRUCTICN Superintendent 1U Assistant Superintendent (Ope r. )  !

E0I-0 IU Assistant Superintendent (Maint.)  !

Administrative Supervisor IMMEDIATE ACTIONS AND DIACNOSTICS Maintenance Supervisor (M)

Assistant Maintenance Supervisor (M) l Maintenance Supervisor (E)

Units 1 & 2 ' Assistant Maintenance Supervisor (E)

Maintenance Supervisor (I)

Results Supervisor 1C Operations Supervisor Quality Assurance supervisor a j , , __

..- - .. ;..

__ ' . Health. Physics. . - . . -- -

Public Safety Services Supv.

i Chief Storekeeper l

f Preop Test Program Coordinator Outage Director I

Chemical Engineer (Results)

Radiochem Laboratory Instrument Shop 1 l Reactor Engineer (Results)

Instrument Engineer (Maint. (I))

Mechanical Engineer (Results)

Staff Industrial Engineer (Pl. Sys.)

_JIL_ Training Center Coordinator PSO - Chickamauga Engrg Unit - SNP Prepared By: J. R. Walker Public Safety Services - SNP Revised By: N/A

,jfL_ Shift Engineer's Office

_jf,_ Unit Control Room QA&A Rep. 'SNP -

Submitted By: , Health Physics Laboratory Supervisor

- lll_ Asst Dir NUC PR (Opm-), 727 EB-C ,

IU Nuclear Document Control Unit, 606 EB--

PORC Review

_jjt_ Superintendent, WBNP Date Superintendent, BENP Superintendent, BENP in NE3, W9C174C-K -

Approved By: Supv., NPHPS ROB, MS Superintendent NRC-IE:II Power Security Officer, 620 CST 2-C Date Approved: Nuclear Materials Coordinatee-1410 CUB:

Manager. OP-QA&A Staff 1c Resident NRC Inspector - SNP I

• tc NSRS, 249A HBB-K .

.

ic Technical Support Center Rev. No. Date Revised pages Rev. No. Date Revised Pages 0 M1 - ~

The~last page of this instruction is Number 12 -

1 4

e .

. SqiP E0I Units 1 and 2 Pase 1 of 1 Rev. 0 PURPOSE I This instructica presents the autcmatic actions, the 1::m:ediate operator actions and the diagnostic sequence which is to be followed in the identification of the fol-lovin4:

A. Spurious Actuation of Safety Injection

3. Loss of Reactor Coolant C. Loss of Secondary Coolant D. Steam Generator Tube Rupture

, The reactor automatic protection equipment is designed to safely shut down the reactor in the event of any of the above emergencies. The safety injection system is designed to provide emergency core cooling and boration to naintain the uafe reactor shutdown condition. These plant safeguards syst'e=s operate with offsita electrical power or from onsite emergency diesel-electric power should offsite power not be available.

In the subsequent doc'.:ments in this series (E0I-1, 2 and 3), instructiens for recovery frcm the event are presented for each particular accident. .

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9

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m * * * ~ = * = .. . ..n . ,- . a.

6 e

. . SQ:TP i E0I Units 1 and 2

) Page 1 of 9 i Rev. 0 3, I. SYMP" CMS

TOTE: The process .-_-iables referred to in this instruction are typically

=enitored by = ore than ene instrumentation channel. The redundant j channels should be checked for censistency while perfor=ing the steps I of this instruction.

A. The folleving sy=ptc=s are typical of those which =ay arise in a plant which is undergoing a loss of reactor coolant, 1 css of secondary coolant or stes= generator tube rupture (one or more sy=ptc=s =ay appear in any order):

1. kw Pressurizer Pressure I. 2. kw Pressurizer Water Level 3 High Pressuri:er Water Level

h. High Centain=ent Pressure 5 High Contain=ent Radiation

6. High Air Ejector Radiation T. High Stea= Generator Slowdown Radiatica

. 8. Steam nev/Feedvater Flev Mis =atch 9 Letdown Isolation / Pressurizer Heater Cutout

10. Lev Lov Reactor Coolant System Average Ccolant Te=perature

, 11. High Containment Recirculation Su=p Water Level

12. kw Stes=line Pressure (one er all Stes=line.5) .

, 13 Lov Steam Generater Water evel ih. Increasing Stes: Generator Water Level 15 Rapidly Changing Reactor Coolant Syste Average Coolant Te=perature f

. 16. Increased Charging Flev 17 High Steam Flov (one or all Steam lines) 18.

High Containment Hu=idity i

L

i A .

• s l

. . SQNP l E0I Units 1 and 2 l Page 2 of 9 l Rev. 0 I. SDtPTCM9 (continued) 19 High Containment Temperature I

20. Iov Feedvater Pu=p Discharge Pressure NOTE: The pressurizer water level indication shculd always be used in  ;

. conjunction with other reactor coolant system indications to evaluate system conditions and to initiate =anual operator actions.

II. I!S!EDIATE ACTICNS p A. Conditions warranting reactor trip or safety injection say be characterized by a number of anomalous situations or unusual instru=ent indications.

i

1. If the plant is in a condition for which a reactor trip is warranted ,

and an automatic reactor trip has not yet occurred, =anually trip the reactor. Continue =enitoring plant conditions as shcvn in Figure 1.

f CAUTICN_: If there is not a rapid drop in nuclear power and the control rods are not inserted, then this is an ATWS event. Atte.pt to trip the reactor by other means, (see E0I-lh as needed).

2. If the plant is in a condition for which safety injection is warranted and an autesatic safety injection has not yet occurred, manually initiate safety injection.

t

3. Verify the folleving actions and system status: .

1. Reactor trip .and turbine trip have occurred.

a. All rods inserted.

b. Turbine steam stop valves closed

2. Bus voltages indicate that the busses are energized sad all intended '

+

loads are being powered.

!- a. Generator breakers open ~(ti=e delay of 30 seconds if no electrical fsult)

b. Station service transferred 3 Feedvater Isolation has occurred

a. F.W. Isolatica valves closed .

b. Main F.W. reg. valves closed

c. Main F.W. reg. bypass valves closed MF ed w

e . .

. SrJTP 20I Units 1 and 2 Page 3 of 9 Rev. O II. I!GtEDIA"E AC*ICNS (continued)

4. Contain=ent Isolatica Phase A, centain=ent ventilation isolation and safety injection have occurred and valves and ds=pers are in proper position:

NOTE: Test all status =enitor lights prior to relying en position

, indication.

a. Safety injection and Phase A:

Panel 6C - Dark Panel 6D - Dark Panel 6E - Light (except for cutlined)

Panel 6F - Light (except for cutlined)

Panel 6G - Dark

?anel 65 - Dark 5 Auxiliary Feedvater Pu=ps have started and the Auxiliar/ Feedvater Syste= valves are in their proper E=ergency Align =ent and are open or closed as appropriate for present S/G level.

6. Safety Injection Pu=ps have started:

s. Centrifugal charging pu=ps (hi head SI) i

b. Safety injection pu=ps (lo head SI) i -

f -

i c. RER pu=ps (lo head SI)

AND Verify flev through the 3IT, as pressure falls, verify SI pumps and RER pumps deliver flev.

7 ERCW and Cc=penent Cooling Water Pu=ps have started. ~

8. Energency Diesel Generators have started.

9 Energency Gas Treat =ent and Auxiliary Suilding gas treat =ent syste=s have started .

C. If any of the above autc=atic actions have not occurred and are required, they should be =anually in2.tiated.

=

• l I

4 .

SQNP E0I Units 1 and 2 Page 4 of 9 Rev. O

II. IMMEDIATE CFZTOR ACTICNS (Continued)

C. Continued I

1 Verify the folleving:

1. Safety Injection flow from at least ene train is being delivered to the reactor coolant system when the Reactor Coolant System pressure is telev the high head safety injection pump shutoff head. If not, atte=pt to operate equipment =anually or locally.

,. 2. Auxiliary Feedvater flow frem at least one train is being delivered to l the steam generators. If not, attempt to operate equipment =anually I or locally.

• f NOTE: Only after steam generator water level is established above the top of the U-Tubes, should the Auxiliary Feedvater System Flow be regulated to maintain reequired level.

3 Verify that heat is being removed from the reactor plant via the stes=

generators by noting the folleving:

a. Autc=atic steam du=p to the condenser is occurring; '

b. Reactor coolant average te=perature is decreasing towards progran=ed no-load temperature.

I.

NOTE: If condenser steam du=p has been blocked due to a control nalfunction or loss of the " Condenser Available" conditien, decay heat renoval vill be effected by autc=atic actuation of the steam generator power-operated relief valves, cr, if these prove ineffective, the steam generator code safety valves. In this event, stesn. pressure vill be =aintained at the set pressure of the controlling valve (s) and reactor ecolant average temperature vill stabilize at approximately the saturation te=perature for the steam pressure being -

=aintained.

D. If running, step contain=ent building " auxiliary" floor and equip =ent drain i

pumps (controls en M-9) and verify contain=ent f1cer and equipment drain ptmps stopped (controls on M-15).

ene g .. _ _.. ._ . _ . _ - . . . _ _ .

SQl?

E0I Units 1 and 2 Page 5 of 9 Rev. 0 II. I:CIDIATE AC"'ICNS (Continued)

E. Whenever the contai=ent hi-hi pressure setpoint is reached, verify the "ollowing:

1. Main Stes: isolation valves have closed

2. Centai=ent spray is initiated 3 Contai=ent isolatica phase 3 is initiated NOTE: Test all status =cnitor lights prior to relying on position indication.

t.

i a. Panel 6C - Dark Panel 6D - Dark Panel 6E - Light Panel 6F - Light Panel 6G - Dark Panel 6H - Dark I' nain stea= isolation valves have not closed, =annai'y close frc= the control board.

If contai=ent spray or phase 3 isolation have not cecurred, =anually initiate.

i . III. ACCICETT DIAGNCSTICS (Refer to Figure 2) i A. Evaluate reactor coolant pressure to dete. ine if it is icv er decreasing in an uncontrolled manner. If it is lov or decreasing, verify that :

1. All pressurizer spray line valves are closed and

2. All pressurizer relief valves are closed ,

If not , manually close the valves frc= the control board.

1 I If the RCS pressure is above the icv pressure reacter trip setpoint and is .3 table er increasing, so to step G.

a- -

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e .

SQIT?

E0I Units 1 and 2 Page 6 of 9 Rev. O III. ACCIDE'TT DIAGNCSTICS (continued)

3. Stop ALL Reactor Coolant Pu=ps after the high head safety injection pu=p operation has been verified and when the vide range reacter coolant pressure decreases to 1300 psig.

I j ~ "~'IC:T : If._ component ecoling vater to the reactor ecolant pu=ps is i isolated on a contain=ent phase 3 isolation,all reacter coola;1t l'

pu=ps should be stopped within 5 =inutes because of less of i =otor bearing cooling.

i b CAUTICIT : If the reactor coolant pu=ps are stopped, the seal injection i

flev should be =aintained.

, NOTE: The conditions given above for stopping reactor coolant pu=ps l should be continuously =cnitored throughout this instructien.

C. Il the condenser air ejector radiation or steam generator b1cvdown radia-tion =cnitor exhibit abnor= ally high readings, AND contai=ent pressure,

'.' centai=ent radiation and contai=ent recirculation su=n. level exhibit nor=al readings, THEN go to E0I-3, " Steam Generater Tube Rupture."

D. _IJ,the steamline pressure is abnor 1'y lever in ene stea= generator than in the other stes= generators, TEE'T go to E0I-2, " Mss of Secondary Coolant."

t h' E. _IF contain=ent pressure, _OR contai=ent radiation _CR contai=ent recircula-

tien su=p levels exhibit either abncrm'77 high readings or increasing L readings, THE'T to ta E0I-1, " b ss of Reacter Coolant".

i NCTE: For very s=all breaks inside the contai=ent building, the centain-

=ent pressure increase vill be very small and possibly not reccgni-

! :able. For very s"" breaks the centai=ent recirculation sump water level vill increase very sicvly and early in the transient

=ay not indicate a level increase. ..

F. E the centai=ent pressure, contai=ent radiation A'ID cen ='a~a*

recirculation su=p water level centinue to exhibit stable readings in the

! normal pre-event range, THE'T go to E0I-2, " Loss of Secondary Coolant".

. G. In the event of a spuricus safety injection signal, the sequence of reacter trip, turbine trip and safeguards actuation vill eccur.

, The operator =ust assu=e that the safety injecticn signal is non-spurious unless the following are exhibitied: .

1. Normal readings for contai=ent temperature, pressure, radiation and recirculation su=p level AND -

e e .

. . SQNP E0I Units 1 and 2 Page T of 9 Rev. O III. ACCIDENT DIAGNCSTICS (continued)

2. Nor=al readings for auxiliary building radiation and ventilation

=cnitoring MTD 3 Nor=al readings for steam generator blevdevn and condenser air ejector radiation I I? all of the sy=pto=s 1 through 3 above are =et and when the folleving h

' through T are exhibited:

1 l' h. Reactor coolant pressure is greater than 2000 psig and increasing mfd 5 Pressurizer water level is greater than progrs==ed no lead water level AND

6. Ihe reactor coolant indicated subecoling is greater than ho F.

t 7 '4ater level in at least one S/G is stable and increasing as verified by auxiliary feedvater flow to that S/G. Auxiliary feedvater flow to the unaffected S/G's should not be reduced belov h00 GPM until indicated level is returned to within the narrev range level instru-

=ent. *

NOTE: Pressurizer water level should trend with reactor coclant syste= te=perature. If the pressurizer water level is icv i

' ~

enough to prohibit pressurizer heater operatica, re-establish vater level by operating the charging system. Energize the heaters.

i THZ'T

8. Reset safety injection and stop safety injection pu=ps not needed for nor=al charging and RCP seal in.f ection flev.

CAUTICN: Autc=atic reiniitiatica of safety injection vil1 not occur '

since the reactor trip breakers are not reset.

8 CAUTICN: Subsequent to this step, should loss of offsite pcVer cccur,

=anual action (e.g., =anual safety injection initiation) i vill be required to load the safeguards equip =ent onto the diesel povered emergency busses.

9 Place all safety injection pu.7s not needed to provide normal changing flow in standby = ode and maintain operable safety injectica flevpaths.

m

.s

a .

. . SQ3?

E0I Units 1 and 2 Page 8 of 9 Rev. O III. ACCIDENT DIAGNOSTICS (centinued)

10. Isolate safety injection flow to RCS Cold Legs via 3oren Injection Tank and establish nor=al charging flow.

a. Close seal injection water L v centrol valve FCV-62-89

b. Open the charging pu=p suction valves from the VCT FCV-62-132 and 133

c. Close the charging pump suction valves frc= the RWST FC7-62-135 and 136.

I

d. Open the centrifugal charging pu=ps miniflev isolation valves FC7 98 and 99 -

e. Close the 3IT inlet isolation valves FC7-62-39 and h0 and cutlet isolatien valves FCV-62-25 and 26.

f. Open the charging line iscaltica valves FCV-62-90 and 91.

g. Open seal water heat exc % ges inlet isolatien valves FCV-62-61 and 63.

h. Gradually open the seal injection vater flow control valve FC7 , 89 Adjust the seal vater flow to 8 GPM per RCP.

. 11. Reestablish nor=al =akeup and letdown (if letdevn is unaffected) to =ain-tain pressurizer water level in the normal operating range and to =aintain

+

reactor coolant pressure at values reached when safety injection is

' terminated. Ensure that water addition during this process does not ,

result in dilution of the reacter coolant system boren concentration.

a. Open the letdown line isolation valve FCV-62-TT

b. Open FCV-62-81 25% on hand and i==ediately open FC7-62-73 (h5 GPM ,

orifice isolation valve). I==ediately adjust FC7-62-81 for desired letdown pressure setting, then place pressure centrol en auto.

I (Nor=al pressure is 320 PSIG ! nor=al letdown temp. )

I NOTE: After charging and letdevn have been established, additional

,' letdevn may be increased as condition per=it.

12. Reestablish operatica of the pressurizer heaters. Tcen reactor ecolant pressure can be centro 11ed by pressurizer heaters alene, return makeup sad letdown to pressurizer water level control only.

l

• n

- SQNP

• i E0I Units 1 and 2 Fage 9 of 9 Rev. O _. _

III. Continued NOTE: IJ, after securt .t ety injectionand attempting to transfer to nor=al pressurizer pr :,.n l' and level c0ntrol, reactor coolant pressure drops below the lov jres. irizer pressure setpoint for safety injection actuation OR if pressurizer vater level drops belov 10% of span, g

(

the reactor coolant sub-cooling drops below h0 F,

• Ei SAFETY INJECTICN MUST BE MANUALLY REINITIATED l

' Rediagnose plant conditions and proceed to the appropriate e=ergency instruction.

1 CAUTION: Stopping and stac ing of the high head safety injection pumps and lov head safety injection pumps can cause pu=p motor overheating or reduced cotor life. Hence if the' pumps are restarted once after ter=ination, an additional 15 F of sub-cooling should be added to the requires sub-cooling prior to the second termination of the high head pu=ps.

NOTE: E after security safety injection and transferring the plant to nor-=1 pressurizer pressure and level control, the reactor coolant pressure does not drop below the low pressurizer pressure setpoint for safety injection actuation AND the pressurizer water level re=ains above 10%

span, AND the reactor coolant indicated subcooling is greater than k0 F, TEI go to AOI-19 for recovery from inadvertent safety injection.

r I

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1 o o SQNP E0I-O - Units 1 & 2

Figure 1 -

Page 1 of 1 C Rev. O t

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' HAS AN AUTCMATIC F. VALUATE NEED FCR l REACTOR )- MANUAL REACIOR l IF NO - - RETURN TO TRIP GCCURRED? TRIP j j NORMAL I '

a l; CPERATION A

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VERIri MANUALLY INITIATE REACTOR ( REACTOR TREP TRIP V

'(

HAS AUTOMATIC EVALUATE NEED SI INITIATION NO FOR MANUAL ,,

IF NO OCCU1'IED7 ) SI INITIATION l

V YES IF M

. CO TO v s, AOI'S

, VERITl MANUALLY SAFECUARDS -

SEQUENCE INITIATE SI l

! CHECK PAMS V

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- SAr!"Y . :::JECn0N Return to - Figure 1 CG 70 =p a tot z.r- . 5 tCO:'" ARY C00*.A:.7 . o . r:c.-: : 6 6 7 __... _.. __ _ _._ ._ _ _. l '} . -Ssquoyah Nuclsar Plant 1C Plant Master File EMIRGENCY OPERATING INSTRUCTION Superintendent IU Assistant Superintendent (Oper.) IU Assistant Superintendent (Maint.) E0I-1 Administrative Supervisor Maintenance Supervisor (M) Assistant Maintenance Supervisor (M) LOSS OF REACTOR COOLANT Maintenance Supervisor (E) ~ Assistant Maintenance Supervisor (Ej l Maintenance Supervisor (I) ' Units 1 f 2 Results Supervisor 1C Operations Supervisor g Quality Assurance Supervisor

• Health Physics Public Safety Services Supv.

Chief Storekeeper , Preop Test Program Coordinator Outage Director Chemical Engineer Radiochem Laboratory Instrument Shop Reactor Engineer Instru.:, ant Engineer Mechanical Engineer Staff Industrial Engineer 1c Training Center Coordinator PSO - Chickamauga Engrg Unit - SNP Prepared By: C.T. Benton Public Safety Services - SNP 1r Shift Engineer's Office . Revised By: J.Jt. Walker . 1e Unit Cuatrol Room j) QA&A Rep. - UNP Submitted By.,- e >>i V / Health Physics Laboratory Sig5ervisor in ' Chief, Nuclear Generation Branch o/ in P Prod Central Office File PORC Review: 414 /Ed in Superintendent, WBNP Cated - Superintendent, BFNP ~ Superintendent, BENP in EN DES - NEB NEG Approved By: .h. 'M O - Supv., NPHPS ROB, MS Q uperintendent-- in NRC-II:II g Power Security Officer ~, - 620 CST 2-C Nuclear Materials Coordinator Date Approved: 7/[d/MO Manager, OP-QA&A Staff / / in P Prod Plant Ehg. Branch ic NSRS, 309 GB-K - -- 1c Technical Support Center ic n. 4 a.h nur man m.- svo Rev. No. Date Revised Pages Rev. No. Date Revised Paees 8 2/9/80 All * ~ _ 9 /f 4 - - O - The last page of this instruction is Number 29 '___2 ~' ~ ..;- ~ ~ ~~

• a

. , ,o, SQNP EOI-l - Units 1 & 2 l Page 1 of 1 Rev. 8 l PURPOSE To provide a discussion of the event, symptoms, automatic actions, inunediate and subsequent operator actions and recovery from: A. Loss of reactor coolant B. Station blackout, with SIS reset, following a LOCA while in recirculation mode. ?)" 8 f l i . i .  ; i i 5 a-k 4 o - - - ___. , _ _ . _ _ _ __. m.._ __ ___ . _ . . _ __ 4 O 4$ m neme f g ,,g' mee-m w r - - - , ,,- g y .,-,--4 -

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SQNP EDI-1A - Unit 1 & 2 Page 1 of 18 Rev. 8 LOSS OF REACTOR C00LAhi I. SYMPTOMS NOTE: The process variables referred to in this instruction are typically monitored by more than one instrumentation channel. The redundant channels should be checked for consistency while performing the steps of this instruction. A. The behavior of the parameters, which the operator can observe in the control room, will be similar in many respects following the initiation of either a loss of reactor coolant, a steam-line break, or a steam gen-erator tube leak. (Refer to Figure 1 and 2) The olperator will have to determine the accident type as soon as possible to carry out the responses necessary to correct or reduce the severity of the accident. The probable magnitude of the parameter changes will be determined by the rate of the loss of reactor coolant. The following symptoms are typical of those , which may arise due to a large leak or loss of reactor coolant:

1. High containment airborne radiation.

2. Rising containment pressure with normal steam generator pressure.

3. Water accumulation in the containment sump with normal steam generator pressure.

B. The following symptoms may also be observed during a loss of coolant:

1. Decrease in pressurizer pressure or level.

2. Increase in charging flow.

3. Decrease in volume control tank level.

4. Increase in containment temperature and humidity.

. C. Possible Alarms

1. " Pressurizer pressure low backup heaters on." (2210 psig)

2. " Ice condenser doors open." (1 lb/ft )

3. " Pressurizer level low, heaters off, letdown secured." (17% level)

4. " Pressurizer level high-low'.'" (-5% from program)

5. " Pressurizer low pressure reactor trip." Reactor first-out annunciation.

