Text

Proposed Tech Specs,Reflecting Effects of Boron Injection Tank Deactivation
	ML20006A273
Person / Time
Site:	Sequoyah
Issue date:	01/12/1990
From:	; TENNESSEE VALLEY AUTHORITY
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L

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jf[ll ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANG?.

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2-DOCKET NOS. 50-327 AND 50-328 3

(TVA-SQN-TS-89-26) 1:

[

LIST OF AFFECTED PAGES Unit 1 p

-VII y

3/4 1-11 3/4 1-12 3/4 5-1 3/4 5-7 3/4 5-11 3/4 5-12 3/4 5-13

~

B 3/4 1-3 n

B 3/4 5-2

.I 1'

B 3/4 5-3' L

l h

Unit 2 l'

VII I

3/4 1-11' l:

3/4 1-12 l

3/4.5-1 3/4 5-7 3/4 5-11 L.

3/4 5-12 3/4 5-13 l-

'B 3/4 1-3 B 3/4 5-2 B 3/4 5-3 18 p

i l -.

L l.

P 1

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INDEX 3

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS Cold Leg Injection Accumulators...........................

3/4 5-1 r

g Q :r " d ?r.j;; tie. Au umul m,b........................

3/4 5 3

\\

D5 N#3/4.5.2 ECCS SUBSYSTEMS - Tavg greater than or equal to 350 F......

3/4 5-5 was-submitled 3/4.5.3 ECCS SUBSYSTEMS - T less than 350 F....................

3/4 5-9 in-T5 avg g

3/4.5.4 BOR0t! Itu:CTION SYSTCh-DELare o

! M ~ge, D'

-Ceren Inj;; tion Ten's.-

S/4 5.t f _ _1 T__m 1.,.

.m

.....................................~.........

-3/0 5-12 3/4.5.5 REFUELING WATER STORAGE TANK..............................

3/4 5-13 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.....................................

3/4 6-1 Containment Leakage.......................................

3/4 6-2 Containment Air Locks.....................................

3/4 6-7 Internal Pressure.........................................

3/4 6 Air Temperature...........................................

3/4 6-10 Containment Vessel Structural Integrity...................

3/4 6-11 Shield Building Structural Integrity......................

3/4 6-12 Emergency Gas Treatment System (Cleanup Subsystem)........

3/4 6-13 l

Containment Ventilation System............................

3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System..................................

3/4 6-16 R73 l

Lower Containment Vent Coo 1ers............................

3/4 6-16b 9

l R120 L

. SEQUOYAH - UNIT 1 VII Amendment No. 67,63 116 June 1, 1989 1

l.

f.7

.z.

o.

REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION i

s l

3.1.2.5 'As a minimum, one of the following borated water sources shall be-OPERA 8LE:

A boric acid storage system and associated heat tracing with:

I a.

1.

A minimum contained borated water volume of 2175 gallons,

-I 2.

Between 20,000 and 22,500 ppa of boron, and i

j'u

(

3.

A minimum solution temperature of 145'F.

I 1

- b.

The refueling water storage tank with:

p.,

i 1.

A minimum contained borated water v of 35,443' gallons, j

3 2.

A minimum boron concentration of ppa, and-4y 3.

A minimum solution temperature of 60*F.

1

-'V APPLICABILITY: MODES 5 and 6.

j

')

ACTION:

l 4

-With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS ll 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

. (

-+

2

.a..

At least once per 7 days by:

[

i 1.

Verifying.the boron concentration of the water, 7

2.

Verifying the contained borated water volume, and C

l 3.

Verifying the boric acid storage tank solution temperature when d

it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water.

h Nl." -

SEQUOYAH - UNIT 1 3/4 1-11 i

e

,s S.

i REACTIVITY CONTROL SYSTEMS I

B0 RATED WATER' SOURCES OPERATING i

LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum,-the following borated water source (s) shall be OPERABLE as required by Specification'3.1.2.2:

a.

A boric acid storage system and associated heat with:

1. -

A minimum contained borated water volume'of

allons, 2.

Between 20,000 and 22,500 ppm of boron, and 3.

.A minimum solution temperature of 145*F.

' b.

The refueling water storage tank with-t 1.

A ' contained borated water volume of between 370,000 and 375,000

(-

gallons 2.

Between nd pm of boron, 3.