(1970 psig)

6. " Pressurizer safety injection pressure low reactor trip." (1870 psig)

A e @49$66 D- O* 6 88* 6 w w SQNP E0I-1A - Unit 1 & 2 Page 2 of 18 Rev. 8 I. SYMPTOMS (Cont.) C. 7. " Containment pressure high safety injection actuate" (1.54 psig)

8. " Containment pressure high safety injection reactor trip" Reactor first-out annunciation.

9. "RA-90-106 containmer.t building lower compartment air monitor high radiation" 10.

D" "RA-90-ll2 containment building upper compartment air monitior high radiation"

11. " Containment vacuum relief isolation valve closed" (1.5 psig)

. 12. " Containment high-high pressure steam line isolation" (2.81 psig) 1

13. " Containment high-high pressure spray actuation" (2.81 psig)

14. " Containment isolation Phase B" (2.81 psig)

15. " Lower compartment moisture high"

16. " Lover compartment temperature high" (120'F incr.)

17. " Upper compartment moisture high"

18. " Containment sump level high" (El. 676.2 in sump)

NOTE: The PZR water level indication should always be used in con-junction with specified reactor coolant system indications ' to evaluate system conditions t.sd to initiate manual operator actions. II. AUTOMATIC ACTIONS A. Reactor trip and resultant turbine trip. B. hafety injection actuation. C. Diesel generator start. D. Containment isolation Phase A actuation and containment vent isolation. E. Containment spray and Phase B containment isolation actuation. F. Feedwater isolation and automatic control of auxiliary feedwater valves on accident signal. 3 , , _ . _ ,_ . . . . . . - . - - - -- -- a- --*** g . m g g *e ww.a 6mhd6he D64& 6 "O O-OO "OO ' ? 3 SQNP E0I-1A - Units 1 & 2 Page 3 of 18 Rev. 9 III IMMEDIATE OPERATOR ACTION CAUTION: If there is not a rapid drop in nuclear power and the control rods are not inserted, then this in an ATWS event, see E0I-14 for instructions. A. Verify turbine trip - reactor trip.

1. All rods inserted.

2. Turbine steam stop valves closed.

3. Auxiliary feedwater pumps running and flow established to S/G's.

4. Generator breakers open and unit station service transferred to start busses.

5. Steam generator feedwater isolation, main regulator, and bypass valves closed.

B. Verify safety injection and containment isolation actuated.

CAUTION: If RCS pressure is below 1870 psig and SI has not auto-matically initiated, manually initiate the SI. / NOTE: Test all status monitor lights prior to relying on position M# indication.

1. Safety injection and phase A monitor lights: Panel 6C - Dark; Panel 6D - Dark; Panel 6E (except for outlined) - Light; Panel 6F (except for outlined) - Light; Panel 6G - Dark; Panel 6H - Dark.

Verify flow through EIT tank. As RCS pressure falls, verify proper flow from SIS pumps and RHR pumps. . 2. Phase B containment isolation and containment spray actuated monitor lights: Panel 6C - Dark; Panel 6D - Dark; panel 6E - Light; Panel 6F - Light; Panel 6G - Dark, Panel 6H - Dark. a: If OB isolation occurs, verify main steam isolation valves closed.

3. Emergency diesel generator start.

C. Stop all reactor coolant pumps after the high head safety injection pump operations have been verified and when the wide range reactor coolant pressure decreases to 1300 psig. , SE Within 5 minutes after component cooling water has been isolated to the reactor coolant pump motor bearings due to OE containment isolation. Maintain seal injection flow to RCP's. If pumps are tripped, refer to Appendix C for natural circulation guidelines. e l o I

• 1 SQNP E0I-1A - Units 1 & 2 Page 4 of 18 Rev. 8 III. IMMEDIATE OPERATOR ACTION (Cont.)

D. Declare gas condition I. IV. SUBSEQUENT OPERATOR ACTION CAUTION: Reference Leg Heatup: High energy line breaks inside containment can result in heatup of level measurement ref. leg's on the SG's and PZR. This will result in erroneous level indication, there-fore it will be necessary to control the levels as follows during post accident S/G's - level greater than 40% and less than 71% on narrow range PZR - level greater than 50% and less than 70%gy - Also use the following backup variables to verify the existence of water level in one or more S/G's:

1. Auxiliary feedwater flow

. 2. Steamline pressure

3. Wide range T and T hot eold 4

Reference Leg Boiling: DO NOT rely upon S/G water level indi-cations in depressurized S/G's during post accident conditions. NOTE: Refer te step Z of this instruction for removing noncondensable gases from RCS. NOTE: Request performance of SI-268 to verify position of P-4 contacts (failure of P-4 contacts may prevent resetting SI) CAUTION: Monitor RWST level closely. If RWST level decreases rapidly (5-10%/ min.) such that the RWST low level alarm- appears imminent, be prepared to move to step "P" of this instruction. CAUTION: Monitor the primary system (RCS) throughout this instruction for indication of INADEQUATE CORE COOLING. Inadequate core cooling exists when: . (a) 5 or more core exit thermocouples exhibit readings at or , above 1200*F. gR (b) If the high range readings from the core exit thermocouples are not available, a condition of inadequate core cooling exists when: The hot leg RTD's are pegged high or five or more incore thermocouples are off-scale above 700*F and SI flow is not being delivered to the RCS and AFW is not being de-livered to the intact steam generators. g g m, s.h m - G =W h @ $ 6 *M= 6 MMM h6 **Me .Q t SQNP E0I-1A - Units I and 2 Page 5 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued) If conditions in (a) or (b) are determined to exist see Appendix D of this procedure for instructions to restore core cooling. NOTE: The process variables referred to in this instruction are typically monitored by more than one instrumentation channel. The redundant channels should be checked for consistency while performing the steps of this instruction. NOTE: Monitor RCS pressure for duration of LOCA for the purpose of de-termining if the RCP's should be stopped, see immediate actions for correct pressure at which to trip RCP's. .,3.. NOTE: Use computer program and Appendix A to determine margin of subcooling being maintained. NOTE: The pressurizer water level indication should always be used in con-junction with other specified reactor coolant system indications to evaluate system condition and to initiate manual operator actions. NOTE: The following instruments are designated as post accident monitors. . Actions should be based on these instrumente.

1. Containment Pressure PdI-30-44, 45

. 2. Refueling Water Storage Tank Level II-63-50, 51

3. Steam Generator Water Level (narrow range) LI-3-39, 52, 94, 107.

4. Steam Generator Water Level (wide range) LR-43, 98

5. Steam Line Pressure PI-1-2A, 2B, 9A, 9B, 20A, 20B, 27A, 27B

6. Pressurizer Water Level LI-68-320, 335A

7. Reactor coolant system temperature TR-68-1, 60

8. Reactor coolant system pressure PI-68-68A, PR-68-66 CAUTION: As water level in the RWST decreases under action of the

, safeguards pumps, check that the RHR recirculation sump water level indicates an increase in water level. If a sump water le,el increase is not evident, then a re-evaluation of the symptoms must be conducted.

• o SQNP E0I-1A - Units 1 and 2 Page 6 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued)

A. Close all PZR power operated relief valves. B. Verify component cooling water pumps running. C. Verify ERCW pumps running. D. Transfer NR-45 to one source range and one intermediate range channel. E. 1. Verify that control room ventilation has isolated from SIS. (Refer to SOI-30.lB). i

2. Close U-2 equipment hatch temporary doors (734iT.), and check equipment transfer hatch cover in place. (734' el. to lower el.).

i F. Verify that all ECCS pumps are operating with less than red line amps. G. Verify that emergency gas treatment system and ABGT system operating. H. Verify that the upper head injection system accumulator isolation valves close on low level. I. If running, stop containment building " auxiliary" floor and equip-ment drain pumps on M-9 and verify containment floor and equipment drain pumps stopped on M-15. NOTE: The conditions given below for termination of safety injection should be continuously monitored throughout this instruction: CAUTION: If SI termination criteria cannot be met and RCS pressure is dropping, proceed to step J. Ensure that containment isolation is maintained, i.e., not re-set until such time as manual action is required on necessary process streams. Safety irijection can be terminated _IF,:

1. RCS, pressure is greater than 2000 psig and increasing

. AND

2. PZR water level is greater than 50*,' of span.

AND,,

3. The RCS indicated subcooling is greater than 40*F AND 9 6h &M D6*8'

• SQNP E0I-1A - Units 1 and 2 Page 7 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued)

4. Water level in at least one S/G is stable and increasing as verified by auxiliary feedwater flow to that S/G. Aux-iliary feedwater flow to the unaffected S/G's should not be reduced below 400 g7m until indicated level is returned to within the narrow range level instrument. l THEN

5. Reset safety f njection and stop safety injection pumps not needed for not._1 charging and RCP seal injection flow.

CAUTION: Auto reinitiation of safety injection will not occur since the Rx trip breakers are not reset. Also subsequent to reset of therSI signal, should loss of offsite power occur, manual action (e.g. , manual SI initiation) will be required to load the safeguards equipment onto the D/G's.

6. Place all SI pumps not needed to provide normal charging flow in standby mode and maintain operable SI flowpaths.

7. Isolate safety injection flow to RCS cold legs via BIT and establish normal charging flow.

CAUTION: If RCS pressure drops below the low PZR pressure setpoint for safety injection or PZR water level i drops below 20% of span following termination of safety injection flow or the reactor coolant sub-cooling drops below the value for SI termination, . manually reinitiate safety injection to establish reactor coolant pressure and PZR water level. Re-evaluate the event. CAUTION: Stopping and starting of the high head safety in-jection pumps and the low head safety injection pumps can cause pump motor overheating or reduced motor life. Hence, if pumps are restarted once after termination, an additional 15 F of sub-cooling should be added to the required sub-cooling prior to the second termination of the high head pumps.

8. Place all available containment upper and lower compartment coolers and CRDM coolers in service.

9. Reestablish normal makeup and letdown to maintain PZR water level in the normal operating range and to maintain RCS pressure at values reached when safety injection is term-inated. Ensure that water addition during the process does not result in dilution of the RCS boron concentration.

@ub 4e 6 h6M MZ**'. * *O* * ***

• SQNP j E0I-1A - Units 1 & 2 '

Page 8 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued)

10. Re-establish' operation of the PZR heaters. When RCS pressure can be controlled by PZR heaters alone, return makeup and letdown to PZR water level control only. (See Appendix B for availability of PZR heaters),

11. Monitor either the average temperature indication of core exit thermocouples or all wide range RCS T indications to H

verify that the RCS temperature is at least 50'F less than saturation temperature at RCS indicated pressure. If 50*F indicated subcooling is not present, then attempt to establish 50*F indicated subcooling by steam dump from the S/G's to the condensers or the atmosphere. See Appendix C for guide-lines relative to natural circulation if RCP's are off. CAUTION: If steam dump is necessary, reduce the S/G pressure 200 psi below the lowest steam safety valve setpoint (864 psig) and maintain a reactor coolant cooldown rate of no more then 50*F/hr, consistent with plant makeup capability. I If 50' indicated subcooling cannot be established or maintained, then manually reinitiate safety injection. Re-evaluate the event. I

12. Perform a controlled cooldown to cold shutdown conditions using normal cooldown procedures if required to affect repair.

' Maintain subcooled conditions (at least 50'F indicated sub-cooling) in the RCS. If subcooled conditions cannot be main-tained, go to step 17.

13. Notify shift engineer that gas condition I may exist.

14. Leave both trains of EGTS in service' till recovery portion

of procedure is carried out.

After E 30 minutes and as time permits, place one train of , ABGTS in standby mode. i

15. After one hour and if offsite power is available, remove D/G's from service.

1 I l' i ( i . -.n . .. .. 4 l SQNP E0I-1A - Units 1 and 2 Page 9 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued)

16. Verify boron concentration sufficient to maintain shutdown I margin. .

Recovery: The unit will remain in cold shutdown with the 1 RHR system in the cooldown mode for an indefinite period of time. l' Instructions for recovery shall be prepared when l the condition of the containment and the accident j evaluation is determined.

17. If the conditions for terminating safety.; injection are not met, maintain necessary safety injection jumps operating.

l- If RCS pressure is above the low head safety injection pump shutoff head, manually reset safety injection so that safe-guards equipment can be controlled by manual action. Stop the low head SI pumps and place in the standby mode and con-tinue with step J. CAUTION: Whenever the RCS pressure dereases below the j low head SI shutoff head, the low head SI pumps j must be manually restarted to deliver fluid to the RCS. NOTE: The conditions previously given'for stopping RCP's should be continuously monitored throughout this instruction. J. In the case of a break characterized by reactor coolant pressure ) quickly decreasing below steam generator pressure, so to step K. In the case of a break characterized by a slowly decreasing reactor l coolant pressure or stabilized reactor coolant system pressure above I the lowest steam side safety valve setpoint (1064 psig), the following additional manual actions should be taken to aid the cooldown and de-pressurization of the reactor coolant system. 1 NOTE: Regulate or verify auxiliary feedwater flow to the steam generators to restore and/or maintain required narrow range water level (PAMS). l If narrow range level increases in an unexplained manner in one l steam generator, go to E0I-3, Steam Generator Tube Leak.

1. If the main condenser is available, After steam line pressure is l equalized, open at least one main steam line isolation valves or

. bypass valves and transfer the steam dump control to steam header pressure mode and dump steam to the condenser to lower the reactor . coolant temperature (PAMS) and consequently the reactor coolant l pressure. e e . ed"* ',e . ese O , SQNP EOI-1A - Units 1 and 2 Page 10 of 18 Rev. 8 IV. SUBSEQNEUT OPERATOR ACTION (continued)

2. If the main condenser is NOT available, dump steam to the atmo-sphere with the steam relief valves to lower the reactor coolant temperature and consequently the reactor coolant pressure.

CAUTION: Reduce the S.G. pressure 200 psi below safety valve setpoint (864 psig) while reducing reactor coolant system temperature approximately 50*F/hr. CAUTION: Use only PAM identified instrumentation for monitoring and evaluating plant conditions. NOTE: Monitor the condensate storage tank levels.and be prepared to switch over to alternate source if needed. K. Just prior to going on recirculation mode, reset the safety injection signal, if not previously reset. L. Verify that the cold leg accumulators dump their contents into the reactor coolant system if RCS pressure decreases below E 400 psig. M. Stop reciprocating charging pump if it is still running. N. Verify A-A and B-B containment air return fans in service (10 minute TD) after Phase B actuation and suction dampers ZS-30-72 and ZS-30-73 i open.

1. If containment spray has been actuated, and if the containment pressure is reduced to nominal operation pressure, reset contain-ment spray. Spray pumps should be shutoff and placed in the standby mode with operable flow paths.

CAUTION: 'If a blackout occurs after switching to the recirculation mode, see Part B of this instruction. NOTE: If RWST low level alarm is not imminent, then consideration should be given to performing a perliminary evaluation of the - plant status as follows:

1. Periodically check auxiliary building area radiation monitors for detection of leakage. If significant leakage is detected attempt to isolate the leakage and implement the required radiation protection. The injection flow to the RCS must be maintained at all times.

l 1 .. . -..e _ . - . . . ~ - - - . - e eewsg96 _ o , SQNP E01-IA - Units I and 2 Page II of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (continued)

2. While the plant is in cold leg injection mode, make pro-visions for an evaluation.of equipment in the plant. This evaluation should include the primary safeguards equipment e.g. , RCS pumps and valves, D/G's, containment coolers, etc. ,

and support equipment, e.g., ECCS HVAC equipment, diesel l fuel supply, diesel start air supply, sampling of RCS for l boron concentration and fuel damage, sampling of contain-l ment atmosphere.

3. Prior to the time specified for the plant switchover to the the cold leg recirculation mode:

a. Ensure the control room valve switches...are aligned in the proper position for cold leg recirculation mode.

b. Re-energize breakers, as required, for valves.needed to affect switchover.

I P. The following automatic phase of switch-over from the injection to the recirculation mode is initiated when the RWST is at low level E 29% (120,000 gal) coincident with a containment sump level of E 10%. NOTE: The times provided at the end of subsequent steps are normal times for the valves to travel full stroke. l NOTE: Do not interrupt the changeover operation until all actions , are completed. If a valve fails to respond or to complete its demanded operation, postpone any corrective action until the subsequent steps are performed except as noted. Loss of one complete train of power will allow the other train to be swapped, and parallel valves required to be closed will re-quire one valve to be closed locally.

1. Verify that valves FCV-63-72 (A-A) and FCV-63-73 (B-B) RHR pumps suction from containment sump, start to open while the RHR pumps continue to run. (2 min)

CAUTION: If a containment sump valve cannot be opened, stop the corresponding RHR pump.

2. Verify the valves FCV-74-3 and FCV-74-21, RHR pumps suction from RWST start to close. (2 min) t l

__ ___ .. . . . . -

• ee.-

a SQNP E0I-1A - Units 1 & 2 Page 12 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION (Cont.) Q. The following manual operations are done upon verification that the automatic switchover phase has begun.

1. Close the following RHR HX outlet crosstie valves.

a. Close FCV-74-33 I

b. Close FCV-74-35

2. Open the following component cooling valves: (Panel M-27B)

a. Open FCV-70-156 RHR HX. A outlet. .,.f60 sec.)

b. Open FCV-70-153 RHR Hx, B outlet. (60 sec.) *

3. Verify flow to the RCS from the safety injection pumps and close the following SI pump miniflow valves:

CAUTION: Repressurization of the RCS following performance of the following steps may cause damage to the SI pumps

if pumps are not removed from service or provided with an adequate flow path.

a. Close FCV-63-4 SI pump A-A miniflow. (10 sec.)

b. Close FCV-63-175 SI pump B-B miniflow (10 sec.)

c. Close FCV-63-3 SI pumps A and B miniflow to RWST.

(10 sec.) .

4. Verify that the automatic valve realignments in step P above have been completed. -

5. Open the following:

a. Open LCV-63-8 RHR HX A outlet to centrifugal charging pumps suction and SI pump A suction. (10 sec.)

~

b. Open FCV-63-ll RHR HX. B outlet to SI pump B suction.

, (10 sec.)

6. Close FCV-63-1 RHR pump suction header from RWST.

30TE: 480V breaker must be closed before operating valve from control room (Rx MOV bd Al-A) l p 66 h mMe= . si 9h - - - N m ""

.

SQNP EOI-1A - Units 1 & 2 Page 13 of 18 Rev. 8 IV. SUBSEOUENT OPERATOR ACTION (Cont.)

7. Open the following parallel valves:

a. Open FCV-63-6 RHR HX. A to SI pump A suction. (10 sec.)

b. Open FCV-63-7 RHR HX. A to SI pump A suction. (10 sec.)

8. After completion of the above steps verify that the two SI pumps and centrifugal charging pumps are receiving suction supply flow from the RHR pump,. and proper flow is established to RCS cold legs on all injection pumps.

9. Close FCV-63-5 safety injection pump suction from RWST

10. Close the following:

i

a. Close FCV-62-135 RWST to centrifugal charging pump suction. (10 sec.)

i

b. Close FCV-62-136 RWST to centrifugal charging pump suction. (10 sec.)

11. Periodically check auxiliary building area radiation monitors for detection of leakage from ECCS during recirculation. If significant leakage has been identified in the ECCS, attempt to isolate the leakage. The recirculation flow to the RCS must be maintained at all times.

R. After going into the recirculation mode, the containment spray pumps will continue using water from the RWST. Upon reaching the RWST low-low level alarm point (= 8% = 50,000 gallons), as indicated on the qualified, PAM indicator channels, realign the Containment Spray System. The following steps are required for the realignment of the Containment Spray System from the injection to the recirculation mode.