A minimum solution temperature of 60'F, and

)-i 4

A maximum solution temperature of 105'F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:-

~

a ' With the boric acid storage system inoperable and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200'F; restore the boric acid storage system tn OPERABLE. status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

'With the refueling' water storage tank inoperable, restore the tank b,

to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(

bo.-

SEQUOYAH - UNIT 1 3/4 1-12

'E

'~

_. =-

L o'

i+

p-3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS

~ 7- ~~-

COLD LEG INJECTION ACCUMULATORS

'i LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:

The isolation valve open, a.

jf7 b.

A contained borated water olume of between and gallons of f

boratort wat er

, VHI re.meval, and W

( 00 j66ficafian has c.

Between and ppm of boron, and 1

l

^ been ' avbmWal PS ge d.

A nitrogen cover pressure of between and psig.

'61-1 0 APPLICABILITY: MODES 1, 2 and 3.8 ACTION:

.With one cold leg injection accumulator inoperable, except,as a result a.

of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STAN0BY within i

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in H0T SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

Withonecoldleginjectionaccumulatorinoperableduetotheisola-tion valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.#

With one pressure or water level channel inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in H0T R128=

SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , With more than one channel (pressure or water level) inoperable per d.

accumulator, immediately declare the af fected accumulator (s) inoperable.

Pressurizer pressure above 1000 psig.

Actions c and d are in effect until the restart of Unit 2 from the Unit 2 Cycle 4 refueling outage.

R128

.)

SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 124 August it. 1989

.t.

I p, y ", 4.

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.1 EMERGENCY CORE COOLING SYSTEMS (ECCS)'

L.m u4 SURVEILLANCE REQUIREMENTS (Continued) i h

... a.w:n. -.? :.

2.

Verifying that each of the following pumps start automatically

_,upop,r3peipt of a safety injection signal, a)'. Centrifugal charging-pump 1

L

' b).

Safety injection pump.

c)

Residual heat removal pump L

f.

-By verifying-that each of the following pumps develops the indicated j

E discharge pressure on recirculation flow when tested pursuant to J

Specification'4.0.5: '

J 1.

Centrifugal charging pump dreaterthanorequalto2400psig

~

2.

Safety Injection pump Greater than or equal to 1407 psig j

3.

Residual heat removal. pump Greater than or equal to 165 psig.

2

g..

By verifying the' correct position of each mechanical stop for the C

following Emergency Core Cooling System throttle valves:

1.

Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s-following completion of each valve stroking operation'or maintenance on the valve when the ECCS subsystems y

j are required to be OPERABLE.

l

[

L 2.

At least once'per 18-months.

4 cx, Injection Safety. Injection Cold Safety Injection Hot.

t p

Throttle Valves Leg Throttle Valves Leo Throttle Valves

. Valve Number Valve Number

~ Valve Number.

l is

1. 63 - 582

1. 63 - 550

1.63-542

2. 63 - 583

2. 63 - 552

2.63-544 j.

3. 63 - 584

3. 63 - 554

3.63-546

4. 63 - 585

4. 63 - 556

4.63-548 1

l f

L4 SEQUOYAH - UNI'T 1 3/4 5-7 n'

l+

1

'a.

1 1

EMERGENCY CORE' COOLING SYSTEMS (ECCS)

)

3/4.5.4_ 00:0W ;4:C7:0:; OY;T;;; DELETED g

g

]

(BORONINJECTION LIMITIN NDITION FOR OPE ION j

1 The bor injection tank s be OPERABLE wit -

4.

A nimum contained ated water vol f 900 gallons, b

Between 20,000 d 22,500 ppm of b on, and

)

L c.

A minimum olution temperatu of 145'F.

1 APPLICABILITY

MODES 1, 2 and 3.

ACTION:

t Wi theboroninject tank inoperable, r tore the tank to ABLE status thin.I hour.or b n HOT STANDBY and ated to a SHUTD0 RGIN equivalent

.to 1% delta k/k f 200'F within the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restor e tank to OPERA 8

-status withi he next 7 days or in HOT SHUTDOWN w in the next 12 ho 1

., Y SURVEILLANCE RE EMENTS 4.5.4

- The.boroninjec n tank shall be d nstrated OPERA by:

a..

Verifyin he contained bora water volume east once per 7

days, b,

erifying the boron ncentration of water in the tan t least once per 7 days,

.d c.

Verifying t water temperat at least once per hours.

V SEQUOYAH - UNIT 1 3/4 5-11

. m m

m

l EMERGENCY CORE COOLING SYSTEMS (ECCS)

Oe leb b'

HEAT TRACING

~

LIMIT CONDITION FOR OP ATION 3.5.4.2 At lea two independent chan s of heat tracing all be OPERABLE for the boro njection tank and fo he heat traced por ons of the associ

!d flow paths AP,PLI ILITY: MODES 1, 2 3.