1. Stop both containment spray pumps (CSP) (" pull to lock in stop" to preclude the possibility of pump restart while realigning

, suction valves).

2. Close the following valves: l

a. Close FCV-72-22 containment spray pump' A-A suction from RWST (1 minute)

b. Close FCV-72-21 containment spray pump B-B suction

, from RWST. (1 minute) l l 9 . .. ; _. . . . . -

• o SQNP E0I-1A - Units 1 & 2 Page 14 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTIONS (Cont.)

• 3. Open the following ERCW valves (Panel M-27A):

a. Open FCV-67-125 containment spray HX A ERCW inlet.

(1 min)

b. Open FCV-67-126 containmer.t spray HX A ERCW outlet.

i (1 min.)

c. Open FCV-67-123 containrient spray HX B ERCW inlet.

(1 min.)

d. 'Open FCV-67-124 containment spray EX B ERCW outlet.

(1 min.)

4. Open the following CSP suction valves:

, a. Open FCV-72-23 containment spray pump A-A suction from containment Jump. (1 min.)

b. Open FCV-72-20 containment spray pump B-B suction from containmenc sump. (1 min.)

, CAUTION: If the pump suction valve frcm sump or the RHR containment sump valve will not open, do not re-start the corresponding containment spray pump. j

5. Observe CSP suction valves from containment sump full open. 4 above. (1 min.)

6. Start containment spray pump A-A (HS-72-27A).

7. Start containment spray pump B-B (HS-72-10A).

l

8. Verify at least 4500 gpm flow on CSP A-A (FI-72-34).

9. Verify at least 4500 gpm flow on CSP B-B (FI-72-13).

CAUTION: Monitor con'.ainment2 H e neentrati n n M-10 and if H concentration reaches % by volume, followstep"Y"of this instruction. CAUTION: DC NOT pe : form the part of following step (s) relative to EGTS till ECN-L-5124 is completed on flow switches FIS-65-31A-B-C-D and 55-A-B-C-D. Completion date is expected to be approximately June 1980. Until ECN is completed, i leave be,th trains of EGTS running until recovery from l acciden; is planned. ' l l j l SQNP EOI-1A - Units 1 & 2 Page 15 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTIONS (Cont.) S. Thirty minutes or longer after the accident occurs, place one of the operating emergency gas treatment trains in standby condition according to SOI-65.1B, and one of the operating auxiliary building gas treatment systems in standby operation according to SOI-30.6. T. After one hour stop the emergency diesel generators if offsite power has not been interrupted. i U. Approximately two (2) hours after a design basis accident, the ice in the ice condenser will be depleted. At this time, containment

pressure will increase and if containment pressure increases to 9.5 psig, place one train of RHR in containment spray mode, but no sooner than (2) hours into the accident.

CAUTION: Step U actions must not be initiated sooner than two (2) hours after the accident to ensure adequate flow to core. V. Sequence of operator actions required to establish RER containment spray flow on train A or B: ~ Train "A" RHR containment spray initiation.

1. Close the RHR cold leg isolation valve FCV-63-93.

s 2. Open the RHR containment spray header isolation valve FCV-72-40.

Train "B" RER containment spray initiation.

1. Close the RHR cold leg isolation valve FCV-63-94.

2. Open the RHR containment spray header isolation valve FCV-72-41.

W. Approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after transferring to cold leg recirculation, realign one train of ECCS for hot leg recirculation to assure against an excessive buildup of boric acid concentration in core. CAUTION: Due to possible cavitation in the vicinity of FCV-74-35 and , piping tee downstream of this valve under high flow and low head conditions; use Train A RER for hot leg recirculation. Use Train B RHR for hot leg recirculation only in extreme emergency. While Train A RHR is in hot leg recirculation, leave Train B RER in cold leg recirculation.

1. Train A RHR Changeover to H.L. Recirculation (Preferred)

a. Close FCV-63-93, RHR pump A-A cold leg isolation valve.

b. Open FCV-74-33, crosstie isolation valve between dis-

charge lines of RHR heat exchanger.

c. Open FCV-63-172, RHR pump H.L. isolation valve.

d. Verify H.L. flow on FI-63-173.

l J I SQNP E0I-lA - Units 1 & 2 Page 16 of 18 Rev. 8 s IV. SUBSEQUENT OPERATOR ACTIONS (Cont.) W. 2. Train B RHR Changover to H.L. Recirculation (Used Only if Train A Cannot be Used)

a. Close FCV-63-94, RHR pump B-B C.L. isolation valve.

. b. Open FCV-74-35, RHR pump B-B crosstie valve.

c. If not open, open FCV-63-172, RER pump H.L. isolation valve.

d. Verify H.L. flow on FI-63-173.

3. Realign the safety injection pumps to deliver through the hot leg injection headers to the reactor coolant system.

a. Stop A-A safety injection pump. .;y-

b. Close FCV-63-152, safety injection pump A-A crosstie isolation valve.

c. Open FCV-63-156, safety injection pump A-A hot leg i

isolation valve.

d. Start safety injection pump A-A and verify flow to the RCS through the hot leg header. (FI-63-151)

e. Stop B-B safety injection pump.

i

f. Close FC-63-153, safety injection pump B-B crosstie isolation valve.

g. Open FCV-63-157, safety injection pump B-B hot leg isolation valve.

h. Start safety injection pump B-B and verify flow to the RCS through the hot leg header. (FI-63-20)

j. Close SIS pump isolation valve FCV-63-22 with HS 22A.

X. Stop the containment spray pump when determined that the containment pressure will stay below 0.25 psig, if not previously done, reset phase "B" and place spray pumps in standby with operable flow path. Y. 'l . Place one train of H, recombiners in service'per SOI-83.1 if con-tainment H2 concentration reaches % by volume and diesels can carry the extra load (75 kW) or if offsite power is available but no later than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident to ensure a mixture of less than 4%, by volume, in containment. Place the other train of H, recombiner in service if containment H e a entrati a reaches 2 1% by volume. Do not exceed 4000 kW on diesel generator. (Each recombiner maximum capacity is 75 kW.) p .S & e h MWGM 6 o SQNP E0I-1A - Units 1 & 2 Page 17 of 18 Rev. 8 IV. SUBSEQUENT OPERATOR ACTIONS (Cont.)

2. Have the containment air sampled and in the event containment atmosphere reaches 3% 2H , by volume, the H2 purge system will also be placed in service per SOI-83.1.

. Z. Operational Guidelines

1. Guidelines for removing H fr a c ntainment ata 8P here.

2

a. Place H rec abiners in service.

2

b. Keep containment air return fans on.

-l'"

4

c. Use containment pressure relief and/or H 2Purge system when necessary.

2. Guidelines for removing H2 and other non-condensables from pri-mary system when trapped and cannot escape to containment.

a. Run RCP #1 or #2 (#2 preferred) if possible. Open spray valve periodically to strip non-condensables from water to

PZR vapor space. Then open PZR PORV momentarily to vent

, non-condensables to PRT where it can be vented to waste gas system or to containment if RPT rupture disc is blown.

b. Keep primary system pressure as high as possible to prevent bubble from enlarging and also to entrain more non-condensables in reactor coolant water. *

c. A slow, controlled depressurization can be utilized to grad-ually burp non-condensables into #2 hot leg from vessel upper head into PZR surge line where it can be vented to PRT. Also spray valves will be utilized to strip non-condensables from water in PZR vapor space. This should be accomplished slowly to prevent flow blockage from large quantities being liberated in a short time period.

d. If normal CVCS letdown can be accomplished, non-condensables can be stripped in VCT vapor space and then vented to waste gas system and stored in GDT's.

. CAUTION: This may be unadvisable if activity of primary coolant indicates fission products. V. RECOVERY ~ The reactor will remain in recirculation mode for an indefinite period of time. Instructions for recovery shall be prepared when the condition of the containment and the accident evaluation is determined. 6 M m 4 6 g@ 6 @h69OM a 6 SQNP EOI-1A - Units 1 & 2 Page 18 of 18 Rev. 8 VI. DISCUSSION OF EVENT l This emergency results in a breach of the primary system such that

leakage will develop enough differential, pressure between the upper and lower compartments to open the ice condenser doors. Safety in-jection and reactor trip will be initiated by the falling pressurizer pressure on a time scale dependent upon the magnitude of the break, or by high containment pressure, or manually. Injection flow, from the

• centrifugal charging and safety injection pumps, will increase with decreasing reactor coolant system pressure. The accumulators will auto-matically discharge their fluid inventory when the reactor coolant system pressure drops below E 1200 psig (upper head injection system) and E 400 psig (cold leg accumulators) In the event of rapid depressurization of the reactor coolant system leading to very low pressure,3the residual heat removal pumps will commence injection. Long-term control and cool-down of the primary system is by recirculation of spilled reactor coolant from

the containment sump. This is carried out by the residual heat removal pumps and by the centrifugal charging and/or safety injectica pumps taking suction from the outlet of the residual heat exchangers.

It is important in the use of this procedure that the UHI isolation valves go closed when the low level setpoint is reached to prevent intrcduction of the gas into the RCS. If not closed properly this could cause partial { flow blockage due to the nitrogen gas. Containment pressure increases because of the release of energy from the l primary system to the containment, and containment isolation will result t from the safety injection signal. The maximum containment pressure ex-pected under the worst conditions with ice condenser and containment spray functioning is approximately 8.0 psig and within six minutes the contain-ment pressure should decrease to less than 5.0 psig, at which time there will still be approximately 63 percent of the ice remaining. The contain-ment spray will be initiated at 2.81 psig, containment high-high pressure. The objectives of this instruction are to: A. Provide immediate core cooling and shutdown reactivity to ninimize damage to the fuel cladding and release of radioactivity. , B. Maintain long-term shutdown and cooling of the reactor by recir-culation of spilled reactor coolant, injected water, containment , spray system drainage, and melted ice. VII REFEEZNCES FSAR 15.31 ~ 6.3 l i l { ~ ~r: _ . . . . . . . . . -. t *

SQNP E0I-1B - Units 1 & 2 l Page 1 of 5 Rev. 8 STATION BLACKOUT WITH SIS RESET FOLLOWING A LOCA IN RECIRCULATION MODE j I. SYMPTOMS l A. Loss of normal control room lights and emergency lights on.

B. Loss of all loads connected to the 6.9-kV and 480-V unit board and 6.9-kV and 480-V shutdown board. C. "6900-V unit board 1A (lB,1C, and ID) (2A, 2B, 2C, and 2D) failure or undervoltage." (Setpoint 68 percent of normal voltage after 5 seconds time delay.) D. "480-V unit board 1A (lB) (2A and 2B) undervoltage or-transfer." (Setpoint 68 percent of normal voltage after 5 secon'ds time delay.) E. "6900-V SD Board 1A-A (lB-B) (2A-A and 2B-B) failure or undervoltage." (Setpoint 44 percent of normal voltage after 1.5 seconds time delay.) F. "480-V SD Board lAl-A (lA2-A and IB1-B and IB2-B) (2Al-A, 2A2-A, 2B1-B, and 2B2-B) failure or undervoltage." (Setpoint 68 percent of normal voltage a'fter 2 seconds time delay.) G. "6900-V start bus failure or undervoltage" on control building recording annunciator. (Setpoint 79 percent of normal voltage , after 2 seconds time delay.) II. AUTOMATIC ACTION A. Diesel generators lA-A, IB-B, 2A-A, and 2B-B start and close on their respective 6900-V shutdown board for boards that lost voltage. B. Blackout sequence initiated for boards that lost voltage. III. IMMEDIATE OPERATOR ACTION A. If a station blackout happens after the LOCA when the emergency core cooling system is in the sump recirculation mode and the , SIS has been reset, proceed as follows (only for the unit which  ; has the LOCA):

1. Verify that the diesel generators close in on the 6900V shutdown boards and the " Blackout Sequencer" starts sequencing loads for boards that lost voltage.

2. Lockout centrifugal charging pumps prior to (preferred) or just af ter they start to prevent pump cavitation until RHR pumps are restarted.

3. Secure the pressurizer heaters (all) "off" to prevent over-loading the diesel generators.

O

• W 46 $ SM- @e M6 i dD Me 4 +66 thw ee

1 . a SQNP E0I-1B - Units 1 & 2 l Page 2 of 5 Rev. 8 III. IMMEDIATE OPERATOR ACTION (Cont.)

4. Start control air compressors A and B locally.

CAUTION: Do not reinstate the SIS, as this would cause manually re:rositioned valves or other equipment to go to the wrong condition for the post LOCA recirculation mode.

5. Watch diesel loads closely. Start the following as soon as possible without overloading the diesel generators. .If diesel loads get excessive, secure equipment in subsequent actions step A.1 below before starting additional equipment.

j, . - RHR pumps a.

b. Centrifugal charging pumps

c. SIS pumps

d. Containment spray pumps (containment pressure 2.81 psig)

e. Containment air return fans, if not running, and 9B has initiated.

6. Announce blackout on PA system.

IV. SUBSEQUENT OPERATOR ACTIONS A. 1. For the unit which underwent the LOCA, secure the following 'l _ loads OFF: (this is done to avoid later overloading the diesels). (a) Reactor lower compartment fans. NOTE: Will not come on j if 9B is not reset. (b) Control rod drive mechanism cooler fans. NOTE: Will not come on if 9B is not reset. (c) Boric ac.d transfer pumps

2. The following loads are fed from'the diesel generators.

Start equipment as necessary to establish conditions prior , to blackout. CAUTICEY: Do not delay the following starts, but at all times keep each diesel generator total load at or below 4000 Kw continuously and 4400 KW for 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. (a) Control building cleanup fan, if not running , (b) Control building pressure fan, if not running (c) Hydrogen recombiner (H2> %) SOI-83.1 W W N* * *

• NOh 6 De== G 44 & ._MM.*

• e SQNP E0I-1B - Units 1 & 2 Page 3 of 5 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION l

B. The following loads are fed from the diesel generators. Verify equipment operating or start as necessary to establish pre-blackout conditions: NOTE: Watch total loads closely. (See CAUTION above) i 1. Essential raw cooling water pumps 6

2. Component cooling water pumps

3. Containment spray pumps

. .y-

4. Control building main control room condensing unit

5. Control building electric board room condensing unit

6. Auxiliary control air compressor

7. Containment air return fans j 8. RHR pump AHU

9. Vent radiation monitor I 10. Spent fuel pit cooling pump

11. 125-V vital battery chargers 1.

12. Emergency lighting cabinet

13. Fuel handling area exhaust fan (after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

14. Shut down board room air conditioner chiller

15. Shut down board room elev. 734 AHU

16. Electric board room AHU

17. Radiation moniter and sampling

, l'8 . Safety injection pump AHU

19. Centrifugal charging pump AHU _

20. Control room air conditioner ANU

21. Containment spray pump AHU

22. Pipe chase AHU

23. Penetration rooms El. 669, El. 690, El. 714 AHU's O e 6. @ e 6 M4G D e e 44-@ 6heMO&* 4- 4

~

• SQNP E0I-1B - Units 1 & 2 Page 4 of 5 Rev. 8 IV. SUBSEQUENTbPERATORACTION B. (Cont.)

24. Auxiliary feedwater pumps

25. Shutdown transformer exhaust fans

26. Control building cleanup fans

27. Elev. 749 board room pressure supply fans

28. Elev. 749 board room AHU's

.j, ."

29. Elev. 749 board room condenser units

30. Elev. 749 board room compressors

31. Control building b .tery room elev. 669 exhaust fans.

32. Emergency gas treatment fans and heaters

33. Auxiliary building gas treatment fans and heaters

34. Emergency gas treatment room AHU

35. CCS and AFW AHU

36. Shutdown board room air conditioner pressure fan

37. Station air compressor

38. Control building pressure fans

39. Shutdown board room air conditioner CW pump

40. Essential raw cooling water strainer

41. Hydrogen recombiner

42. Boric acid transfer pump 4'3 . Boric acid tank heater 44 Motor operated valves _

45. Charging pump auxiliary oil pump

46. 24 volt microwave battery charger -

47. Diesel starting air compressor

48. Diesel battery board exhaust fan

+ e

• e*

SQNP E0I-1B - Units 1 & 2 Page 5 of 5 Rev. 8 IV. SUBSEQUENT OPERATOR ACTION B. (Cont.)

49. Diesel building exhaust fans

50. Diesel building fuel oil transfer pump

51. Diesel building lighting cabinet

52. 125-volt battery charger

53. Diesel engine day tank fuel transfer pump 54.

D" AERCW pump station trace heating cabinet

55. Room exhaust fan

56. ERCW pump station fan and heater

57. ERCW traveling screen and wash pump

, 58. Spent fuel pit pump AHU

59. Diesel building electric heat i

V. RECOVERY Once the required electrical loads are restarted to establish pre-blackout conditions, continue RCS cooldown in accordance with part A of this instruction. - VI. DISCUSSION OF EVENT If a station blackout happens after a LOCA when the emergency core cooling system is in the sump recirculation mode, operator action is required to restore plant conditions to pre-blackout conditions. Extra caution must be exercised so that the diesel generators are not overloaded as accident loads are applied. NOTE: Under LOCA conditions do not reset the Safety Injection Signal (SIS) , until the first manual action is required following the LOCA. This is to ensure initial water cooling of the core, regardless of whether normal or diesel power has to be used. The first procedural reset of the SIS takes place no earlier than about 16 minutes after the LOCA to initiate the last valve manipulations (closure of two parallel valves in the line feeding water from the refueling water storage tank to the centrifugal charging pumps) which completes switchover of the ECCS pumps to the recirculation mode. VII. REFERENCES FSAR 6.3 . - _ _ . ._ _e_. . - . . __. .,.,e we w---_ g- ee ee-o [e =.e-* mut* *= *-- o . SQNP EOI Units 1 & 2 Appendix A Page 1 of 1 , Rev. 8 SATURATIONSTEAbfTABI.E (Temperatures rounded to nearest *F) Sat. 50*F Sat. 50*F. PSIG Temp *F Subcooled PSIG Temp *F Subcooled 300 422 372 1350 584 534 350 436 386 1400 588 538 400 448 398 1450 593 543 , 450 459 409 1500 597 .y- 547 500 470 420 1550 602 552 550 480 430 1600 606 556 600 489 439 1650 610 560 650 497 447 1700 614 564 700 505 455 1750 618 568 750 513 463 1800 622 572 800 520 470 1850 626 576 , 850 527 477 1900 630 580 900 534 484 1950 633 583 950 540 490 2000 637 587 1000 546 496 2050 ' 640 590 1050 552 502 2100 644 594 , 1100 558 508 215'O 647 597 1150 , 563 513 2200 650 600 1200 569 519 2235 653 603 ~ 1250 574 524 2250 654 604 1300 579 529 Saturation temperatures may be read from hot leg temperature RTD's or incore T/C's. WG M emmE gS @M $ n .+ e sipuD w. a O .w. - SQNP E0I-l - Units 1 & 2 Appendix B Page 1 of 1 Rev. 8 4 PZR Her's available as follows: lA-A: 1) Manual operation allowed immediately following SI.

2) Auto operation allowed 90 sec. following blackout

3) Manual operation allowed 90 sec. following B0 and SI.

IB-B: 1) Manual operation allowed immediately following SI.

2) Auto operation allowed 90 sec. following blackout

3) Manual operation allowed 90 sec. following B0 and SI.

IC: 1) Manual / auto operation allowed following B0 by resetting BO signal.

2) Manual / auto operation allowed following SI by resetting

, SI signal. 1D: 1) Manual / auto operation allowed following B0 by resetting , B0 signal.

2) Manual / auto operation allowed following SI by resetting SI signal.

n i l 6 4 ~ D 0 SQNP E0I-l - Units 1 & 2 - Appendix C Page 1 of 1 Rev. 8 NATURAL CIRCULATION A. The following are guidelines to determine if natural circulation is taking , place in the primary system.

1. Core AT as read on wide range RTD's (hot and cold) or an indicated AT between W.R. cold leg and incore T/C's should be stable or dropping.

A relatively stable AT with values less than 55*F with a gradual de-i crease indicates natural circulation. 9"

2. Incore T/C's temperature indicating below saturation temperature for

the existing primary system pressure. -

3. Heat is being removed from the primary system by secondary system, i.e. , S/G's steaming and water being added to S/G's and secondary I system pressure near saturation for the primary system.

B. The following are guidelines to enhance natural circulation.

1. Keep S/G levels in narrow range (Tubes covered)

CAUTION: See caution at start of subsequent actions this procedure relative to Post Accident indication for S/G level require-ments.

2. Keep primary system pressure above saturation pressure for the existing hot leg (W.R.) temperature or incore T/C temperature if possible.

3. Use condenser steam dump or S/G PORV's to steam off and cool primary system at desired rate. -

9 SQNP 1EDI Units 1 & 2 Appendix D Page 1 of 1 Rev. 8 INSTRUCTIONS TO RESTORE CORE COOLING DURING A SM.U.L LOCA A. PURPOSE To specify precaution and operation actions aimed at restoring a condition of core cooling during a small LOCA. I B. ACTIONS

1. Throughout this instruction, continue efforts to provide safety in-

jection and/or charging flow to the RCS and/or feedwater flow to the S/G's. Attempt to operate equipment manually locally., if possible.