N:

With only one cha el of heat tracing either the boro injectiontankor the heat trace ortion of an asso ted flow path OP BLE, operation ma continue for p to 30 oays provi d the tank and f path temperature are

. verified-be greater than or qual to 145'F a east once per 8 h s;

otherwi

,-be in at least H STANDBY within hours and in HOT DOWN with the.following 6 h rs.

%, 2 SURVElk ANCE REQUIREME

~,

4.2 Eac eat tracing chan for the boron inj ion tank and assoc ed flow path all be demonstra OPERABLE:

At least once er 31 days by ener ing each heat trac

channel, and b.

At I t once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months y verifying the ta and flow path t

eratures to be gre r than or equal to 5'F.

The tank emperature shall b etermined by measu nt.

The flow p temperature shal e determined by ei r measurement or circula-tion flow unt establishment of e librium temperat s within the tank.

Oc

,A SEQUOYAH - UNIT I 3/4 5-12

'N j

l

2

' EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION s

y 3.5.5-The refueling water storage tank (RWST) shall be OPERABLE with; i,

A contained borated water volume of between 370,000 and 375,000

- a.

gallons, b.

A boron concentration of between and ppe of boron.

l A minimum colution temperature of 60'F, and K16 c.

d.

A maximum solution temperature of 105'F.

APPLICABILITY: MODES 1, 2, 3 cnd 4.

ACTION:

b With the RWST ' inoperable, restore the tank to OPERABLE status within i hour or I'

be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within'the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , rN q

L 1

h SURVEILLANCE REQUIREMENTS

{

4.5.5-The RWST shall-be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

-Verifying the contained borated water volume in the tank, and 2.

Verifying the boron concentration of the water.

b.

At-least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying t'he RWS1 temperature.

i;

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gs -

.4

%=

MAR 251582

-k--

SEQUOYAH - UNIT I 3/4 5-13 Amendment No. 12

F b'

n REACTIVITY CONTROL SYSTEMS BASES gallons of 20,000 ppm borated water from the boric acid storage tanks or

" '00 gallons of m borated water from the refueling water storage tank.

82,082 With the RCS temper'ature below 200'F, one injection system is acceptable i

without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200'F, is sufficient to provide a EHUT00WN MARGIN of 1% delta k/k after xenon decay and cooldown'from 200*F to 140 F.- This condition requires either 635 gallons of 20,000 ppm borated water

-from the boric acid storage tanks or 9,690 gallons of ppm borated water from the refueling water storage tank.

ESoC)

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

(

The limits on contained water volume and boron concentration of the RWST.

BRl also ensure a pH value of between 7,5 and 9.5 for the solution recirculated I

within containment after a LOCA.

This pH band minimi.zes the evolution of

. iodine and minimizes the effect of chloride and caust'ic stress corrosion on mechanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications.of this section ensure that (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUT 00WN MARGIN is maintained, and (3)' limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment' and insertion limits.

L SEQUOYAH - UNIT 1 B 3/4 1-3 R'evised 08/18/87 Bases Change

~

I l

(.--

EMERGENCY ~ CORE COOLING SYSTEMS

. BASES s

With the RCS temperature below 350*F, one OPERA 8LE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensurs OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses.

are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration,.

(2) provide the proper flow split between injection points in accordance'with the assumptions-used in the ECCS-LOCA analyses, and (3) provide an acce' table-p level of total'ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

3/4.5.4 BORON INJECTION SYSTEM Ql&

. The OPERABILIT the boron injec n system as part the ECCS ensur

' that sufficient ative reactivity ' injected into th ore to countera any positiv crease in reacti caused by RCS s em cooldown. R cooldown can be sed by inadverten pressurization; oss-of-coolant dent or a ste ine rupture.

The limits njection tank si um contained vol and boron conc ra-tion ensure the assumptions ed in the steam e break analysis e

met. Th ntained water.vo limit includes allowance for w not usua ecause of tank.

harge line loca or other physic characteristics The-0PERABIL of the redundan eat tracing chann associated with the boron in on system ensur at the solubilit the boron-solut n will be m alned above the ubility limit of F at 21000 ppm on.

8 SEQUOYAH - UNIT 1 B 3/4 5-2

...o l

L EMERGENCY' CORE COOLING SYSTEMS BASES 3/4.5.5 REFUELING WATER STORAGE TANK

-The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits on RWST minimum volume and boron concentration ensure that-

1) sufficient. water.is available within containment to permit recirculation cooling flow to the. core,-and 2) the reactor will remain subcritical in the- _

t cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST-f also ensure a pH value of between 7.5 and 9.5 for the solution recirculated g'

within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes-the effect of chloride and caustic stress corrosion on mechanical systems and components.