2. Continue monitoring of core exit thermocouples to determine effective-

'  ; ness of subsequent action.

3. Depressurize the RCS by:

i I

a. Dumping steam to the condenser, or

b.  !

If the condenser is not available, dump steam through the at- , mosphere relief valves, or, i CAUTION: Depressurization through use of the S/G's should only be attempted if there is an effective water level and 8 auxiliary or main feedwater is available.

c. Open the PZR PORV's only if:

I

1. SIS or charging.is available to deliver to the RCS.

l i l

2. RCS depressurization cannot be accomplished by steam relief from che S/G's. .

3. Feedwater is not available to maintain che steam generator secondary water level at an effective level.

4. If no means for RCS depressurization are available, or if the de-pressurization did not result in decreasing core exit theenocouple temperatures, then start a reactor coolant pump, if possible.

DAUTION: During the co'nduct of this instruction, the RWST level chould be monitored for switchover to cold leg recir- ) culation, if required. ) g, g __ _ emg( q 6-6 *MW -W' B GM SQNP E0I-l - Units 1 & 2 Figure 1 Page 1 of 1 Rev. 8 l l [ l _ _ . . HAS AN AUTC:!ATIC EVALUATE NEED FOR i REACTOR ) MANUAL REACTOR l IF NO t TRIP OCCURRED? TRIP , RETURNTOl )l NODIAL ' OPERATION ..;s - I YE3 YES s/ - Y I_ i VERIFY l MANUALLY INITIATE l REACTOR ( --- --- REACTOR TRIP TRIP s/ 4 EAS AUTOMATIC EVALUATE NEED SI INITIATION N0 s FOR MANUAL - IF NO OCCURRED? SI INITIATION s/ YES - IF YES C0 TO q s, AOI-6 ~ VERIFY MANUALI.Y SAFECUARDS - INITIATE N SEQUENCE SI CHECK PAMS V C0 TO FIGURE 7 ?! CURE I og - .. . - - . . . .. .- .__..____...__?_. . . .S^4TP - . .. l . ,- -

• E0I Units 1 and 2 Pace 1 of 1

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r- e .e , .r e C0 TO lsaww7IniT 7 " '. { LoN'M2 stC:'OARY c:c A:: r: cry: : I . _ ~ - . . - . . s _ _ _ _ _ _ . _ , ,_._.~ ___ . .. _ _ _ . _ . . <l~ - ~ ~ .-. ....... . - . DISTRIBUTION IC Plant Master File . ENERGENCY OPERATING INSTRUCTION Superintendent 1U Assistart Superintendent (Oper.) 1U Assistant Superintendent (Maint.) E0I-2 Administrative Supervisor Maintenance Supervisor (M) l Assistant Maintenance Supervisor (M) LOSS OF SECONDARY COOLANT Maintenance Supervisor (E) f . ' Assistant Maintenance Supervisor (E) Maintenance Supervisor (I) Results Supervisor Units 1 & 2 1C operations Supervisor Quality Assurance Supervisor Health Physics Public Safety Services Supv. Chief Storekeeper { Preop Test Program Coordinator Outage Director l r Chemical Engineer (Results) Radiochem Laboratory Instrument Shop Reactor Engineer (Results) Instrument Engineer (Maint. (I)) Mechanical Engineer (Results) Staff Industrial Engineer (Pl. Sys.) IC Training Center Coordinator , PSO - Chickamauga Engrg Unit - SNP Prepared By: J. R. Walker Public Safety Services - SNP IC Shift Engineer's Office Revised By: J. R. Walker IC Unit Control Room QA&A Rep. - SNP Submitted By: , Health Physics Laboratory Supervisor IU Asst Dir NUC PR (Oper), 727 E3-C IU Nuclear Document Control Unit, 606 E3-C PORC Review: IU Superintendent, WBNP Date Superintendent, BFNP Superintendent, BENP 1U NE3, W9C174C-K Approved By: Sapv., NPHPS ROB, MS. Superintendent JJL_,,NRC-IE:II ' Power Security Officer', - 620 CSTE-C Nuclear Materials Coordinator-1410 CU35 Date Approved: Manager. OP-QAEA Staff _LE Resident NRC Inspector - SNP

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f 1r Technical Support Center i Rev. No. Date Revised -Pa res, Rev. No. -- Da te Revised Paees .___..;. ,2/9/80 7 All - 8 2/14/80 5 9 All - The last page of this instruction is Number 16 - i l m - l 1 SG2 E0I Units 1 L 2 Page 1 of 1 Rev. 9 e FURPOSE The objectives of these instructions are as follows:

1. To establish stabilized reactor coolant syste= and stea= generator condi-tiens prior to plant cooldown.

2. To minimize the energy release due to the break by isolation cf the break where possible.

3 To prevent the ?"R safety valves from lifting by du= ping stea= fro: all stea= generaters to the =ain condenser when possible or te the nt=os-phere from the unaffected steam generators. !. . To ' elate the auxiliary feedvater flov to the affected stea= generator, to .axi=1:e au.xiliarf- feedvater flow to the intact stea= generaters, and =inimize the energy release. 5 To berate the reactor ecelant to establish and =aintain reactor shut-down =argin. >e i . 4 4 a M d' . w mRP e , SQNP E0I Units 1 & 2 Page 1 of 6 Rev. 9 I. IMMEDIATE OPERATOR ACTICNS Refer to section on I==ediate Operator Actions of EIO-0, I==ediate Actions and Diagnostics, if not already performed. II. }4A'iUAL AC'"ICNS : If not actuated, =anually init-Verify the actuation of stea=line isolation. iate stea=line isolation. III. SUESEQUENT OPERATOR ACTIONS . CAUTION: The diesels should not be operated at idle or =ini=u= load for extcmded periods of time. If the diesels are shut down, they should be prepared for restart. NOTE: The process variables referred to in this Instruction are typically monitored by more than ene instru=entation channel. The redundant channels should be checked for consistency while perfor=ing the steps of this Instruction. t i N0'"E : The pressurize.r vater level indication should always be sued in conjunc-tion with other reacter coolant syste= indications to evaluate system conditions and to initiate =anual operator actions. A. If reactor coolant pressure is above the lov head safety injection pu=p shut-off head, manually reset safety injection so that safeguards I equipment can be controlled by manual action. Ensure that containment isolation is maintained. Stop the lov head safety injection pu=ps and place in the standby mode and request perfor=ance of SI-268 to verify P L contact position (failure of P h contacts vill prevent reset of SI . signal). CAUTION : Vnenever the vide range reactor coolant pressure decreases below the lov head safety injection shutoff head, the lov head safety injection pu=ps should be =anually restarted to deliver fluid to the reactor coolant syste=,

i

• CAU'ICN: Aute=atic reinitiatics of safety injection vill not occur since

- the reacter trip breakers are not reset. C AU'" ION : Subsequent to this Step, should loss of offsite power occur, manual action (e.g. , =anual safety injection initiation) vill be required to load the safeguards equip =ent ente the diesel povered e=ergency busses. M w I G.

• e

^ ' SUP E0I Units 1& 2 , Page &Loff Rev. 9 I!!. SUESEOUE

• 0?IRATOR AC"' IONS (cont. )

3. Stop all reactor coolant pu=ps after high head safety injection pu=p operation has been verified and when the vide range reactor coolant pressure decreases to 1300 PSIG.

CAUTION: If component cooling water to the reactor coolant pu=ps is j isolated on a PHASE "B" contain=ent isolation signal, all reactor coolant pu=ps should be stopped within 5 =inutes because of loss t of motor bearing cooling. CAUTION: If the reactor coolant pu=ps are stopped, the seal injection flov ' should be =aintained. UCTE: The conditiens given above for stopping react,or coolant pu=ps should be continuously =cnitored throughout this instruction. UCCE: See Appendix "A" for guidelines on natural circulation if RCF's are tripped. C. Determine which stea= generator is affected by observing the individual stea=line pressures (?AMS). A lov steanline pressure co= pared to the , others denotes the faulted loop; terminate auxiliary feedvater to that stea= generator and verify =ain feedvater isolated. CAUTICU: Secondary syste= breaks inside contain=ent =ay cause PZR-PCRV(s). To fail open, should this occur, isolate associated block valve. j Secondary syste: breaks in area of S/G PCRVs r.ay cause their failure in open position, should this occur, isolate if possible. Should the PZ3-PORV fail open and not be isolable, go to IOI-1. UCTE: If no loop has a lov stea=line pressure ec= pared to the others and all steenlines have been isolated, determine if a break has occurred in the stea=line, in the =ain feedline or in any piping syste= that connects with the secondary pressure boundary. If no indication of , a break in the pressure boundary is found, go to Section III of E0I-0 and re-evaluate the accident with particular emphasis on the Loss of

Reacter Coolant. If a leak from the secondary systems is found, continue te follow these instructions.

4

• e

L sq:7p E0I Units 1 & 2 Pagel5.of3 Rev. 9 'III.-SUBSE;UE';T OPERATCR AC" IONS (eent.) D. Regulate the auxiliary feedvater flow to the steam generators to restore and/or =aintain an indicated narrow range steam generator water level ~ (PAMS) or indicated vide range level (PMIS) sufficient to assure that i , the steam generator tubes are covered. If less of secondary coolant is inside containment, maintain S/G 1evel between h0% and 71% on narrow range for possible instrument error. If water level increases in an unexplained =anner in one steam generator, go to E0I-3, Steam Generator Tube Rupture. i NCCE: Monitor the primary water supply (Condensate Storage Tank) for the { auxiliary feedvater pumps and upon reaching a lov level, verify auto, svitch over to ERCW at 6" level in condensate storage tank. If auto. , switch over does not' occur, manually switch over. l E. Monitor Refueling Water Storage Tank level (PMIS).

1. If containment spray has been actuated, and if the containment

pressure is reduced to no=inal operating pressure (-0.1 to *0.3 PSID) { reset containment spray. Spray pu=ps should be shut off and pl.tced in the standby = ode with operable flow paths. 4 [ 2. If a lov Refueling Water 5torage Tank level alarm (or29%) is reached i vhile the contain=ent spray pumps are still runting, reset contain- [ ment spray. Spray pumps should be realigned to the recirculatica mode using the procedure presented in table E-2.1 (can not be , achieved until RHR pu=ps t_re changed over to recircultion mode). , 3 The high head and lov head safety injecticn pu=ps should re=ain aligned to the Refueling Water Storage Tank. If the Refueling Water Storage Tank low level alarm (229%) is reached, reset safety injection. Realign all safety injection pu=ps to the cold leg ,recirculatien mode using the procedure presented in Table E-2.2. Note, if the reactor coolant system pressure is above the shutoff , head of the lov head safety injection (SI) pumps, stop these pumps and place in a standby =ede prior to transfer to cold les re-circulation. F. Safety injection shculd be terminated {: 1 NOTE: , The conditions given below for termination of safety injection should be centinuously =enitored throughout this precedure.

1. a.

One video range reactor ecolant te=perature TH (PM:S) is less than 350 F. ' AND

b. Wide range reacto'r coolant pressure (PM:S) is peater than 700 psig and is stable or increasing.

i 4 k T t , . , . - m.m. ,. - - - < - , - e . , . . gg77 E0I- - Units 1 and 2 Page h of S Rev. 9 III. SUBSEOUE' T OPED.ATOR ACTIO:G (cont. ) MC

c. PZR vater level (FAMS) is greater than 20% of span and rising.

i A?iD

d. The reactor coolant indicated subcooling is greater than ho F.

NOTE: If all vide range reactor coolant temperature indicators go l above 350 F vhen atte=pting to satisfy the conditions of FI, initiate SI =anually and continue operation until conditions of F2 er F3 are satisfied. OR

2. a. Containment pressure or contain=ent radiation er contain=ent recirculation su=n levels do not exhibit either abner = ally high or increasing readings.

AND . b. All vide range rgaetor coolant temperature T., (PA!G) are greater than 350 F, AID

c. Wide range reactor coolant pressure (?AMS) is greater than 2,000 psis, and is stable er increasing, 1 ATJ

d. Wide-range indicated water level'in at least one S/G is at er '

above Toy A?Q

e. P2R vater level (PAI5) is greater than 20% of span,

. xm

f. The reacter coolant indicated subcooling is greater than h0 F.

i . TOTE: If centainment pressure, er contain=ent radiation, or i containment recirculaticn su=p level exhibit either ! abnor-ay high or increasing readings when attempting ta satisfy the conditions of F2, initiate safety in-l p jection and centinue operation until the following cenditions are satisfied. CR c em, 4 e . Sqyp E01 Units 1 and 2 Page 5 of 6 Rev. 9 III. SUBSEQUEN* OPERATOR AC* IONS (cont.) 3 a. Contain=ent pressure g contain=ent radiation, o_r contain=ent recirculation su=p level exhibit either abnormally high or increasing readings. AND

b. All vide range rgaetor coolant te=perature TH "#*

greater then 350 F, A!TD

c. Wide range reactor coolant pressure (PAMS) is greater than 2,00 psig, and is stable or increasing.

i

A.70 i

d. Harrow range water level in at least one S/G is at or above h0%.

AND .

e. PZR vater level (PAMS) is greater than 50% of span, AITD f.

Thg?. h0 reactor coolant indicated subcooling is greater than THEN , 4 Reset safety injection-and stop the safety injection pu=ps not needed for normal charging and reactor coolant pu=p seal injection flov. CAUTION: If vide range reactor coolant pressure decreases by 200 psi . ' eg- ?"R vater level decreases by 10% of span frc= the point of safety injection ter=ination er reactor coolant sub-cooling drops belov'h0"F, Manually Reinitate safety injec-tion to maintain reacter coolant pressure and F1R level. Centrcl reactor coolant pressure to the nc=inal value which existed when safety injection was initially ter=inated (Ty or to a no=inal value of 2000equaltoorlessthan350}F). psig (Ty greater than 350 Go to EDI-0 to reliagnose the event. CAUTION: Stopping and starting of the high head safety injection pu=ps and the charging / safety injection pu=ps can cause pump =otor overheating or reduced =otor life. Hence, if pu=ps are restarted once after termination, an addi-tienal 15 ? of subcooling should be added to the required subcooling prior to the second ter=ination of the high head pu=ps. b I .* ** i E0I Units 1 & 2 Page 6 of f Rev. 9 III. SUBSEQUEZ CPERATOR ACTIONS (cont.) . 5 Place all non-operating safety injection pu=ps in the stand-by .= ode, and =airtain operable safety injection flovpeths. i I

6. Isolate flow to the reactor coolant syste= cold legs via the boren injection tank and establish normal charging.

7 Reset containment isolation (Phase A). Re-establish nor=al =akeup to maintain syste= pressure at values reached when' safety injection ' was ter=inated (TgG350 F) or to a monical value of 2000 psig (T 350 F). Ensure that water addition during this process does nok>resultindilutionofthereactorcoolantsyste=boroncon-centration.

8. Re-establish operation of the pressurizer heaters after verification of sufficient pressurizer level to assure coverage of the pressur-1:er heaters, e.g. through ec=parisons of pressuriser surge line, water space, and vapor space temperatures and maintain PZR vater level between.50% and 70% for instru=ent error if loss of secon-dary coolant is inside conta!n=ent. When system pressure can be controlled by pressurizer b sters, and contain=ent te=peratures are lov enough to ass re p: per operation of control syste=s, restore nor=al pressurizer level control.

CAUTICN: Should RCS temperature decrease below ND1" for the RX vessel, do not allow pressure to increase abovo required pressure - te=perature li=its. G. Monitor either the average te=perature indication of core exit ther- =occuples (if available) or all vide range reactor coolgn: te=perature T H (PAMS) to verify that RCS te=perature is at least go F less than set-uration temperature at RCS indicated pressure. IfSgFindicatedsub-cooling is not present, then atte=pt to esgablish 50 F indicated sub-j cooling by steam du=p fro = the stea= generators to the condenser or the . at=osphere. CAUTION: If steam du=p is necessary, reduce the stes= generator pressure to 86h psig (200 psi below the lowest stea= safety valve setpoint) and =aintain a reactor coolant cocidown rs e f of no more than 50 F/ER, consistent with plant nake-up capability. g t Stes= du=p should be initiated in the folleving manner. t 8

• I l

l l n .g 9 O SCe E0I Units 1 & 2 Page 7 cf 7 Rev. 9 III. SUESEOUZTT OPEP.ATOR AC"'ICNS (cent. )

1. Establish a 1cv path in at least ene stea=line in an intact loop (if possible) IF the =ain condenser is available and IF an uncontro-lied stea= release vill not be reinitiated upon opening the MSIV.

,. a. Transfer the stea= du=p syste= to stea= header pressure cen-t trol. F

b. Set the steam header pressure control setpoint to the pres-sure in the intact stea generator (s) at the time safety injection is ter=inated.

c. Close stea= real supply valve 1-500.

d. On the intact S/G(s), open the bypass varning valves for j the MSIV(s).

I I e. With MSIV differential pressure less than 100 psig, open f MSIV(s) on intact S/G(s). . f. With conditions stabilized, establish =ain turbine seals and vacuu: per GOI-2. _OR

2. IF_

F all stea=line stop valves are C10SD and cannet be reopened, ! the main condenser is not available, or the rupture is devnstrea= {' of the main stea=line isolation valves, du=p stea to the at= csp- [ here frc= the intact loops using the stea generator power operated i relief valves. Set each stea: generator power operated relief valve pressure centrol'setpcint tc the pressure in the intact stea: generator (s) at the time safety injection is terminated. If 50 F indicated subcooling cannet be, established or =sintained., then manually reinitiate safety injectien. Go to Section' III of - ECI-0 to re-evaluate the event, unless this re-evaluation has already been perfor=ed. I=ple=ent e=ergency plan as required H. I. Verify control rco: vent isolation (See SCI-30.13). J. Verify U-2 contain=ent equipment hatch temporary door closed (73hel.) K. Verify Fuel handling floor equipment transfer hatch cover in place (73L el to lower elevations). L. Place additional CROM cooling fans and icver contain=ent cooling fans in service if break is inside contain=ent and Phase 3 isolation has net yet occurred. M. Transfer NR L5 to 1 SR and 1 IR detector. 3 0 - SQNP E0I Units 1 & 2 Page 8 off Rev. 9 III. SUBSIOUENT 0?IRATOR AC"'ICNS_ (cont. ) N. '4 hen the reacter coolant temperature and pressure (PAMS) are stable, i borate the reacter coolant syste= to cold shutdown conditions, as ? necessary. O. After offsite power is available, establish the auxiliary syste=s neces-sary for a centrolled cooldown to dold shut-down. If offsite poser is available and all reactor coolant pu=ps are stepped, restart at least ene reactor coolant pu=p in an intact lecp (with the pressurizer spray line If possible) for cocidown purposes in accordance with procedures, i Maintain subcooled conditions in the reactor , coolant system consistent with the nornal cooldown curve. If these subcooled conditions cannot be =aintained, restart safety injection pumps. NOTE: If there is significant radioactivity in ene er = ore stea= gen-erster's secondary side due to tube leaks and stea= is being du= ped to the at=csphere, im=ediately isolate .the stes: generator associated with the break. If all steam generators with sign-ificant radioactivity cannot be isolated, begin cocidown and depressurization of the reactor coolant syste= to li=it the re-lease of radioactivity to the environs. . j P. Stop D/Gs after2'30 minutes (socner if popossible) if not needed. NOTE: Safety injection pu=p operation should be reinitiated if an un-centrolled reactor coolant syste= depressurizatien er an uncontro-11ed drcp in pressurizer water level occurs during the cooldown process. These criteria apply in lieu of those given in Step F. Q. After establishing operation of auxiliary syste=s, initiate a controlled cooldown and depressuri:stion to cold shutdown conditiens using Nor=al Cooldown Procedures. NOTE: Safety Injection should be reinitiated if an uncontrolled reactor . coolant syste= depressurication or an uncontrolled drop in pressurizer water level occurc daring the cooldown precess. These criteria apply in lieu cf those given in Step F. NOTE: During the centrolled cocidewn, the reactor coolant syste= pres-I l sure vill decrease belev 1300 psis. Tripping the operating reactor coolant pu=p(s) due to the pressure criterion of Step 3 is not required. Other criteria of Step 3 are still appliable at this time. R. Recovery precedures for the aprticular event =ust be developed and i=plemented to effect plant return to service. j cg w

• e

• S^4TP E0I Units 1 & 2 Page 1 of.1 Rev. 9 NATUPaL CIRCUIATION OPEPATIONAL GUIDELINES A. The following are guidelines to determine if natural circulating is taking

' . place in primar/ syste=. j. I

1. Core T as read on vide range RTD's (het and cold) or an indicated T between WR cold leg and incere T/C's, should be stable and te=perature dropping a relatively stable T vith valves less than 55 F vith a gradual decrease indicates natural circulation.

2. Incore T/C's te=perature indicating below saturation te=perature for the existing primary syste= pressure.

3. Heat is being removed for= primar/ syste= by secondary syste=

1.e. , SG's stes=ing and water being added to SG's, and secondar-/ syste= pressure near saturation pressure for the primary syste: temperature. l s 3. Instructions to enhance nstural circulation. ' , 1. eep SG 1evels in narrow range (tubes covered), between h0% and 71% i for post accident instru:nent errer. 2. Keep pri=ary syste= pressure above saturation pressure for the existing het leg (WR) or incere T/C te=perature if possible. 3 Use stes= du=p or at=cspherie reliefs to stea= off and cool primary syste=. G 9 4 I l l ) b r - ..a_ _m_1:.c ," SQNP E0I Units 1 & 2 Page 1 of E Rev. 9 CONTAINMENT SPRAY Sk'ITCHOVER TO RECIRCULATIC'i MODE A. With a lov RWST level alar = (cd29%) and contain=ent spray pumps still running, reset contain=ent spray with ES-72-h2 & 72-h3 and: . 1. Stop both contain=ent spray pu=ps (CSP) (" pull to lock in stop" to preclude the possibility of pump restart while rerligning suction valves,). 2.

2. Close the following valves: ,

j a. Close FCV-72-22 containment spray pu=p A-A suction frc= RWST (1 r.1nute)

b. Close FCV-72-21 containment spray pu=p 3-3 suction from RWST (1 minute)

3. Open the following ERCW valves (Panel M-27A):

a. Open FCV-67-125 contain=ent spray HX A ERCW inlet (1 rin)

b. Open FCV-67-126 contai::=ent spray HX A ERCW cutlet (1 min)

'. c. Open FCV-6T-123 centain=ent spray HX 3 ERCW inlet (1 =in)

d. Open FCV-67-12h contain=ent spray HX 3 ERCW outlet (1 =in)

h. Open the following CSP suction valves:

a. Open FCV-72-23 contain=e'nt spray pu:2p A-A suction .

fro containment su=p (1 =in) l

b. Open FCV-72-20 centain=ent spray pu=p 3-3 suction fro contain=ent su=p (1 =in)

CNrIC's : If the pu=p suction valve fro = su.p or the RHR contain- =ent su=p valve vill not cpen, do not restart the corresponding containment spray pump. er M . _= . . - _ . . .- . - - . . - - SQNP  ! . E0I Units 1 & 2 Page 2 of a l Rev. 9 OONTAITENT SPRAY S*vCTCHOVER TO RECIRCULATICN MODE (cent) TABLE E-2.1 A. 5 Observe CSP suction valves frc= contain=ent su=p full cpen. Step L above. (1 min) t' 1

6. _ Start containment spray pu=p A-A (HS-72-27A)

I

• 7 Start contain=ent spray pu=p 3-3 (HS-72-10A)

8. Verify at least h500 gym flow en CSP A-A (FI-72-34) 4 9 Veri."y at least h500 gym flov en CSP 3-3 (FI-72-13) i

'i l e I s l e 9 I ? e ? l l' 1 ) 15 9 , - . . ,- 7.-,,,. -.- SqNF E01 Units 1 & 2 Page 1 cf 4 Rev. 9 TABLE E-2.2 COLD LEG F.ECIRCULATICN S*JITCHOVER INSTRUC"'ICNS I. OPERATICNAL STEFS A. The following automatic phase of switch-over from the injection to the recirculation mode is initiated when the RWST is at icv levelu29% (120,000 gal) coincident with a containment su=p level of M O%. , NOTE: "'his times provided at the end of subsequent steps are ner=al f times for the valves to travel full stroke. NOTE: All operator actions must be perfer=ed expeditiously, in a precise, orderly sequence. Do not interrupt the changeover operatien until all actions are completed. If a valve fails to respond or to ecm-pleted its demanded operation, postpene any corrective action . f until the subsequent steps are performed except as noted. Loss of , one ec=plete train of pcuer vill allev the other train to be swapped, and parallel valves required to be closed vill require one valve to be closed locally. ' ' "erify that valves FCV-63-72 (A-A) and FCV-63-73 (3-3) RER pu=ps suction from contain=ent sump, start to open ' while PER pumps continue to run. (2 =in) ' CAUTION: If a containment su...p valve cannot be opened, step the corresponding PER pump.

2. Verify the valves FCV-7h-3 and FCV-Th-21, PER pu=ps suction from RWST start to close. (2 min)

3. The following manual operations are dene upon verification that the -

automatic switchever phase has begun. CAUTION: Immediately stop any pu=ps taking suction frc= the RWST cn indication of the RWST being empty. Complete the switchover 1 steps listed below, then restart required pu=ps. I 1. Close the following RER EX outlet cresstie valves. l . a. Close FCV-7h-33 (ho see) i

b. C1cse FCV-Th-35 (hosec) 1 1

l l a e o e SQNF E0I Units 1 & 2 Fage 2 of 4 Rev. 9 TABLE E-2.2 I. OPERATIONAL S"EPS (cont)

2. Open the follocing ec=penent cooling valves: (Fanel M-273)

a. ' pen FCV-70-156 RER EX. A outlet (60sec)

b. Open FCV-70-153 RER EX. 3 outlet (60 ecc) 3 Verify flow to the RCS frc= the safety injection pu=ps and close the folleving SI pu=p =iniflow valves:

i Close FCV-63 k SI pu=p A-A =iniflov (10 see) a.

b. Cicse FCV-63-3 SI pu=ps A and 3 =iniflow to RWST (10 see)

k. Verify that the a_,c=stic valve realign =ents in step T above have been ec=pleted.

5 Open the folleving:

a. Open FCV-63-8 RER EX A outlet to centrifugal charging j- pumps suction and SI pu=p A suction (10 see)

I I b. Open FCV-63-11 RER EX 3 outlet to SI pu=p 3 suction (10 see)

6. Close FCV-63-1 RER pu=p suction header frc= RWST (2 =in)

NOTE: k80V breaker =ust be closed before operating valve frc= li control roc = (Rx MOV bd Al-A) - j, T. Open the following parallel valves:  ; s l I a. Open FCV-63-6 RER EX A to SI pu=p A suction (10sec) l

b. Open FCV-63-7 RER EX A to SI pu=p A suction (10 see)  !

l

8. After ec=pletion of the above steps verify that the two l Si pu=ps and centrifugal charging pu=ps are receiving i ruction supply flov frc= the RER pu=p, and proper flov is )

estat113hed to RCS ccid legs en all inject 1cn pu=ps. CAUTICN: Do not perterm steps 9 and 10 until the above varificatien ~ ) is made. I 14 m sqny . E0I Units 1 ti 2 Page 3 of4 Rev. 9 TA3LE E-2.2 I. CFERATICNAL S~TPS (cont) 9 Close FCV-63-5 safety injection pu=p suction fro: RWST (2 =in)

10. Cloase the fo11cving:

a. Close FCV-62-135 RWST to centrifugal charging pu=p suction (10 see)

b. Close FCV-62-136 RWST to centrifugal charging pu=p suction (10 see) i

' 11. Periodically check auxiliary building area radiation monitors for detection of leakage frc= ECCS during recirculation. If significant leakage has been identified in the ECCS, atte=pt to isolate the leakage. The recirculation Flow to the RCS =ust be maintained e.t all ti=es, i ! 12. 'a~hile the plant is in cold leg recirculation = ode, =ake provisiens for an evaluation of equipment in the plant if not previcusly done. f II. VERIFICATION: t A. After ec=pleting the preceding steps, verify that the safety injection syste= is aligned for cold leg recirculation as follows:

1. One icv head safety injection pump is delivering frc= the contain-

=ent recirculation su=p directly to two reactor coolant syste: cold legs and to the suction of two charging / safety injection pu=ps. I 2. The other lov head safety injection pu'=p is delivering frc= the , centain=ent recirculation su=p directly to two reactor coolant f system cold legs and to the suctice of two high head safety in-jection pu=ps. [ i 3 The two high head safety injectics and two charging / safety in-4 jection pe=ps are taking suction from the lov head safety injection pu=ps and tre delivering to fcur reacter coolant syste= cold legs.

h. The suction paths frc= the RWST to all safety injection pu=ps have been isolated.

5 If contain=ent spray is required, verify that flow is being del- - ivered.

3. If any failures have occurred, proceed to contingency actions.

• o SQi?

E0I Units 1 & 2 Page /)Fof 4 Rev. 9 TABLE E-2.2 III. CCICINGE'7CY AC"' IONS A. CO?!TAINME'IT RECIRCULATION SUMP VALVE FAILS TO OPEN If a containment recirculation su=p valve cannot be opened, stop the corresponding lov head safety injection pu=p and verify that:

1. One lov head safety injection pu=p is delivering flov to two

, reactor coolant syste= cold legs and to the suction of the two high head safety injection and two charging / safety injection , pu=ps.

2. The two high head safety injection and the two charging / safety injection pumps are delivering to four reactor coolant system cold legs.

3. LOSS OF ONE TRAIN OF ELEC"'RICAL POWER NOTE: If the single active failure is the failure of one of the e=er-gency diesel generators to staM in conjunction with a LOCA

, and a 1 css of offsite power, electrical power vould not be i available to one' of the vital safeguard busses. As a conse-quence, all engineered safeguards equip =ent assigned to that corresponding electrical power train vould not be available for operation until power could be restored to that bus. The l instruction for switchover to cold leg recirculation, assu=ing i a train failure, is essentially the same as the instruction, which assumed no single failures. The operator could follow the

instruction which assumed no single failure, with the understand-ing that those valve.s, without power, do not have to be reposi-tiened.

It should be noted that if a train failed subsequent to. the initiation of the safety injection signal additional steps =ay ' be required. For efa=ple, if no failure is assu=ed, the parallel suction valves in the line fro = the RWST to the charging / safety injection pu=p suction header vould open on a safety injection signal. Should a subsequent failure of one of the electrical trains occur, one of the parallel suction valves could not be closed frc= the =ain control board. Therefore, position isola-tien of the RWST to charging / safety injection pu=p suction path would have to acco=plished locally. O 3 __ . _ . ._ o , . Sequoych Nucles? Plant DISTRIBUTION EMERGENCY OPERATING INSTRUCTION IC Plant Master File Superintendent 1U Assistant Superintendent (Oper.) E0I-3 1U Assistant Superintendent (Maint.) Administrative Supervisor STEAM GENERATOR TUBE RUPTURE Msintenance Supervisor (M) Assistant Maintenance Supervisor (M) Maintenance Supervisor (E) Units 1 & 2 ' Assistant Maintenance Supervisor (E) j Maintenance Supervisor (I) Results Supervisor g Operations Supervisor { Quality Assurance Supervisor s . Health. Physics - - - - - Public Safety Services Supv. Chief Storekeeper Preop Test Program Coordinator Outage Director . Chemical Engineer (Results) Radiochem Laboratory Instrument Shop Reactor Engineer (Results) Instrument Engineer (Maint. (I)) Mechanical Engineer (Results) Staff Industrial Engineer (Pl. Svs.) 1c Training Center Coordinator PSO - Chickamauga Engrg Unit - SNP 1 Prepared By: George Wilson Public Safety Services - SNP ~ tc Shift Engineer's Office Revised By: J. R. Walker ic Unit Control Room QA&A Rep. - SNP Submitted By: ~ Health Physics Laboratory Supervisor 1tt Asst Dir NUC PR (Oper), 727 EB-C . PORC Review: It? Nuclear Document Control Unit, 606 EB-C 1tr Superintendent, WBNP Date Superintendent, BFNP Superintendent, BENP ~ - ~ 1 t' NEB, W9C174C-K -~ - Approved By: Supv., NPHPS ROB, MS Superintendent i t? NRC-IE:II Power Security Officer, 620 CST 2-C Nuclear Materials Coordinator-1410 CUBE Date Approved: Manager. OP-QA&A Staff , 1c Resident NRC Inspector - SNP sc NSRS, 249A HBB-K . . . .. _ ic Technical Support Center Rev. No. Date _ Revised Pages Rev. No. Date Revised Paees 5 2/9/80 All

• l 6 2/14/80 2

~7 All I l i I -The last page of this instruction is Number 12 - A h. e- .q e , . SQNP E0I Units 1 & 2 Page 1 of 1 Rev. 7 STEAM GENERATOR TUBE RUPTURE PURPOSE The objectives of these instructions are as follows:

1. To minimize the release of radioactive material by identifying and isolating the faulted steam generator and by reducing reactor coolant steam pressure below the staam generator safety valve settings.

2. To establish capability to supply feedwater to all steam generators and to isolate feedwater to the faulted steam generator.

l 3. To maintain the ability to remove the necessary residual heat from the reactor through the intact steam generators via the steam dump valves or power operated relief valves.

4. To maintain the reactor coolant system in a subcooled state during the recovery.

5. To prevent overflooding of the faulty steam generator.

I I 1 SQNP E0I Units 1 & 2 Page 1 of 9 _. Rev. 7 STEAM GENERATOR TUBE RUPTURE I. IMMEDIATE ACTIONS Refer to section on immediate actions or E0I-0, Immediate Actions and Diagnostics, if not already performed. II. SUBSEQUENT OPERATOR ACTIONS CAUTION: The diesels should not be operated at idle or minimum load for ex-tended periods of time. If the diesels are shut down, they should be prepared for restart. l . NOTE: If at any time during the conduct of steps A through H the faulted ' steam generator is positively identified, immediately proceed to step I. Following completion of this step, the remainder of the recovery must be accomplished from the last step of steps A through H which had been com-pleted prior to identifying the faulted steam generator. NOTE: Make arrangements to sample containment atmosphere and steam generators to identify presence of abnormal radioactivity. NOTE: The process variables referred to in this instruction are typically i monitored by more than one instrumentation channel. The redundant channels should be checked for consistency while performing the steps of this instruction. NOTE: The pressurizer water level indication should always be used in con-junction with other reactor coolant system idications to evaluate system conditions and to initiate manual operator actions. l A. Verify that all pressurizer power operated relief valves are closed.-- Verify the open status and availability of power to all pressurizer power operated relief valve backup isolation valves. I

' B. Stop all reactor coolant pumps after the high head safety injection pump operation has bee'n verified and when the wide range reactor coolant pres-sure decreases to 1300 psig.

CAUTION: If component cooling to the reactor coolant pumps is isolated on a containment Phase B isolation signal, all reactor coolant pumps are to be stopped within 5 minutes because of loss of motor bearing cooling. CAUTION: If reactor coolant pumps are stopped, the seal injection flow should be maintained. NOTE: The conditions given above for stopping reactor coolant pumps should be continuously monitored through Step J of this instruction. NOTE: See Table 1 or computer subcooling program for subcooling margin l and see Appendix A for guidelines on natural circulation. l 1 l 9 .) 'SQNP E0I Units 1 & 2 Page 2 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) C. If offiste power and the condenser are available, open bypass valves, equalize pressure across MISV(s), open any closed main steam line iso-lation valves to provide a flowpath to the condenser dump valves. D. Establish power sources necessary to operate at least one pressurizer power operated relief valve, at least one steam generator power operated relief valves, and charging and letdown flowpaths. NOTE: Ensure that containment isolation is maintained, i.e., not reset ' until such time as manual action is required on necessary process streams. E. Stabilize the reactor coolant system at approximately no-load temperature by steam dump to the main condenser if offsite power and the condenser are avaiilable. If offsite power or the condenser is not available, utilize the steam generator power operated relief valves to stabilize the l reactor coolant system at approximately co-load temperature. F. Regulate the auxiliary feedwater flow to the steam generators to restore and maintain steam generator water level in the narrow range span, or in the wide range span at a level sufficient to assure that the U-tubes are covered (76%). G. If reactor coolant system pressure is above the low head safety injection pump shut-off head, manually reset safety injection so that safeguards equipment can be controlled by manual action. Stop the low head safety injection pumps and place in the standby mode and request performance of SI-268 to verify position of P-4 contacts (failure of P-4 may prevent resetting safety injection). CAUTION: Ifthereactorcoolantsystempressuredecreasesudcontrollablp

. below the low head safety injection shut-off head, the low j head safety injection pumps must be manually restarted to de-liver fluid to the reactor coolant system.

t

• CAUTION: Automatic reinitiation of safety injection will not occur since the reactor trip breakers are not reset.

CAUTION: Subsequent to this step, should loss of offsite power occur, manual action (e.g. , manual safety injection initiation) will be required to load the safeguards equipment onto the diesel ~ powered emergency busses. H. Identify the faulted steam generator by one or more of the following methods:

1. An unexpected rise in one steam generator water level with auxiliary feedwater flow reduced or stopped.

s. SQNP E0I Units 1 & 2 Page 3 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) H. (Cont.)

2. High radiation from any one steam generator blowdown line radiation monitor.

3. High radiation from any one steam generator blowdown line, as de-termined by analysis or radiation detector.

4. High radiation from any one steam generator main steam line.

I. When the faulted steam generator has been positively identified, then:

1. Stop all feedwater flow to the faulted steam generator.

i !

2. Close the main steam isolation valve and bypass valves associated with the faulted steam generator.

3. Verify the closure of all power operated relief valves associated with the faulted steam generator.

4. Verify.the steam driven auxiliary feedwater pump is not being supplied with steam from the faulted S/G.

CAUTION: Do not proceed to step J until the faulted steam gen-erator has been identified and isolated. f NOTE: With faulted S/G isolated @ RCS temperature of 547*F, the faulted S/G pressure will be E 1000 psig. j' J. After the faulted steam generator has been identified and isolated, begin a cooldown of the reactor coolant system, the rate of cooldown should be'

relatively fast, but not so fast as to cause the UHI accumulators to dump J

on low pressure. Terminate the cooldown at 497*F on the RCS.

1. If offsite power and the condenser are available, dump steam to the

- main condenser from the intact steam generators by manual control of the steam header pressure controller.

2. If offsite power is not available or the main condenser is not avail-able, dump steam from the intact steam generators through the steam generator power operated relief valves. .

'K. Declare Gas Condition I or in the event blackout conditions exist, delcare Gas Condition II. L. Sound the plant radiological emergency siren to expedite assembly of per-sonnel and to reduce onsite doses to personnel. i SQNP E0I Units 1 & 2 Page 4 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) M. Survey meteorological information and dispatch the shift HP technician to survey the downwind sector at the plant boundary and request HP section perform survey of secondary site of plant. N. Transfer NR-45 to 1 SR and 1 IR detector. O. After the reactor coolant system temperature has beza reduced to 50 F i below the no-load temperature, if necessary begin a de-pressurization of the reactor coolant system to a value equal to the faulted steam gen-erator steam pressure while maintaining 50'F subcooling. ! NOTE: With RCS temperature @ 497*F, 50*F subcooling will be maintained down to E 1000 psig on the RCS. I j NOTE: During subsequent controlled reactor coolant system depressur-

ization, the reactor coolant system pressure criteria for tripping i the reactor coclant pumps established in step B DOES NOT APPLY.

j If the RCP's are in service, use the pressurizer spray to reduce the 1 pressure. If offsite power is not available, or the RCP's are not in service open

one pressurizer PORV to decrease pressure.

NOTE

Prior to the initiation of a controlled RCS depressurization, there may be no indicated PZR level. As the depressurization process proceeds, an increase in indicated PZR level is expected as liquid replaces steam in the PZR.

CAUTION

Monitor containment indications to verify that a loss of reactor coolant other than the steam generator tube rupture is not in progress. If recirculation sump level or a containment sample (if available at this time) are not in the normal pre-event range, further accident recovery must be directed according to Emergency Instruction E0I-1, Loss of Reactor Coolant, step M (Small LOCA).

P. As the reactor coolant system pressure decrease's, due to the quenching of the steam by the pressurizer spary or due to the steam release through the pressurizer PORV, monitor the pressurizer water level indications and stop the depressurization operation:

1. If the indicated water level in the pressurizer rises above 50 per-cent of span OR l

2. As soon as the reactor coolant system pressure decreases to a value I equal to the faulted steam generator steam pressure.

SQNP E0I Units 1 & 2 Page 5 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) Q. When the reactor coolant system preasure is reduced to 1500 psig, iso-late the upper head injection system as follows or verify isolated:

1. Close FCV-87-21 and GAG

2. Close FCV-87-22 and GAG

3. Close FCV-87-23 and GAG

4. Close FCV-87-24 and GAG R. Isolate the cold leg accumulators by closing the following valves when RCS pressure drops below 1000 psig if content of accumulators has not

, . been dumped to RCS. l 9 NOTE: Power will have to be placed on these valves.

1. FCV-63-ll8

2. FCV-63-98

3. FCV-63-80 i l

4. FCV-63-67 S. After the depressurization operation has been verified to have been term-inated (using the pressurizer PORV stem-mounted position indicators or spray valve demand signal), continue to monitor the reactor coolant system ,

pressure and the pressurizer water level. , ___ _ _ _ , . _ .

l i ,

1. If the pressurizer water level continues to rise or remains nearly constant concurrent with a reactor coolant system pressure decrease,

- suspect leakage from the pressurizer steam space. Monitor the pres-i sure relief tank (PRT) pressure, temperature and level to identify  ; continuously increasing conditions. Close the PORV isolation valves  ; ~ if a reactor coolant leak to the PRT is identified. Monitor PRT con- i ditions to verify PRT integrity. 1 CAUTION: If pressurizer relief tank integrity is 16st, abnormal con-tainment conditions could exist and may not be true indi-cations of a continued loss of reactor coolant. If con-ditions of step S1 persist after closing the pressurizer PORV isolation valves, further recovery must be directed according to E0I-1, Loss of Reactor Coolant, step M. The conditions of step S2 must be satisfied before proceeding to step T. l l 1 e SQNP E0I Units 1 & 2 Page 6 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) S. (Cont.)