(

iAdd Addi+ionoI19 the ~OPERABIUTY of th e RW3T' as pari efthe Ecc5 e.nsures ihai sol %cien+ negaHve reacHvHg is

~

injec+ed ink the sore h counterac+ any posHive ir1 crease ir) reac+ivHy caused by R66 cy6km

]

Cooldown.

1 SEQUOYAH - UNIT I B 3/4 5-3 Revised 08/18/87

.h-

,1

~INDEX-(j LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1

L

.W c

SECTION PAGE l

3/4.5 EMERGENCY CORE COOLING SYSTEMS t

L 3/4. 5.1 ACCUMULATORS Cold Leg Injection Accumulators...........................

3/4 5 ??:7 " :d Inj;;ti n ^.;;;;;N t0 %........................

-3/4 Ph

,g L This cheTc /4.5.2 ECCS SUBSYSTEMS - T greater than or equal to 350 F.....

3/4 5-5 3

, web avg y

sam ed 3/4.5.3 ECCS SUBSYSTEMS - T less than 350*F....................

3/4 5-9 Jn TS avg c.hong C 3/4.5.4 - "0"07: It05CT!0f? SYSTEP. O ELETE O l 8't -2.5.

90 ca I a j a ' H e - T ; ; ';......................................

3/4 ', 11 t

l

"; ; t i n ; i n i;..............................................

O/4 i,12 3/4.5.5 REFUELING WATER STORAGE TANK........................'......

3/4 5-13 L

3/4.6 CONTAINMENT SYSTEMS i

3/4.6,1 PRIMARY CONTAINMENT Containment Integrity.....................................

3/4 6-1 I

Containment Leakage.......................................

3/4 6-2 Containment Air Locks.....................................

3/4 6-7 f

Internal Pressure.........................................

3/4 6-9 1.

I Air Temperature...........................................

3/4 6-10 1:

Containment Vessel Structural Integrity...................

3/4 6-11 Shield Building Structural Integrity......................

3/4 6-12 Emergency Gas Treatment System (Cleanup Subsystem)........

3/4 6-13 Containment Ventilation System............................

3/4 6-15 l

l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 1'

i Containment Spray System..................................

3/4 6-16 Lowe r Co nta i nme nt Ve n t Coo l e rs............................

3/4 6-16b R61 April 4, 1988 SEQUOYAH - UNIT 2 VII AmendmentNo.//,61

i y

. REACTIVITY CONTROL' SYSTEMS BORATED WATER SOURCE - SHUTOOWN LIMITING CONDITION FOR OPERATION l.

\\

N, l.

3.1.2.5 As.a minimum,-one of the following borated water sources shall be 0PERABLE:.

a.

A boric acid storage system and at least one associated heat tracing system with:

I 1.

'A minimum contained borated water volume of 2175 gallons, 2.

Between 20,000 and 22,500 ppm of boron, and l

3.

A-minimum solution temperature of 145 F.

b.

.The refueling water storage tank with:

I 1.

,A minimum contained borated water of 35,443 gallons, 1

son l'

2.

A minimum boron concentration of ppm, and

'D 3.

A minimum sclution temperature of 60'F.

d APPLICA81LITY:: MODES 5 and 6.

ACTION:

With no' borated water source OPERABLE, suspend all operations involving CORE

~

ALTERATIONS or positive reactivity changes.

1 L

SURVEILLANCE-REQUIREMENTS l'

L 4.1.2.5 The above required borated water source shall be demonstrated l'

-OPERABLE:

p l

a.

At least once per 7 days by:

J L

1.

Verifying the boron concentration of the water, 2.

Verifying t'he contained borated water volume, and l-3.

Verifying the boric acid storage tank solution temperature when i-

'it'is the source of borated water.

.l b.

'At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it' l^

is the source of borated water.

V

(,r SEQUOYAH UNIT 2 3/4 1-11 L

1 i

1 g,

e m3 3

. REACTIVITYCONTROLSYSTE!g BORATED WATER SOURCES - OPERATING J' %

LIMITING CONDITION.FOR OPERATION A

3.1.2.6 As a minimum, the following borated water source (s) shall b'e OPERABLE' A

as required by Specification 3.1.2.2:

q A boric acid storage system a,nd at least'one associated heat tracing a.

system with:

1.