2. If the pressurizer water level subsequently continues to increase concurrent with a reactor coolant system pressure increase con-current with verified PRT integrity, the safety injection flow is greater than the leak.

Then, when reactor coolant system pressure has increased by at least 200 psi (after shutting the spray valve or verified closure of the pressurizer PORV) and an indicated water level has returned in the pressurizer, stop all operating safety injection pumps not needed for normal charging and reactor coolant pump seal injection flow. [ CAUTION: The diesels should not be operated at idle or minimum load for extended period of time. If the diesels are shut down, they should be prepared for restart. NOTE: Following termination of safety injection, pressurizer pressure should decrease to a value equal to the faulted steam generator steam pressure. T. Place all safety injection pumps in a standby mode and maintain operable i safety injection flow paths. U. Verify main control room ventilation isolation (See SOI-30.lB). V. Verify U-2 containment eqdipment hatch temporary doors closed (734' El.) W. Verify fuel handling floor equipment transfer hatch cover closed (734 El. ! to lower elevations). . X. Re-establish charging and letdown flows to maintain the pressurizer water level in the operating range (approximately 25 percent indicated level): CAUTION: If, during subsequent recovery actions, pressurizer water level cannot be maintained above 20 percent indicated level, manually initiate safety injection flow to re-establish pressurizer water level in the operating range. If pressurizer water level cannot be established by this method, return to step d and proceed with the instruction from that point.

1. Close seal injection water flow control valve FCV-62-89.

2. Open the charging pump normal suction valves FCV-62-132 and FCV-62-133 from the VCT.

3. Close the charging pump suction valves FCV-62-135 and FCV-62-136 from the refueling water storage tank.

7- . i - w SQNP E0I Units 1 & 2 Page 7 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (C'est.) X. (Cont.) , 4. Open the centrifugal charging pumps miniflow isolation valves FCV- } 62-98 and FCV-62-99.

5. Open the charging line isolation valves FCV-62-90 and FCV-62-91.

. 6. Open seal water heat exchanger inlet isolation valves FCV-62-61 and FCV-62-63.

7. Gradually open the seal injection water flow control valve FCV l 89. Adjust the seal water flow to 8 gpm per RCP.

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8. Open letdown isolation valves FCV-62-69 and FCV-62-70.

9. Open the letdown line isolation valve FCV-62-77.

10. Open the 45 GPM letdown orifice isolation valve FCV-62-73.

11. Position PCV-62-81 to control pressure at letdown orifices above steam flash point.

12. 'Close the BIT inlet isolation valves FCV-63-39 and FCV-63-40 and outlet isolation valves FCV-62-25 and FCV-62-26.

I NOTE: Flush the injection lines and reestablished BIT concentration per AOI-19, IV., L thru M. Y. Re-establish the use of the PZR heaters to maintain the RCS pressure. If offsite power is available, establish the required conditions for -- operation of a reactor coolant pump and start the pump in a non-faulted , loop (preferably in loop 2 or if not available, in loop 1). If all RCP's are running, trip all but one RCP so as to maintain one pump operating in the loop connected to the PZR (loop 2), or if this is the faulted loop, in loop 1. Z. If offsite power is available, begin a controlled cooldown of the RCS at a rate of about 50*F/hr. by use of the steam dump to the main condenser from the non-faulted S/G's. Control the water levels in the S/G's ?.c maintain S/G water level in the narrow range span or in the wice range span at a level sufficient to assure that the U-tubes are covered (76%). If offsite power is not available, dump stemp from the non-faulted S/G's through the S'G PORV's to provide a controlled cooldown of the reactor coolant system at a rate of about 50*F/hr. 3 - e' , SQNP E0I Units 1 & 2 Page 8 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) AA. Simultaneous with the cooldown using the non-faulted S/G, slowly decrease the faulted S/G pressure by opening the MSIV bypass valve to the con-denser (if available), or using the S/G PORV. BB. As pressure is reduced in the faulted S/G, control the RCS pressure at a valae approximately equal to the steam pressure in the faulted S/G to , minimize the leakage flow. RCS pressure control should be accomplished by use of the PZR heaters and action of one of the following:

1. Normal PZR spray (if a RCP is in service)

OR l

2. Use of PZR auxiliary spray (if spray is heated by letdown through

. the regen. HX.) OR

3. Brief intermittant opening of one PZR PORV.

t NOTE: Maintain RCS temperature and pressure within the limits of the normal cooldown curves in TI-28. CAUTION: If RCS pressure control is accomplished by use of the PZR PORV, continuously monitor the PRT pressure, temperature, and water level .and take appropriate actions to verify and , maintain PRT integrity. Verify PZR PORV closure using the PORV stem-mounted position indicators, ACQUSTIC VALVE POSITION MONITORING SYSTEM and PRT conditions. _If a reactor coolant - leak to the PRT is identified, close the PORV isolation valves.' CC. Periodically sample and analyze the reactor coolant boron concentration during the continuing cooldown. Borate as necessary to maintain the required shutdown margin at all times during the cooldown. DD. Continue to cooldown and depressurize the reactor coolant system and faulted steam generator until the reactor coolant. hot leg temperatures are below 400 F in the non-faulted loops and the reactor coolant pres-sure has reached about 400 psig (do not collapse the presrurizer steam bubble). EE. Place the residual heat removal system in operation using Normal Cooldown Procedures. NOTE: Throughout this cooldown procedure, maintain a steam bubble in the pressurizer. Solid water pressure control may not be effective. _9 SQNP E0I Units 1 & 2 Page 9 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.) FF. Continue the plant cooldown in a normal mode except that after the RCP operation has been terminated, continue to simultaneously control the faulted steam generator steam pressure and reactor coolant pressure to minimize the leakage flow. GG. When the reactor coolant system hot leg temperatures are reduced below 200*F, the pressure in the pressurizer may be reduced by using auxiliary l j spray until reactor coolant system pressure and the faulted system gen-erator pressure equilibrate. j HH. Continue the operation of the residual heat removal system to remove j l the core residual heat and maintain the charging and letdown in service to control the pressurizer water level and provide a boration path. I III. RECOVERY l Following a steam generator tube rupture, the exact procedure will be planned l by the Plant Operations Review Committee for repairing the affected tube or tubes and decontamination of the secondary system. The procedure for repairing the affected steam generator will include necessary decontamination and exposure precautions for maintenance personnel. Decontamination of the secondary system will be carried out after the extent and type of radiation present is analyzed, with protection to plant personnel being the mose important consideration. IV. REFERENCES FSAR 15.4.3 l l l .i i l \ l l I l l ^ j i o SQNP E01 Units 1 & 2 ' Appendix A Page 1 of 1 Rev. 7 NATURAL CIRCULATION A. The following are guidelines to determine if natural circulation is taking place in the primary system. i

1. Core AT as read on wide ranga RTD's (hot and cold) or an indicated AT between W.R. cold leg and incore T/C's should be stable or dropping.

A relatively stable AT with values less than 55*F with a gradual de- , crease, indicates natural circulation.

2. Incore T/C's temperature iudicating below saturation temperature for the existing primary system pressure.

3. Heat is being removed from the primary system by secondary system, i.e., S/G's steaming and water being added to S/G's and secondary system pressure near saturation pressure for the primary system temperature.

B. The following are guidelines to enhance natural circulation.

1. Keep S/G levels in narrow range (tubes covered).

, 2. Keep primary system pressure above saturation pressure for the existing hot let (W.R.) temperature or incore T/C temperature if possible.

3. Use condenser steam dump or S/G PORV's to steam off and cool primary system at desired rate.

e j - l 5 r go aasog'S > f. .').,-.- .. , UNITED STATES NUCLEAR REGULATORY COMMISSION g g y - t w ASHINGTON, D. C. 20555 I Co%,'.w..*,/

• February 29, 1980 g

Docket No: 50-327 1 l Mr. H. G. "arris Manager of Power Tennessee Valley Authority 4 500A Chestnut Street Tower II Chattanooga, Tenne:see 37401

Dear Mr. Parris:

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SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNIT I - ISSUANCE OF LICENSE NO. DPR-77

- The Nuclear Regulatory Conintssion (the Commission) has issued the enclosed License No. OPR-77 to the Tennessee Valley Authority for the Sequoyah I Nuclear Plant, Unit 1, located in Hamilton County, Tennessee. License No. DPR-77 authorizes fuel loading and low power testing.

Also enclosed is a copy of the Notice which has been forwarded to tne Office 5 f.-

of the Federal Register for publication.

h( 1 Two signed originals cf Amendment No. 6 to Indemnity Agreement No. B-82 which covers the activities authorized under License No. DPR-77 are enclosed.

Please sign and return one copy to this office.

Sincerely,

~

.h" , SMk:

D. F. Ross, Jr., Acting Director '

Division of Project Management Office of Nuclear Reactor Regulation

Enclosures:

1. License No. DPR-77 ,

2. Federal Recister Notice

3. Amenament 6 to Indemnity Agreement B-82 c::s: See next page i

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UNITED STATES n r

• NUCLEAR REGULATORY COMMISSION j

(f 1 . ],

, wAsmmoToN. D. C. 20558

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-321 SEQUOYAH NUCLEAR PLANT, UNIT 1 LICENSE FOR FUEL LOADING AND LOW POWER TESTING

- License No. DPR-77

i. The Nuclear Regulatory Comission (the Comission) having found that:

l A. The application for licenses filed by the Tennessee Valley Authority comolies with the standards and requirements of the Atomic Energy Act ,

of 1954, as amended (the Act), and the Commission's rules and regulations '

set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. Construction of the Sequoyah Nuclear Plant, Unit 1 (the facility),

has been substantially completed in conformity with Drovisional Construction Permit No. CPPR-72 and the application, as amended, the provisions of the Act and the rules and regulations of the Comission; C. The facility requires exemptions from certain requirements of Appendices G and J to 10 CFR Part 50. These exemptions are described in the Office of Nuclear Reactor Regulation's Safe.ty Evaluation Report, Supplement No. 1. These exemptions are authorized by law and will not endanger life or property or the common defense and se:urity and are otherwise in the public interest. The exemptions are, therefore, I hereby granted. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the'Act, and the rules and regulations of the Comission; D. There is reasonable assurance: (1) that the activities authorized by this license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Comission; I

E. The Tennessee Valley Authority is technically qualified to engage in ]

the activities authorized herein in accordance with the rules and i regulations of the Comission; F. The Tennessee Valley Authority is financially qualified to engage in the activities authorized herein in accordance with the rules and regulations of the Commission; G. Thc Tennessee . Valley Authority has satisfied the applicable provf sions of 10 CFR Part 140, " Financial Protection Requirements and Indemnity l

Agreements", of the Comission's regulations;

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- H. The issuanco of this license will not be inimical to the common defense and security or to the health and safety of the public;

1. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of license No. OPR-77, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and J. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70, including

, 10 CFR Sections 30.33, 40.32, 70.23 and 70.31.

2. Pursuant to approval by the Nuclear Regulatory Commission at a meeting on Feoruary 28, 1980, license No. DPR-77 is hereby issued to the Tennessee Valley Authority to read as follows:

A. This license applies to the Sequoyah Nuclear Plant, Unit 1, a pressurized water nuclear reactor and associated equipment (the facility),

owned by the Tennessee Valley Authority. The facility is located in Hamilton County, Tennessee, about 9.5 miles northeast of Chattanooga, and is described in the " Final Safety Analysis Report" as supplemented and

, amended (Amenoments 14 through 63), and the Final Environmental Statement prepared by the Tennessee Valley Authority.

(-

S. Subject to the conditions and requirements incorporated herein, the I Commisson hereby licenses the Tennessee Valley Authority:

(1) Pursuant to Section 104(b) of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities", to possess, use, and operate the facility at the designated location in Hamilton County, Tennessee, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursua'nt to the Act and 10 CFR Parts 30, 40 and'70', to receive, cessess, and use at any time any byproduct, source and special nuclear material

, as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; -

l I

- , l (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

.(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the l conditions specified in the following Commission regulations in 10 CFR l Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40,  ;

Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and J is subject to all applicable provisions of the Act and to the rules, )

regulations, and orders of the Commission now or hereafter in effect; i i

and is subject to the additional conditions specified or incorporated below: \

l (1) Maximum Power level The Tennessee Valley Authority is authorized to (a) load fuel, (b) proceed to initial criticality, (c) perform startup testing at zero power in Operational Mode 2, and (d) after prior written approval Dy the Director of Nuclear Reactor Regulation, operate the facility e r- for testing at reactor core power levels not in excess of 170 Megawatts s- thermal (five percent of rated power). Prior to attaining that five percent power level, the Tennessee Valley Authority shall complete the items identified in Paragraph C(5) below to the satisfaction

• of the Director of Nuclear Reactor Regulation.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B attached l hereto are hereby incorporated in this license. The Tennessee Valley Authority shall operate the facility in accordance with the Technical Specifications.

(3) Initial Test Procram The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program that has been reviewed and approved by the Commission at the time of issuance of this license without making ,

l any major modification of this program. Major modifications are deemed to involve unreviewed safety questions under 10 CFR {50.59 and are defined as: .

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' a. Elimination of any test identified in Section 14 of the Final Safety Analysis Report as essential.

j

b. Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of the Final Safety Analysis Report as essential.

c. Performance of any test at a power level different from there described.

d. Failure to complete any tests included in the described program (planned or scheduled for nower levels up to the authorized power level).

(4) Fuel Load and Zero Power Testino Conditions The following conditions shall be complettd to the satisfaction of the Commission prior to fuel loading. The following conditions are related to matters specified in the TMI Action Plan, Near Term Operating License (NTOL) Requirements dated February 6, 1980, and applicable to fuel load and zero power testing. Each of the following conditions references the appropriate section of Part II of Supplement No. I to the Safety Evaluation Report (NUREG.00ll) for the Seouoyah Nuclear Plant and follows the numbering secuence utilized in the February 6,1980 NTOL Requirements list. The below designated sections of Part II of Supplement No. I to the Safety Evaluation

(] Report (NUREG-0011) are hereby incorporated by reference. In case of any inconsistency between the license and the Supplement to the Safety Evaluation Report, the terms of the license shall govern.

a. Shift Technical Advisor (I.A.l.1)

As defined and clarified in Section I.A.l.1, the Tennessee Valley Authority shall provide a Shift Technical Advisor on each shift with the duties and training set out therein.

b. Shift Suoervisor Duties (I.A.1.2)

As defined in Section I.A 1.2: (1) the senior Tennessee Valley Authority officer responsible for plant operations shall review the administrative duties of the Shift Suoervisor, and (2) administrative duties that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room.

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-S-c.

Shift Manning (I.A.l.3)

• - - In addition to the minimum plant staff shown in Table 6.2.1 of the Technical Specifications, Tennessee Valley Authority shall provide an additional senior reactor operator on shift at all times during reactor operation in Modes 1, 2, 3 and

4. The normal work station for this individual shall be in the control room.

- - Tennessee Valley Authority shall have administrative procedures to assure that qualified individuals to man the operational shifts are readily available in the event of an abnormal or emergency situation. These administrative procedures shall include provisions which limit the amount of overtime worked by licensed operators.

The need for a licensed operator to exceed the limits on overtime shall be infrequent. The limits on overtime werk hours are:

. An individual should not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight.

. There should be at least a 12-hour break between all work periods.

7_

'-- . An individual should not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

. An individual should not work more than 14 consecutive days without having 2 consecutive days off.

However, for those circumstances which arise requiring deviation from the above, such deviation may be authorized by the plant superintendent or high levels of Tennessee Valley Authority management in accordance with established procedures and with appropriate documentation of the cause.

d. Revised Scope and Criteria for Licensing Examinations (f.A.3.ll As further defined in Section 13.2 of Part I of Supplement 1 to NUREG-00ll, the Tennessee Valley Authority shall conduct a substantially augmented operations training program during the Sequoyah test program.

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' The Tennessee Valley Authority shall have additional contractor support to provide operational experience on each shift.

These individuals shall have at least one year of operating experience in large pressurized water reactors and a previous NRC license as a reactor operator.

At least one licensee's contractor startup engineer shall be on duty each shift as needed whenever the reactor is not in cold shutdown condition. The contractor startup engineer shall have previously participated in the startup of at least three pressurized water reactors.

e. Organization and Management Criteria (I.B.1.1)

In addition to the shift staff requirements indicated in Section 6.2.2.d of the Technical Specifications, Tennessee Valley Authority shall provide one health physics technician on shift at all times during reactor operation in Modes 1, 2, 3 and 4.

f. Safety Enoineering Groue (I.B.1.2)

The Tennessee Valley Authority shall establish an onsite Safety Engineering Group composed of a minimum of four engineering personnel who are not part of the plant staff and who shall be present on each day shift to act as independent observers. The

' /~ duties and responsibilities of this group are as follows:

- Be cognizant of the scope and intent of the test program.

- Be cognizant of the tests being conducted.

- Be familiar with the operation of a pressurized water type reactor.

- Provid. daily status reports to the Assistant Director of i Nuclear Power (Operations, TVA).

g. Licensee Onsite Ooer ating Exoerience Evaluation Cacability (I.B.l.4)

The Tennessee Valley Authority shall establish an onsite capaoility ,

to evaluate the operating history of the plant and plants of similar i design. l

- The corporate Nuclear Experience Review Panel (NERP) shall  !

j be a multidiscipline review group and shall review nuclear industry operational experience. l l

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- - The NERP shall inform the Sequoyah shift technical advisor o'f all matters so discovered applicable to Sequoyah.

4

- The shift technical advisor shall be the Sequoyah plant focal point for disseminating this information.

- This information shall be made a part of the reactor operational staff's routine upgrade training and shall be presented to operational personnel during initial training and retraining at the Tennessee Valley Authority Power Operations training center.

h. Shift Relief and Turnover Procedures (I.C.2)

The Tennessee Valley Authority chall develop and implement shift and relief turnover procedures that will provice assurance that the oncoming shift possesses adequate knowledge of critical plant status information and system availability. A checklist or similar hard copy will be completed by and signed by offgoing and oncoming shifts at each shift turnover. These checklists will be periodically reviewed by the operations supervisor or his assistant and will be held in the operations supervisor's office files for one month following review. The Tennessee Valley Asthority shall establish a ,

system to evaluate the effectiveness of the turnover procedures.

1. Shift Personnel Responsibilities (I.C.3) s- The Tennessee Valley Authority shall implement procedures which properly define the duties, responsibilities and authority of the shift supervisor and control room operators.

j. Control Room Access (I.C.4)

The Tennessee Valley Authority shall implement procedures that establish the authority and responsibility of the person in charge of access to the control room and that establish a clear line of authority and responsibility in the control room in the event of an emergency. -

k. Deoraded Core - Trainino (II.B.4)

The Tennessee Valley Authority shall establish a training program for the use of installed equipment and systems to control or  ;

mitigate accidents in which the core is severely damaged.

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1. Relief and Safety Valve Test (II.D.2)

The Tennessee Valley Authority shall carry out a testing program to qualify the relief and safety valves under expected operating conditions for design basis transients and accidents as provided in NUREG-0578, Section 2.1.2, as clarified in NRC letter to operating license applicants dated November 9,1979.

m. Relief and Safety Valve Position (II.D.5)

The Tennessee Valley Authority shall provide reactor system relief and safety valves with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. The indication shall comply with the requirements contained in NUREG-0578, Section 2.1.3.a. as clarified in NRC letter to operating license applicants dated Noverber 9,1979.

n. Auxiliary Feedwater Initiation and Indication (II.E.1.2)

The Tennessee Valley Authority shall provide for automatic initiation of the auxiliary feedwater system and shall provide for indication, in the control room, of auxiliary feedwater flow to each steam generator. These requirements shall comply with NUREG-0578, Sections 2.1.7.a and 2.1.7.b, as clarified in NRC

letter to operating license applicants dated Novenber 9,1979, s

o. Inadecuate Core Coolino - Subcoolino Meter (II.F.2)

The Tennessee Valley Authority shall provide a subcooling meter to provide on-line indication of coolant saturation condition. The meter shall comply with the requirements of NUREG-0578, Section 2.1.3.b, as clarified in NRC letter to operating license applicants dated Neverser 9,1979.

p. Inadecuate Core Coolino - Additional Instrumentation (II.F.2)

The Tennessee Valley Authority shall provide a design of additional instruments to provide an unambiguous indication of inadeouate core cooling. This requirement shall comply with NUREG-0578, Section 2.1.3.b, as clarified in NRC letter to operating license applicants dated November 9,1979.

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q. Emeroency power for Pressurizer Ecuioment (II.G)

The Tennessee Valley Authority shall provide emergency power for ,

the power-operated relief valves (PORY's), the PORY block valves I and pressurizer level instrument channels. This requirement shall comply with NUREG-0578, Section 2.1.1, sub-section 3.2, as clarified in NRC letter to operating license applicants dated November 9,1979.

r. IE Bulletins on Measures to Miticate Small Break LOCAs and Loss of Feeowater Acc1oents (II.K.1)

The Tennessee Valley Authority shall:

- Review the operating procedures and training instructions -

requested by Item 7(a) of IE Bulletin 79-06A. This review should ensure that operators have been instructed not to l override automatic operations of the engineered safety features, j unless their continued operation would result in unsafe j plant conditions, or until the plant is clearly in a stable, i controlled state, and engineered safeguards are no longer '

required.

- Revise the Sequoyah plant prccedures to include instructions to the operator for dealing with non-condensible gases in the

( primary system.

s. Imorove Licensee Facilities for Respondino to Emercencies - Onsite Tecnnical Succort Center (III.A.I.2)

The Tennessee Valley Authority shall establish a technical support center to meet the January 1,1980 recuirements of NUREG-0578, Section 2.2.2.b, as clarified by NRC letter to operating license applicants dated November 9,1979.

t. Imorove licensee Facilities for Respondino to Emeroencies -

Onsite Ooerational Sucoort Center (III.A.I.2)

The Tennessee Valley Authority shall provide an onsite operational support center to meet the requirements of NUREG-0578, Section 2.2.2.c, as clarified by NRC letter to operating license applicants dated November 9,1979.

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u. Uograde Licensee Emercency Precaredness (III.A.3)

During the period of this license, the Tennessee Valley Autnority shall maintain in effect an emergency plan that meets:

- Regulatory requirements of 10 CFR Part 50, Appendix E.

- Regulatory Position Statement in Regulatory Guide 1.101 (March 1977)

- The Essential Pla'nning Elements in NUREG-75/111 and Supplement 1 thereto defined by NRR as significant for fuel load and low power testing.

This plan shall provide an emergency operations facility as a base for coordinating onsite and offsite activities and iriterface with State, local, and Federal agencies.

v. In-Plant Radiation Monitorine - Partial (III.D.3.3)

The Tennessee Valley Authority shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under

_ y accident conditions.

' w. Consnunications (III.A.3.3)

The Tennessee Valley Authority shall install and maintain direct dedicated telephone lines between the Sequoyah plant control  ;

room, the Tennessee Valley Authority Emergency Operations Facility and the NRC Incidence Response Center in Bethesda, Maryland.

(5) Conditions on Operations Beyond Zero Power Testing The following conditions shall be coupleted to the satisfaction of the Commission prior to proceeding beyond zero power testing. The following conditions are related to matters specified in the TMI Action Plan, Near Term Operating License (NTOL) Requirements dated February 6,1980, and applicable to operation beyond zero power testing. Each of the following conditions references the appropriate section of Part II of Supplement No. I to the Safety Evaluation Report (NUREG-00ll) for the Sequoyah Nuclear Plant and follows the numbering sequence utilized in the February 6,1980 NTOL Requirements list. The below designated sections of Part II of Supplement No. I to the Safety Evaluation Report (NUREG-00ll) are hereby incorporated by reference. In case of any inconsistency between the license and the Supplement to the Safety Evaluation Report, the terms of the license shall govern.

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a. Short-term Accident Analysis and Procedure Revision Small breat LOCAs to Inaceouate Core Cooling (I.C.1)

The Tennessee Valley Authority shall revise its emergency operating instructions for dealing with small break LOCAs arid inadequate core cooling based on its analysis of these events and the vendor guidelines derived from these analyses.

b. Vender Review of Procedures: Low Power Test'ff.C.7)

I The Tennessee Valley Authority's low power test procedures shall be i reviewed by the nuclear steam supply system vendor, Westinghouse, )

and documentation of the review submitted to NRC. l l

c. Low Power Test Procram (I.G.1) l As set forth in Section I.G.1, the Tennessee Valley Authority shall octain staff approval of a low power test program.

D. The Tennessee Valley Authority shall maintain and fully implement the physical security plan entitled '" Physical Security Plan for the l Sequoyah Nuclear Plant" dated August 25, 1978, as revised on April 2,  ;

1979, June 29,1979, September 19, 1979, and as amended in actordance with the provisions of 10 CFR {50.54(p).

( In addition to all other comitments contained in the physical security plan, all keys, locks, combin1tions, and related equipment used to control access to protected and vital areas shall be controlled to reduce the probability of compromise. Whenever there is evidence that any key, lock comoination, or related equipment may have been comoromised it shall be changed. Upon termination of employment of any employee, keys, locks, combinations, and related equipment to which that employee had access, shall be changed.

E. This license is subject to the following additional condition for the protection of the environment:

Before engaging in additional construction or operational activities which may result in an environmental impact that was not evaluated by the Comission, Tennessee Valley Authority will prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than that evaluated in the Final Environmental Statement prepared

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by the Tennessee Valley Authority and the Environmental Impact Appraisal prepared by the NRC staff in May 1979, the Tenncssee Valley Authority shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation.

F. This license is effective as of the date of issuance and shall expire f one year after that date. )

l FOR THE NUCLEAR REGULATORY C07911SS10N )

1

$& $ s.,,,,

Harold R. Denton, Director l Office of Nuclear Reactor Regulation  ;

l

Attachment:

Aopendices A and B Technical Specifications i Date of Issuance:

February 29, 1980 r

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I o

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o.. , .. --

.- Sequoysh Nuclear Plant DISTRh, 1

_ E _ Plant Master File Superintendent g 1U Assistant Superintendent (Oper.)

, 1 Assistant Superintendent (Maint.) l ZERGINCY OPERATING INSTRUC"' ION Administrative Supervisor )

g Maintenance Supervisor (M) '

EOI-lh Assistant Maintenance Supervisor (M) '

Maintenance Superviser (E)

ANTICIPATED TRANSID:TS VITHOUT Assistant Maintenance Supervisor (E)

I SCPJLM Maintenance Superviser (I)

Results Supervisor 1C Operations Supervisor Quality Assurance Supervisor j k Health Physics t

_ Public Safety Services Supv. l Chief Storekeeper l j.' Preop Test Program Coordinator j Outage Director 1 Chemical Engineer g, Radiochem Laboratory  :

L Instrument Shop Reactor Engineer Instrument Engineer L

, Mechanical Engineer Staff Industrial Engineer 1 Training Center Coordinator o

PSO - Chickamauga Engrg Unit - SNP Prepared By: J.R. Walker Public Safety Services - SNP

_ ic Shift Engineer's Office Revised By: A J.P. Valke"

't_ J C_ Unit Control Room QA&A Rep. - SNP . . . . . .

c .. .

r Submitted By:

j f Health Physics Laboratory .

pervisor S Jq_ Chief, Nuclear Generation Branc'h l

atJ_ P Prod Central Office File PORC Review: #2. / 6 MO .

1U_ Superintendent, VBNP Ifa te / St.perintendent, BDIP b Supe rintendent, BENP

^ IU EN DES - NE3 NEG Approved By:

/

O N [la uperi h ndent Supv., NPHPS ROB, MS NRC-IE:II t

Power Security Officer, 604 PRB-C Date Approved: __ h//,7/20 Nuclear Materials Coordinator Manager, OP-QA&A Staff

, / / 1U P Prod Plant Eng. Branch 1C NSRS, 309 CB-K IC Technical Support Center .

1C Resident NRC Instecto-, SNP Rev. No. Date Revised Pages Rev. No. .Date Revised Pages i 0 12/28/79 All 1 2 All The last page of this instruction is Number b

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• SQNP EDI Units 1 & 2 Page 1 of 6 i

Rev. 1 g PURPOSE To provide the corrective actions in the event a reactor trip is not obtained following an automatic signal.

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Also to provide the corrective action in the event any of the immediate actions

, following a reactor trip are not achieved automatically.

• a l

I. SYMPTOMS

• g, A. The failure to obtain a reactor trip following an automatic signal.

. - 1. Rods not inserted as indicated by RPI's or the rod bottom lights. --

.L 2. No rapid drop in nuclear power as indicated by NIS readout, e.g.,

< 10*. of initial power in 5 seconds.

• L. II. IMMEDIATE OPERATOR ACTIONS A. Turbine Trip - Reactor Trip:

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1. Verify reactor trip, if not then immediately: - ~ - " '

a. Attempt to manually trip the reactor trip breakers by placing

... the reactor trip switches in the trip position, or

-..I. Gebu ~ ~ E

b. Immediately dispatch someone to the 480V unit boards A and B

.t to trip the breakers powering the control rod drive MG sets, or

c. Dispatch someone to the rod drive MG set room to trip the re-

,g actor trip breakers. -

- - ~ ~ ~

If a' reactor ' trip hafnot yet 'oEcurrei,7t'eiipt to insert

' d.~~

control rods manually and:

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1. Start centrifugal charging pump.

"" CAUTION: "B" ECCS equipment will not normally have com- 1 ponent cooling water aligned.

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2. Open charging pt=p suction valves from RWST and close '

suction valves from VCT.

3. Open BIT inlet and outlet valves. . ._-.__.. . . _ _ _ _ _ .

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4. Isolate normal charging path and charging pump recir-culation valves. -

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SQNP E01 Units 1 & 2 Page 2 of 6

/* Rev. 1 t II. IMMEDIATE OPERATOR ACTIONS

5. Decrease turbine load as boron and/or control rods

, decrease reactor power.

["

2. Verify turbine trip (turbine steam stop valves closed), if not then immediately:

L a. Attempt to manually trip the turbine by placing the turbine trip switch in the trip position, or f

( b. Immediately dispatch someone to the main turbine front standard to manually trip the turbine or, -

g* c. Immediately dispatch someone to EHC pump control station to stop and lockout both EHC pumps.

d. If a turbine trip has not yet occured, immediately close MSIV's.

L .

CAUTION: DO NOT open the main generator circuit breaker nor field breaker until the steam source is known to be L. closed and kw seneration is known to be zero or negative.

B. Determine Reactor Coolant System Status:

1. Verify AFW pumps running and flow established to S/G's if not then innaediately:

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a. AFW Pumps -

1. . Attempt to manually start A and/or B AFW pumps and/or

7. ; . , , . . . , _ . steam driven AFW pump from main control room, or.._.__ ..

2. Attempt to manually start' A and/or B AFW pump from 6.9-kV L S.D. boards and/or attempt to manually start steam driven AW pump locally.

c_ b. AW LCV:

MTR Driven AFWP LCV's .

NOTE: These valves fail open on loss of control air or control power. , ,

1. (aa) If not open, isolate control air and bleed down pressure or, (bb) If not open, remove control power or  !

1 2

e_ .- . . . . . . . . .. . ..

[ SQNP E0I Units 1 & 2 I

Page 3 of 6 Rev. 1 k.

[ II. IMMEDIATE OPERATOR ACTIONS

2. (aa) If open, control feedwater flow with motor driven pump I

PCV or k (bb) If open, use manual valve to control flow

}

Turbine Driven A WP LCV's

.[ NOTE: These valves fail closed on loss of control air or I control power I

3. (aa) If not open, use handwheel to open and control flow (bb) If open, use handhweel to control flow.

I c. MTR Driven AWP PCV: .. . .. - . - . . - . . .--

1 L 1. See SOI 3.2 V.D. for manual operating instructions.

2. PZR Pressure and Level Within Exoected Range, if not then immediately:

a. PZR Pressure I- High - -

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1. Check sprays on, heaters off and PORV's open if needed.

2. Check heat sink (S/G's) operable (at least one S/G in c , , . . narrow range er in wide range sufficient to cover S/G , . .;e_v tubes). - - - - - - -- - -- -- -

Low

' 1. Check sprays off, heaters on, PORV's closed and safety valves closed. -

2. Check PZR level normal 1

3. Verify and/or actuate SI if below 1870 PSIG. '

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- - - b c -- PZR Level - '

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l. Check charging system flow rate, letdown system in service and make adjustment to charging and letdown as needed.

2. Check PORV's and safety valves closed if pressure is normal.

Low

, 1. Adjust' charging flow to return level to normal

2. Secure letdown if desired

3. Place on additional charging pump if desired l . :-. . s.

- ~

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

SQNP E0I Units 1 and 2 Page 4 of 6 Rev. 1 II. IMMEDIATE OPERATOR ACTIONS (Cont.)

3. Tavg Controlling to 547'F, if not then immediately

, a. High Temp L 1. Verify steam dump valves opening and/or S/G PORV's opening, if not open

2. Attempt to open steam dump valves by going to pressure

[. mode control and manually opening

3. If steam dump valves are still not open, verify 540*F steam dump interlock reset.

4. If steam dump valves are still not open, use S/G PORV's to control Tavs @ 547'F.

5. Verify PZR pressure sufficient to provide subcooling requirement.

Z 6. Verify PZR level on scale.

, 7. Determine if a need to safety inject exists.

8. Verify S/G 1evels sufficient to provide heat sink.

1

b. Low Temu

1. Verify steam dump valves and S/G PORV's closed, if not closed

2. Attempt to close steam dump valves by going to pressure mode control and manually closing or, a 3. Place steam dump interlock selector switch (BOTH) in OFF position.

4. If steam dump valve (s) are still open, close MSIV's

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5. Attempt to close S/G PORV's by goirg to manual with controls and closing or by using PORV's handswitch and placing in close position.

6. If S/G PORV(s) are still open, remove control power from PORV or If S/G PORV(s) are still oped, send operator to manually

'7.

close or isolate

8. If RCS pressure is reduced to SI actuation setpoint, verify auto SI actuation and if not received, manually actuate.

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4. Containment Pressure and Temperature Normal, if not them immediately:

a. Containment Pressure

1. High Press (aa) Verify. containment humidity and radiation, if either are high see AOI-6 for excess leakage or (bb) Verify containment temp. normal, place additional coolers in service if needed

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' SQNP EOI-14 Page 5 of 6 a Rev. 1 II. IMMEDIATE OPERATOR ACTIONS (Cont.) ,

2. Lou Press a

(aa) Verify containment temperature normal, remove from

/- service connected coolers

,' b. Containment Temperature i

1. High Temo g, (aa) Verify sufficient number of upper and lower con-tainment coolers in service

, 2. Low Temo

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" (bb) Remove from service upper and lower containment coolers not needed L

5. Gen Breakers Open, 6.9-kV Unit Station Service Transferred, if not then immediately:

a. Gen Bkr's Doen -

1. Verify all steam sources are closed to main turbogenerator.

then attempt to open gen bkr's from main control room. . . - -

and/or elect. control room. ~~~ " - ~ ~~ -

2. If gen. bkr's are still not open, attempt to open them from switchyard local control.

b. 6.9-kV Unit Station Service Transferred, if not then immediately 1.

~ ' - - '~~~-

Verify start bus energized and then energize 6.9-kV unit boards per SOI-57.3 ~~"-~~ ~~~ ~~ ~

( 6. Steam Gen F.W. Regulator, Regulator Bvoass and Main Isolation Valves Shut, if not then immediately:

1, a. F.W. Rerulator

1. Attempt to close F.W. reg. valve (s) from main control j room by going to manual control and close.

2. If still open, attempt to close associated F.W. main l

isolation valve if not closed.  ;

[ 3. If unable to isolate in this manner, reduce main feed _ _ _ . . _ _ _ _ ;

I pump speed (disch. pressure) to point where pumps will l not deliver water to S/G's.

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t SQNP

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EDI-14 Page 6 of 6

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g Rev. 1 i L II.

( IMMEDIATE OPERATOR ACTIONS (Cont.)

b. Regulator Bypass I. 1. Attempt to close reg. bypass valve (s) from main control Foom

2. If still open, attempt to close associated F.W. main j* isolation valve if not closed

3. If unable to isolate in this manner, reduce main feed pump speed (disch. pressure) to point where pump will

._ not deliver water to S/G's.

t.

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c. Main Isolation .

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h 1. Attempt to close main isolation valve from main control room i

2. If still open, attempt to close associated F.W. reg. and

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reg. bypass if not closed.

3. If unable to isolate in this manner, reduce main feed pump speed (disch. press.) to point where pumps will not deliver water to S/G's.

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4. If desirable to continue feeding S/G's with main F.W.

attempt to close main isolation valve from 480V Rx MOV board or local by manual H.W.

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[ . In S:qusych Nucicer Plcnt DISTRIBUTION

< o 1 Plant Master File lm Superintendent 1 Assistant Superintendent (Oper.)

j 1 Assis tant Superintendent (Maint.)

i Administrative Supervisor AEI:C??.AI, OPERAT2;G EISTRUC"' ION Maintenance Supervisor (M)

Assistant Maintenance Supervisor (M)

I AOI-6 Maintenance Supervinor (E)

Assistant Maintenance Supervisor (E) b L EXCESS PRDiARY PLAN" LEATAGE Maintenance Supervisor (I)

Results Supervisor Units 1 and 2

. 1C Operat. ions Supervisor

. Quality Assurance Supervisor L Health Physics Public Safety Services Supv.

- Chief Storekeeper

[ Preop Test Program Coordinator Outage Directcy.-

. Chemical Engineer

. Radiochem Laboratory -

L Instrument Shop g Reactor Engineer g Instrument Engineer L Mechanical Engineer

, Staff Industrial Engineer e

j

,_1g_ Training Center Coordinator b

l PSO - Chickamauga Engrg Unit - SNP Prepared By: C.T. Benten Public Safety Services - SNP

,. 1C Shift Engineer's Office L Revised By: J.R. Walker 1C Unit Control Room L , QA&A Rep. - SNP

, Submitted By: __m of Health Physics Laboratory

~ ~

Sup'ep'isor 1U Chief, Nuclear Generation Branch y _1p_ P Prod Central Office File PORC Review: > /9/ga _,,,,1y,,_ Superintendent, WBNP Date Superintendent, BFNP Superintendent, BENP L

Approved By: J) dV_ _ ,. L 1U EN DES - NEB NEG Supv., NPHPS ROB, MS I Superintendent / NRC-IE:II

'L , Power Security Officer, 604 PRB-C

• Nuclear Materials Coordinator Date Approved: y/9/ga Manager, OP-QA&A Staff

~

__1p_ ? Prod Plant Eng. Branch

,,,_1,g_ NSRS, 3C9 GB-K

,, . 1C Technical Support Center

,, l.Q._ Resident NBC Increeter. SI ?

Rev. No. Date Revised Paces Rev. No. _Date Revised Pages k -

11/5/79 A11 5 12/21/79 11,12, added 12a 6 1,/9/9p A11 The last page of this instruction is Number 6

t SQNP J ,

AOI Units 1 and 2 Page 1 of 1 Rev. 6 b.

A PL7tPOSE To provide a discussion of the event, symptoms, automatic actions, it: mediate and subsequent operator actions and recovery from excess RCS leakage.

t A. Small RCS Ieak - A small leak is defined as one where the pressurizer level

. can be maintained by one charging pump.

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I SQNP AOI Units 1 and 2 l-g Page 1 of 4 Rev. 6 i SMALL REACTOR COOLANT SYSTEM LEAK

b. '

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I. SYMPToy A. Leakage Into Containment L- 1. Upper or lower containment building air monitor high radiation alam RE112A, 112B, RE106A, 106B l 2. Containment floor and equipment drain pump, and containment auxiliary L floor and equipment drain sump level rate of change alarm at an  :

inflow of greater than 1 GPM from computer. -b" 1

(~ 3. Containment floor and equipment drain sump and containment auxiliary floor and equipment drain su=p alarm on high level (M-5).

f-

. 4. Containment atmosphere humidity abnormally high alam.

L i g 5. More frequent than normal operation of the containment floor and

[ equipment drain sump pumps as indicated by the elapsed time meters L on switchgear.

6. Lower compartment temperature high.

B. Steam generator Tube Leakage I L

[ 1. Coedenser vacuum pump air exhaust, high radiation alarm. RE119 -

L.

2. Steam generator blowdown-liquid and/or sample monitor high activity if the leak is into the secondary system. RA124A, RA 120A b

C. Miscellaneous - - - -

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1. Reactor coolant pump (s) thermal barrier cooling water differential E l press high alam (one per pump) at 45 psid increasing or themal l barrier return temperature high alam.

l t 2. Component cooling system high radiati'on and/or surge tank level high alam. _

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CC surge tank level high 67.5*..

0-RA-90-123A ~

1-RA-90-123A

": 2-RA-90-123A l 1

L 2-

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b g.

g,- -

SQNP i

, , AOI Units 1 and 2 Page 2 of 4 L. Rev. 6 h."

3. Reactor vessel head seal leakage alarm at 20*F greater than g- ambient.

\- 4. Indication of leak during visual inspection of equipment.

i 5. Pressurizer relief or safety valves outlet temperature high alarm at L. 20 F. above ambient.

g 6. Auxiliary bulding vent monitor alarm RE101A and RI-101B.

L

7. Reactor coolant pump (s) stand pipe level high/ low..j,High 12" above L

, orifice outlet, low 12" below orifice outlet. ~

L- D. Reactor Coolant System Indications g[ 1. CVCS charging flow increases to maintain pressurizer level.

L

2. Increased makeup to CVCS G-NOTE: See attached Figure AOI-6.1 for aid in determining source of lea kage.

[L II. AUTOMATIC ACTION A. In the event a containment vent isolation occurs, the purge air dampers, containment pressure relief valves, and containment air monitor valves will g; close.