A sinimum contained borated water volume'o

gallons, n.

-\\

2.

Between 20,000 and 22,500 ppm of boron, and i

.j 3.

A minimum solution temperature of'145'F.

~

b.

The refueling water storage tank with:

1.

A contained borated' water volume of between 370,000 and U

375,000 s,M

[

2.

Between and-i!+00 ppm of boron, and e

3.

.A minimum solution. temperature of 60*F.

4.

A maximum solution temperature of 105 F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

W.

+

'a,

.With the boric acid storage system inoperable and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTOOWN MARGIN equivalent to at least 1% delta k/k at 200'F; restore the boric acid storage system to OPERABLE status.within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t M

SEQUOYAH - UNIT 2 3/4 1-12 a

4 C '

3/4.5.1 3/4.5 EMERGENCY CORE COOLING SYSTEMS ACCUMULATORS COLD LEG INJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1.1-Each cold leg injection accumulator shall be OPERABLE witn

a.

The isolation valve open, J,.7" hb.

A contained borated water volume of between ar.d allons of u H a, r emova l, eng borated water, hsWAcati n has c.

Between and ppm of boron a nri been submiHe4 by E *nte oms.

[1.

A nitrogen cover pressure of between and p

APPLICABILITY:

MODES 1, 2 and 3.*

ACTION:

Withonecoldleginjectionaccumulatorinoperable,exceptasa a.

result of a closed isolation valve, restore the inoperable

(

accumulator to OPERABLE status within one hour or be in at least HOT l

STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l b.

With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation

.i

- valve or be in H0T STANDBY within one hour and be in HOT SHUTDOWN l

within-the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , c.# With one pressure or water level channel inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT R113 SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d.# With more than one channel (pressure or water level) inoperable per accumulator, immediately declare the affected accumulator (s) inoperable.

" Pressurizer pressure above 1000 psig.

I

Actions c and d are in effect until the restart of Unit 2 from the Unit 2 Rll3 Cycle 4 refueling outage.

SEQUOYAH - UNIT 2 3/4 5-1 Amendment No. 113 August 11, 1989

~

g, J

s

(

EMERGENCV CORE COOLING SYSTEMS

. 5URVEILLANCE RE0VIREMENTS (Continued)

~

. - =

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection signal:

a)

Centrifugal charging pump b)

Safety injection pump c)

Residual heat removal pump I

f.

By verifying that each of the following pumps develops the indicated discharge pressure-on recirculation flow when tested pursuant to Specification 4.0.5:

j 1.

Centrifugal charging pump Greater than or equal to 2400 psig 2,

Safety Injection pump Greater than or equal to 1407 psig 3.

Residual heat removal pump Greater than or equal to 165 psig

- 1 U

g.-

By verifying the correct position of each mechanical stop, for the

.following ECCS throttle valves:

1..

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking

. operation or maintenance on the valve wnen the ECCS subsysteins are required to be OPERABLE.

2.

At least:once per 18 months.

Chargin3 Ce,... Injection Safety Injection Cold Safety. Injection Hot Throttle Valves Leo Throttle Valves Leo Throttle Valves--

pornp Valve Number Valve Number Valve Number

1. 63 - 582 1, 63 - 550

1.63-542

2. 63 - 583

2. 63 - 552

2.63-544

3. 63 - 584

3. 63 - 554-

3.63-546.

4. 63

.585

4. 63 - 556

4.63-548 4

I h

SEQUOYAH - UNIT 2 3/4 5-7 9

'f:

- m - e mm-

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3

.]

i 1

EMERGENCY CORE COOLING SYSTEMS

,j J.

3/4. 5. 4 M D E L E TE D -

b6lOkC BORON INJECTION TANK-l LIMITING COND ON FOR OPERATION 1

1 3.5

.1 The boron inject tank shall be OPER with:-

a.

A minimum ntained borated wate olume of 900 gallo j

b.

'Ab n concentration of b een 20,000 and 22, ppm, and c.

minimum solution to erature of 145'F.

.APPEICABILITY:. MODES 1 nd 3.

CTION With the boron njection tank inop able, restore the tank o OPERABLE status j,

within I he or be in HOT-STAN and borated to a SHU WN MARGIN equivalent K

to 1% de k/k at 200*F wit n the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; re ore the tank'to OPERAB statu ithin the next 7 s or be in HOT SHUTO within the next 12 ho s.

1 1;

l l.'

SURVEILL CE REQUIREMENTS Y

/

5.4.1 The boron jection tank shall demonstrated OPER by:

a.

Ve ying the contained rated water volume least once:per

-)

l days,.

Verifying the ron concentration the water-in the ta at least s

once per 7 ys, and J

a,

[

c.

Ver ing the water temp ture at least once p 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

L. +

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o l

SEQUOYAH - UNIT 2 3/4 5-11 L

l l

l N -- -

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..m L

l C

EMERGENCY CORE COOLING SYSTEMS glgtg 9

HEAT TRACING (IMii:HG 00MDI. N FOR OPERATICH 1

1 1

3.$.)

At least two inc ncent channels of h tracing shall be RABLE the boron injec* ion ank and for the het raced portions of associ-ted flow paths.

APPL!CABILITY:

DES 1, 2 and 3.

ACTION:

With ly one channel of t tracing on eithe he boron injectio ank or on l

t heat traced portio f an associated f1 path OPERABLE, op tion may l

ontinue for up to 3 cays provided the

k and flow path t eratures are l

verifieo to dw gr-er than or ecual t 45'F at least one per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; i

otherwise, be i at least HOT STANO within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months an in HOT SHUT 00WN i

i within the f, owing 6 hoces.

l l

SURVEILLAN REQUIREMENTS 1

1 1

1

/

4 4.2 Eacn heat cing enannel for t boron injection t and associat low path shall b demonstrated OPERA a.

A east once per 31 s by energizing ea heat tracing

nnel, nd At least onc r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by veri ing the tank a flow path temperatu o be greater than r equal to 145'. The tank tempera shall be determi by measuremen The flow path tempe ure shall be dete ned by either surement or recirc ti flow until estabi ment of equillbp fum temperatures wit n the l

SEQUOYAH - UNIT 2 3/4 5-12

\\

E

o+.

i 1

EMERGENCY CORE COOLING SYSTEMS l

l {.

3/4.5.5 REFUELING WATER STORAGE TANK i

l

_ LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A contained borated water volume of between 370,000 and

[

375,000 gallons, b.

A boron concentration of between and e of buron.

R2 i

A minimum solution temperature of 60'F, and c.

d.

A maximum solution temperature of 105'F.

l-

)

APPLICABILITY: MODES 1, 2, 3 and 4.

A_CTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD S4UT00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

8 SURVE1LLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

I

a. \\At least once per 7 days by:

1.

Verifying the contained borated water volume in the tank, and 2.

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.

l i

g, Amendment 2 1

SEQUOYAH - UNIT 2 3/4 5-13 9/15/81

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REACTIVITY CONTROL SYSTEMS BASES 1

BORATION SYSTEMS (Continued) provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and cooldown to 200'F.

The maximum expected boration capabliitu reoutrement occurs at EOL from full power equilibrium xenon 604 conditions and requires 5 H 96 gallons of 20,000 ppm borated water from the boric acid storage tanks or ",100 allons off 994 ppm borated water from the refueling water storage tank. gg g

With the RCS temperature below 200'F, one injection system is acceptable without single fai'.,re consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE j

ALTERATIONS and positive reactivity chances in the event the single injection system becomes inoperable.

The boron capability required below 200*F is sufficient to provide a SHUTOOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200'F to 140*F. This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks or 9,690 gallons of4 000 ppm borated water from the refueling water storage tank.

g The contained water volume limits include allowance for water not f

available because of discharge line location and other physical

(

characteristics.

The limits on contained water volume and boron concentration of the RWST

,i also ensure a pH value of between 7.5 and 9.5 for the solution recirculated BR ll within containment after a LOCA.

This pH band minimizes the evolution of t,

iodine and minimizes the effect of chloride and caustic stress corrosion on

,j mechanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in NODE 6.

r i

3/4.3.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained (2) the minimum SHUTOOWN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

s L

SEQUOYAH - UNIT 2 B 3/4 1-3

. Revised 08/18/87 Bases Change

EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses t

are met and that subsystem OPERABILITY is maintained.

Surveillanc.e requirements for throttle valve position stops and flow balance testing provide assurance

^

I that proper ECCS flows will be maintainpd in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each i

injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, i

(2) provide the proper flow split between injection points in accordance with l

the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable l

L level of total ECCS flow to all injection points equal to or above that assumed l

l in the ECCS-LOCA analyses.

l 3/4.5.4 BORON INJECTION SYSTEM Ddh The OPE LITY of the oninjection tem as part of e ECCS ensures that suff ent negative ctivity is in ted into the e to counteract any p ive increase reactivity ca d by RCS syste ooldown.