. 'B. In the event of CCS high radiation, the CCS surge tank vent valve, FCV 70-

] 66, automatically closes.

C.

In the event coolant loss into the auxiliary bulding results in high

']

Ya radioactive particulate or gas levels in the auxilia' ry building ventilation exhaust stack, the auxiliary building ventilation system will be isolated and the auxiliary building gas treatment system will be started.

." D .

In the event the leak is into the secondary system, high activity in blowdown effluent will automatically close the steam generator blowdown

' isol,ation valves to cooling tower blowdown and divert blowdown to conden-sate DI inlet.

E.

• In the event of RCP thermal barrier cooling coil leak, component cooling water will be isolated on high differential flow.

5.

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j SQNP AOI Units 1 and 2 Page 3 of 4 Rev. 6 III. I.*?E.DIATE OPERATOR ACTION I. ,A. If pressurizer level is falling, start charging pumps as necessary to maintain level.

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B. Determine if containment pressure, temperature, humidity, and/or radiation level is increasing.

IV. SUBSEOUENT OPERATOR ACTION CAUTION: If loss of PZR level is imminent, trip reactor, initiate safety injection and determine reason for loss of PZR water level and refer E. to appropriate emergency procedure (E0I-1, 2 or 3).

lv-

- A. If PZR level is stablized by additional charging pumps placed in service g per immediate actions, determine source of leakage. .

1. If determined to be via PZR PORV, isolate associate block valve and determine need for controlled shutdown.

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2. If determined to be via CVCS letdown system, isolate and determine requirements for repair and need for controlled shutdown of unit.

A

3. If determined to be via S/G tube leak see AOI-24.

4. If determined to be via RCP thermal barrier, verify isolated and I -. determine need for controlled shutdown.

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5. If determined to be via excess letdoun Ex, isolate from RCS and g, determine need for controlled shutdoun.

- 6. If determined to be via reactor vessel head seal leak off, isolate and i- determine need for controlled shutdown.

_, CAUTION: If,afterpeformanceofstbpsA1,2,3,4,5and6theleak has not been identified and isolated', and it is apparent to

t. you that the leak rate is greater than Tech Spec limits without reliance on SI-137 and an auto. trip signal will not be generated, trip the reactor and proceed to cold shutdown.

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B. Verify appropriate automatic actions (dependent upon point and amount of

.. leakage).

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SQNP

- . AOI Units 1 and 2 Page 4 of 4 f, Rev. 6 C. Determine the amount of leakage. Initiate performance of SI-137 " Reactor Coolant System Leak Rate Test." While the leak rate test is being con-E' ducted, refer to steps IV.D and IV.E of this instruction.

If not determined in step A above, determine following:

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1. Location of the leak, if possible.

2. Wheter it is pressure boundary leakage

3. How the leak will affect continued operation EL- 4. If the problem can be corrected (maintenance), or isolated

5. Likelihood of leak to propagate. . . y-

8. D. In the event the leak is into the containment building: (1) Place spare (2) Monitor containment upper and lower compartment coolers in service.

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pressure (pressure relief valves are isolated).

L. E. See Technical specification 3.4.6.2. for limits and actions on leakage.

F. In the event the leak propagates, refer to E0I-1, " Loss of Reactor Coolant."

CAUTION: Should leak be determined to be in the Auxiliary Building (i.e. CVCS or RHR system) take corrective action to isolate and L Onplement appropriate radiation protective measures.

V. DISCUSSION I_ This instruction provides symptoms, automatic actions, and operator actions for primary plant leakage that allows the pressurizer level to be maintained using one charging pu=p and does not cause an increase in containment pressure to the e- . point of safety injection actuation.

L

, Primary plant leakage is monitored on a continuous basis by radiation I monitors,

, humidity detectors, and other indicators. In addition, leakage will be calculated L. on a shift basis. Therefore, gradual increases in leakage will be identified before serious problems are encountered. However, it is possible that sudden

. leaks will develop. This instruction provides assistance in identifying sudden v.

u

' leaks.

. Normal RCS leakage is .004 GPM. One gpm unidentified leakage and 10 gpm identi-

... 'fied, fro'm other than controlled leakage paths is allowed by technical L- specifications unless it is determined to be pressure boundary leakage in which case shutdown is required.

t_ For the methods of reactor coolant system leak detection, refer to FSAR, Section 5.2.7 and 16.3.4.6.1.

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, , Icak Identificat' on i Procedure - '

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3 PZR Pressure or (1)PZH level may

, level (1) decreasing Indicate high if l is lui y r y space of PZR.

Charging Flow Increasing / Decreasing Malfunction of s charging system or '

AOI-20 i

" LeaFIFV3por ~ sppce, of PZH 4 AOI-18

! Cont sinment Pressure l Increasing /No Change i .

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U/G Pronntire tai nno it'l? N D y l

! S/G compared to othern Air H,)octor Hadlallon or

,8',yjij'.j m p ,j

• l Decreasing /No Change S/U Blowdown Radiation op g 3 Increasing /No Change

l H l Steam or Feedvater line Containment Radiation '

break inside containmer L Ievel Increasing S/G Tube Icak Steam Pressure EDI-2 AOI-21: or E01-3 Decreasing /No Change or SteamPreasureLoweronk Y.. ,, one S/G .than others et l Can PZR level t i. J be T YES 'NO l l. , maintained with one y y h

l . charging pump Steam Break,or Feedwate .- Small RCS Iaak  !

YES No break outside contain- AOI-6 l s y ment E0I-2 l

( i AOI-6 ,

E0I-1 I

(

,, ', 'Sequoych Nuclear Plant DI & wm od 1 Plant Master File l

, Superintendent j 1 Assistant Superintendent (Oper.)

A? NORMAL OPERATriG INS *RUC"' ION 1 Assistant Stiperintendent (Maint.)

, Administrative Supervisor f Maintenance Supervisor (M)

! -AOI-2h Assistant Maintenance Supervisor (M)

STFJJ4 GENERA".'CR "'UBE LIAK Maintenance Supervisor (E)

Assistant Maintenance Supervisor (E)

{

tinits 1 and 2 Maintenance Supervi cr (I)

Results Supervisor l 1 Operations Supervisor Quality Assurance Supervisor Health Phyr.ies Public Safety Services Supv.

Chief Storekeeper L Preop Test Program Coordinator Outage Director Chemical Engineer g Radiochem Laboratory i Instrument Shop i Reactc,r Engineer Instrument Engineer 1.- Mechanical Engineer Staf f Industrial Engineer 1c Training Center Coordinator

-t_ PSO - Chickamauga Engrg Unit - SNP Prepared By: J.R. Walker Public Safety Services - SNP in Shift Engineer's Office 1

Revised By: m N/A Mn , r-Unit Control Room m f0 h* //40A/l) ) QA&A Rep. - SNP

/

_. i Submitted By: Health Physics Laboratory

~~#

Supervisor ~ 7tv Chief, Nuclear Generation Branch La p r./2 ey /sf o 7ti P Prod Central Office File PORC Review: - -

, it Superintendent, WBNP Date Superintendent, BFNP i

1 Superintendent, BENP 3 q / 1 EN DES - NEB NEG Approved By: [// ( , t 1 Supv., NPHPS ROB, MS 4

~ Superintendent / / 1U NRC-IE:II J-Power Security Officer, 604 PRB-C Nuclear Materials Coordinator Date Approved: 2/29 #d Manager, OP-QA&A Staff

1. 1U P Prod Plant Eng. Branch 1C NSRS, 309 GB-K IC Technical Support Center 1C Resident NRC Insnector, SNP I

Rev. No. Date Revised Pares Rev. No. Date . Revised Pares

. O g 80 All 1

l s .-

The last page of this instruction is Number 7

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_ . . . . . ... .. - . . .b.

SQNP AOI Units 1 & 2 Page 1 of 7 Rev. O STEAM GENERATOR TUBE I.EAK

.I I. SYMPTOMS l

This procedure covers tube leaks of such a size that a detectable change in charging flow is required but is limited in size such that a trip i signal is not received, and PZR level can be maintained with additional

.L charging flow. But it is large enough that you will have a flooding problem with the faulted S/G.

(, Leaks of less size than that covered by this procedure are assumed to be detected by sampling. In this size of leak the Standard Tech Spec would be used to determine the proper action and shutdown would be via q normal operations procedures.

A. The following symptoms will indicate a S/G tube leak:

. 1. Condenser vacuum pump exhaust high radiation.

l 2. Steam generator blowdown high radiation.

i B. The following parameters will also be affected by steam generator tube leak.

l. 1. Decreasing pressurizer pressure and level.

~

2. Increase in charging flow.

I C. Probable alarms .

9

1. "RA-90-119A condenser vacuum pump air exhaust monitor high I radiation".

P

2. "RA-90-120A steam generator blowdown liquid sample monitor i high radiation".

3.

" Low pressurizer pressure, backup heaters on." ($ 2210 psig)

4. " Volume control tank level high-low." (20%)

5. " Pressurizer level low heaters off and letdown secured." ($ 17%

, of span)

~~

II. AUTOMATIC ACTIONS

~

A. Steam generator blowdown effluent isolation from high radiation and divert to condensate DI inlet. ,

' B. PZR level low, heaters off and letdown secured. ,

L * * *

• = m. .

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'l SQNP i

AOI Units 1 & 2

. Page 2 of 7 Rev. 0 1 i III. INTEDIATE OPERATOR ACTION y A. Manual Actions I

, CAUTION: Verify all PZR PORV's are closed.

l 1. Place additicnal charging pumps in service and attempt to restore

. pressurizer level. )

1 i

n 2. Attempt to maintain pressurizer level normal.

t

3. If pressurizer level can be maintained near program, reduce power at 2%/ min. Announce SG tube leak on PA system and go to

%- Subsequent Actions when unit is shut down.

[ CAUTION: If loss of PZR level is imminent, trip reactor, initiate

_ safety injection and refer to EDI-3.

l IV. SUBSEQUENT OPERATOR ACTION 3- NOTE: The process variables referred to in this instruction are typically I

monitored by more than one instrumentation channel. The redundant channels should be checked for consistency while performing the .

y_ steps of this instruction.

NOTE: The pressurizer water level ind; cation should always be used in con-junction with other specified reactor coolant systems indications to evaluate system response and to initiate manual operator actions.

, A. Notify shift engineer to determine gas release condition and notify Health

'._ Physics to survey the turbine building.

B. Verify that all PZR PORV's are closed.

C. Stablize tte RCS at approximately no-load temperature by steam dump to the main condenser. ,

l

.- D. Start motor driven auxiliary feedwater pumps and regulate the auxiliary feedwater flow to the S/G's to maintain S/G water level in the narrow I

t range span (E 33%).

E. Identify the faulted S/G by one or more of the following methods:

. 1. An unexpected rise in one S/G water level with feedwater flow less than other S/G's feedwater flow.

2. High radiation from any one S/G blowdown line.

3. Survey the individual S/G blowdown line with a portable monitor.

,. 4. High radiation from any one S/G, as determined by analysis of a sample. '

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• SQNP AOI Units 1 & 2 Page 3 of 7

[ Rev. O t

IV. SUBSEQUENT OPERATOR ACTION (Cont.)

g F. When the faulted S/G has been positively identified, then:

L.

1. Stop all feedwater flow to te. faulted S/G.

1. Close the MSIV and bypass valve associated with the faulted S/G.

3. Verify the closure of the SG PORV associated with the faulted S/G.

f.'

4. Ensure the steam driven auxiliary feedwater pump is not being supplied with steam from the faulted S/G.

CAUTION: Monitor containment indications to verify that a loss of reactor coolant other than the S/G tube leak is not in progress.

I g- NOTE: With faulted S/G isolated with RCS temperature of 547'F, the faulted S/G steam pressure will be E 1000 psig.

l_ CAUTION: DO NOT proceed to step G until the faulted S/G has been 4

identified and isolated.

I

( G. After the faulted S/G has been identified and isolated, begin a rapid j cooldown of the reactor coolant system to 50*F below the no-load temp-erature by use of the steam dump, b8

1. If offsite power and the condenser are available, dump steam to the main condenser from the intact S/Gs by manual control of the steam heauer pressure controller.

I

2. If offsite power is not available or the main condenser is not i

available, dump steam from the intact S/Gs through the S/Gs PORVs. I f H. After the reactor coolant system temperature has been reduced to 50*F i below the no-load temperature, being a depressurization of the RCS to a value equal to the faulted steam generator steam pressure.

i

NOTE

With RCS temperature at 497'F, 50*F subcooling will be maintained j down to E 1000 psig on the RCS.

NOTE: During subsequent controlled reactor coolant system depres-surization, the reactor coolant system pressure criteria for tripping the reactor coolant pumps DOES NOT APPLY.

t If the RCPs are in service, use the PZR spray to reduce _the pressure.

g If the RCPs are not in service, open one PZR PORV as necessary to de-crease pressure.

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SQNP AOI Units 1 & 2

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(, Rev. 0

.~ IV. _SURSEQUENT OPERATOR ACTION (Cont.)

When the reactor coolant system pressure is reduced to 1500 psig, iso-g I.

,\ late the upper head injection system as follows or verify isolated:

. 1. Close FCV-87-21 and GAG

[ 2. Close FCV-87-22 and GAG

.' 3. Close FCV-87-23 and GAG

4. Close FCV-87-24 and GAG J. Isslate the cold leg accumulators by closing the following valves when Rf.S pessure drops below 1000 psig if contents of accumulator has not been dumped to RCS.

.- NOTE: Power will have to be placed on these valves.

g 1. FCV-63-118 (480V Rx HOV Bd 1Al-A)

2. FCV-63-98 (480V Rx MOV Bd IB1-B) *

3. FCV-63-80 (480V Rx MOV Bd 1Al-A)

4. FCV-63-67 (480V Rx MOV Bd 1B1-3)

L CAUTION: Monitor containment indications to verify that a loss of reactor coolant other than the S/G tube leak is not in g- progress. If recirculation sump level or a containment sample are not in the normal pre-event range, further

,, ' accident recovery must be directed according to E0I-1.

C K. As the reactor coolant system pressure decreases, due to the quenching of the steam by the PZR spray or due to the steam release thru the PZR t PORV, monitor the PZR water level indications and stop the depres-L surization operation.

, 1. If the indicated water level in the PZR rises above 507. of span L.' OR l' 2. As soon as the RCS pressure decreases to a value equal to the

[f faulted S/G steam pressure.

t L. After the depressurization operation has been verified to have been

• ~

terminated, continue to monitor the reactor coolant system pressure and the PZR water level.

1. If the PZR water level continues to rise or remains nearly constant

.- concurrent with a RCS pressure decrease, suspect leakage from the PZR steam space. Monitor the PZR relief tek (PRT) pressure temp-erature and levels to identify continuously increasing conditions.

Close the PORV isolation valves if a reactor coolant leak to the PRT is identified. Monitor PRT conditions to verify PRT integrity.

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L s SQNP AOI Units 1 & 2 Page 5 of 7 i Rev. O

~

IV. SUBSEOUENT OPERATOR ACTION (Cont.)

t CAUTION: If PRT integrity is lost, abnormal containment conditions could exist and may not be true indications of a con-tinued loss of reactor coolant. If condition of step

'6 L.1 persists after closing the PZR PORV isolation valves, further recovery must be directed according to E0I-1. The

,. condition of step L.2 (this procedure) must be satisfied i a, before proceeding to step M (this procedure).

i

. 2 If the PZR water level subsequently continues to increase con-l current with a reactor coolant. system pressure increase concurrent 1  % with verified PRT integrity, the charging flow is greater than the leak.

t q., THEN, when RCS pressure has increased by at least 200 psi (after shutting the spray valve or verified closure of the PZR PORV) and

. an indicated water level is present in the PZR, stop all operating charging pumps not needed for normal charging and reactor coolant pump seal injection flows.

l' i _.

NOII: Following termination of charging pumps, PZR pressure should decrease to a value equal to the faulted S/G steam pressure,

,~

M. Re-establish charging and letdown flows to maintain the PZR water level

. in the operating range, (E25%). (If previously isolated).

CAUTION: If, during subsequent recovery actions, PZR water level can-

-- not be maintained above 25% indicated level, start additional charging pumps to re-establish PZR water level in the op-erating range.

1. Close seal injection water flow control valve FCV-62-89.

2. Open the charging pump normal suction valve FCV-62-132 and

FCV-62-133 from the VCT.

3. Close the charging pump suction valves FCV-62-135 and FCV-62-136

from the refueling water storage tank.

4. Open the centrifugal charging pumps miniflow isolation valves i FCV-62-98 and FCV-62-99.

^

5. Open the charging line isolation valves FCV-62-90 and FCV-62-91.

.- 6. Open seal water heat exchanger inlet isolation valves FCV-62-61

.. and FCV-62-63. I ,

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• AOI Units 1 & 2 Page 6 of 7 Rev. 0

{

IV. SUBSEQUENT OPERATOR ACTION (Cont.)

k 7. Gradually open the seal injection water flow control valve FCV-62-89.

Adjust the seal water flow to 8 gpa per RCP.

IL- 8. Open letdown isolation valves FCV-62-69 and FCV-62-70.

9. Open the letdown line isolation valve FCV-62-77.

Ic 10. Open the 45 gpm letdown orifice isolation valve ECV-62-73.

11. Position FCV-62-81 to control pressure at letdown orifices above

k. steam flash point.

. 12. Close the BIT inlet isolation velves FCV-63-39 and FCV-63-40 g+ and outlet isolation valves FCV-62-25 and FCV-63-26.

N. Re-establish the use of the PZR heaters to maintain the RCS pressure.

l O. If all RCP's are running, trip all but one RCP so as to maintain one pump operating in the loop connected to the PZR (loop 2), or if this is in the faulted loop, in Icep 1.

P. If offsite power is available, begin a controlled cooldown of the RCS

at a rate of about 50*F/hr by use of the steam dump to the main con-ls densair from the non-faulted S/Gs. Control the water levels in the I,' S/G's to maintain S/G water level in the narrow range span.

If offsita power is not available, dump steam from the non-faulted

, S/G's through the S/G's PORVs to provide a controlled cooldown of the reactor coolant system at.a rate of about 50*F/hr.

Q. Simultaneous with the cooldown using the non-faulted S/G, slowly de-crease the faulted S/G's pressure by opening the bypass valve around the MSIV to the condenser (if available), or using the S/G's PORV.

..- CAUTION: Request RCS boron sample and determine shutdown margin within first hour of cooldown, and adjust boron concen-

.tration as needed to maintain shutdown margin throughout cooldown.

R. As pressure is reduced in the faulted S/G, control the RCS pressure at I a value approximately equal to the steam pressure in the faulted S/G  ;

to minimize the leakage flow. RCS pre =sure control should be accom-plished by use of the PZR heaters tcf utiJn of one of the following:

.i 1. Normal.PZR spray (if a RCP is in service)

OR

2. Use of PZR auxiiIary spray (if spray is heated by letdown through the reg. Ex.)

OR

3. Briefintermittanto%ningofonePZRPORV. ,

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Page 7 of 7 g Rev. 0

4 IV.

,t. SUBSEQUENT OPERATOR ACTION (Cont.)

. NOTE

Maintain RCS temperatur'e and pressure within the limits j of the normal cooldown curves.

t-

• CAUTION: If RCS pressure control is accomplished by use of the

(_,

PZR FORV, continuously monitor the PRT pressure, temp-erature and water level and take appropriate actions to verify and maintain PRT integrity. Verify PZR PORV closure when not in use and if a reactor coolant leah

aL, to the PRT is identified, close the PORV block valve.

. S. Continue to cooldown and depressurize the RCS and faulted S/G until i

the reactor coolant hot leg temperatures are below 400*F in the non-faulted loops and the RCS pressure has reached about 380 psig (D0 NOT collapse the PZR steam bubble).

g_, T. Place the RER system in service using normal cooldown procedure.

NOTE: Throughout this cooldown procedure, maintain a steam bubble in the PZR. Solid water pressure control may not be effective.

U. Continue the plant cooldown in a normal mode except that after the RCP

, I'

• operation has been terminated, continue to simultaneously control the

! a_. faulted S/G steam pressure and reactor coolant pressure to minimf re the leakage flow. (Cooldown per appropriate sections of GOI-3C) .

t I

V. When the RCS system hot leg temperatures are reduced below 200'F, the I

pressure in the PZR may be reduced by using auxiliary spray until re-actor coolant system pressure and the faulted S/G pressure equilibrate.

m-

W. Continue the operation of the RHR system to remove the core residual i

heat and maintain the charging and letdown in service to control the i PZR water level and provide a boration path,

!s. NOTE: Secondary plant should be monitored for radiation release from various systems:

+*

1. Cond vac. pump exhaust i

2. Turbine driven AFWP exhaust 1

3. S/G PORV J..

m

4. Gland steam exhauster

5. T.B. station sump V. RECOVERY ,

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l Following a S/G tube leak, the exact procedure will be pla,nned by the Plant  !

Operations Review Committee for repairing the affected tube or tubes and decontamination of the secondary system.

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