RCS down c

e caused by i vertent depres ization, a los f-coolant accid or a

( Ni

/

eam line rupt

'j r

i The its on injecti tank minimum e ained volume an oron concen 1

tion e re that the a mptions used in e steam line bre analysis ar met he contained ter volume limi neludes an allow e for water ot ble because o ank discharge li location or oth physical ch eteristics.

The 0 ABILITY of the r undant heat traci channels as ciated with the bor injectionsyste nsure that the co.

111tgofth oron solutio will e maintained abov he solubility li p of 135 F a 1,000 pom bor l

3/4.5.5 REFUELING WATER STORAGE TANK r

E The OPERABILITY 'of the refueling water storage tank,(RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in tne event of a LOCA. The limits on RWST minimum vol-use and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the l

w ws SEQUOYAH - UNIT 2 B 3/4 5-2 J

i..1

= nen1 p

~.,,.

- - + - - -

i 1

EMERGENCY CORE COOLING SYSTEMS BASES j

i REFUELING WATER STORAGE TANK (Ccntinued) j i

RWST and the RCS water volumes with all control rods inserted except for the i

most reactive control assembly.

These assumptions are consistent with the LOCA analyses, j

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated BR within containment after a LOCA.

This pH band minimizes the evolution of

)

icdine and minimizes the effr:ct of chloride and caustic stress corrosion on mechanical systems and components.

1 Add l

AdE4ionally, 4ht OPERABIL!TV o P the Ov'sT'as part of j

l

% s ECC5 ensares +)mi sutheien4 ne ake reac+ivify j

aci on3 posdive.

is inJeded in% +he core to coon increase in readivi4 9 caused by RC.S sys+ern cooldorvn l

l

)

i l

[

SEQUOYAH - UNIT 2 B 3/4 5-3 Revised 08/18/87

F-

]

x 4

l ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PIANT UNITS 1 AND 2 1

DOCKET NOS. 50-327 AND 50-328 j

(TVA-SQN-TS-89-26)

DESCRIPTION AND JUSTIFICATION FOR BORON INJECTION TANK DEACTIVATION

'1 i

l

)

l i

I 1

1 i

i.

a l

l ENCLOSURE 2 i

Description of Channe Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to reflect the effects of the boron injection tank deactivation. The refueling water storage tank boron concentration will be changed in Limiting Condition for Operation (LCO) 3.1.2.5.

The volume of the boric acid storage system and the boron concentration of the refueling water storage tank will be i

changed in LCO 3.1.2.6.

In Surveillance Requirement 4.5.2.g.2, the j

reference to boron injection throttle valves will be changed to charging pump injection throttle valves. TSs 3/4 5.4.1 and 3/4 5.4.2 for the boron j

injection system are being deleted. LCO 3.5.1.1 will be revised with a i

new boron concentration for the cold leg injection accumulators, and LCO 3.5.5 will be revised with a new boron concentration for the refueling water storage tank.

i l

Reason for Change The boron injection tank is a component of the safety injection system whose sole function is_to provide concentrated boric acid to the reactor coolant to mitigate the consequences of postulated steamline break accidents.

In order to verify that the criteria for radiation releases are met. TSs are applied to the boron injection tank and associated equipment. Specifically, the TSs currently ensure that the boric acid concentration is maintained in excess of 20,000 parts per million (ppm),

approximately a 12 weight percent solution. Heat tracing is necessary to j

maintain the tank and associated piping at a sufficiently high temperature

'I so that the minimum concentration requirements may be met.

Furthermore, the safety-related nature of the boric acid system requires that the heating systems be redundant.

The required solubility temperature imposes a continuous load on the heeters, and the potential for low-temperature alarm actuation and heater burnout exists. Violation of the TS on concentration in the boron injection tank poses availability problems in that recovery is required within a very short time.

If the concentration is not restored within one hour, the plant must be taken to the hot' standby condition and borated to the equivalent of 1 percent delta k/k at 200 degrees Fahrenheit.

Thus, this requirement has a potentially serious impact on plant availability.

In addition, the high boric acid concentration makes recovery from a spurious safety injection signal (which results in injection of the boron t.

injection tank fluid into the reactor coolant system) time consuming and costly.

i These potential difficulties unfavorably affecting plant availability, operability, and maintainability can be drastically reduced in severity or eliminated by the boron injection tank deactivation.

l l

l 1

6

Justification for Channe The only accident analyses that are significantly affected by boron reduction, boron injection tank removal, or bypassing are the steamline break transients. These transients are affected with respect to both core integrity and mass and energy release to containment.

The following steamline break cases were considered in the core integrity analysis for SQN (1) " hypothetical" steamline break, with and without offsite power available, for the largest double-ended rupture of a steam pipe upstream of the flow restrictor (4.6 square feet); (2) " hypothetical" steamline break, with and without offsite power available, for the largest double-ended rupture of a steampipe downstream of the flow restrictor- (1.4 square feet) and (3) " credible" steamline break, with offsite power available, for the largest single failed open steam generator relief, safety, or steam dump valve.

(Both uniform and nonuniform cases were analyzedt uniform refers to an equal blowdown from all four steam i

generators; and nonuniform refers to a blowdown from only one steam generator.)

For the hypothetical breaks, the same criteria were applied as are applied in the Final Safety Analysis Report (FSAR). That is, for the most severe Condition IV break, the analyses show that the radiation releases are within the raquirements of 10 CFR 100 by demonstrating that the departure from nucleate boiling design basis is met. The steamline break dose calculations performed for the FSAR use a conservative fuel failure level of one percent, although the core analyses show that no consequential fuel failures are anticipated.

The credible steamline break analysis was performed using a new criterion whereby the plant may return to criticality but no damage may occur to the fuel. This constitutes a relaxation of the conservative internal Westinghouse Electric Corporation criterion for Class II events.

This relaxed criterion is in compliance with the criteria used by NRC, which require that releases during steamline break accidents remain within the limits set forth in 10 CFR 20.

This limit is met with a return to criticality if it is assured that there is no consequential fuel damage.

Tor SQN, the system was analyzed assuming that the boron injection tank remains installed, without heat tracing, and with the boric acid concentration reduced to zero ppm. This combination provides the most limiting case for the analyses.

The analyses for the hypothetical casen show that the departure from nucleate boiling design basis is met, and that no consequential fuel failures are anticipated. The analysis for the credible break shows a return to criticality, but the departure from nucleate boiling design basis is met and no fuel failures are predicted.

l l

,o j

s,

t The mass and energy analysis considered two cases:

(1) large or j

double-ended steamline ruptures and (2) small or split steamline

{

ruptures. The small break mass and energy calculations were proven to be the limiting case because of the higher containment temperatures reached.

Assuming the boron injection tank remains installed, without heat tracing, l

and with the boric acid concentration reduced to zero ppm, the temperatures and pressures reached in the small break calculations fall below the containment design limits.

+

i Increasing the refueling water storage tank boron concentration is proposed to address the future need (beyond Cycle 4) for a boron i

concentration increase, which was identified when the Cycle 4 reload I

safety evaluations were performed.

In fact, the Unit 2 Cycle 4 reload l

safety evaluation stipulated that the boron injection tank needed to remain in cperation during Cycle 4.

For future fuel reloads, with or without. Vantage 5 Hybrid fuel, the boron concentration needs.to be increased to accommodate the higher enrichments resulting from extending i

the fuel cycles (in the process of going from 12 to 18 months) and decreasing the number of fresh fuel assemblies (of the 193 total assemblies, instead of changing out 72 to 80 new assemblies, changing out 60 to 68).

In performing this evaluation, the strategy employed was to select the highest boron concentration possible that would accommodste the removal of the boron injection tank (approximately 55 ppm), accosanodate removal of upper head injection (approximately 45 ppm), meet the post-loss of coolant accident sump potential hydrogen-ion activity (pH) requirements specified in the FSAR and TSs and be acceptable to NRC in order to provide the maximum margin available for future fuel reloads.

The evaluations performed to support boron injection tank deactivation accommodate the' effects from the following modifications planned for the Cycle 4 outages for each unitt j

t

~

1.

Resistance temperature detector bypass elimination 2.

Eagle 21 digital protection system implementation i

L 3.-

Upper head injection remova!

b 4.

Vantage 5 Hybrid fuel impienentation

'5.

Use of new steamline break-protection 6.

Reactor trip on steam flow / feed flow mismatch elimination In summary, plant specific analyses have been performed for SQN's steamline break transients. These analyses have shown that the boron injection tank may be bypassed, eliminated, or reduced in boron concentration and the heat tracing system removed. Additionally, the analyses performed for SQN require an increase in the minimum and maximum boron concentrations for both the refueling water storage tanks and the colei leg accumulators. This increase is neesssary to mest the boron requirements in the postaccident sump. Also, to meet the increased boron 1

requirements associated with future core reloads, the volume of the boric acid storage system will increase.

1

- - - - - - ~. -

o I

4-Environmental Impact Evaluation f

r The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change.would nott 1.

Result in a significant increase in any adverse environnental. impact previously evaluated in the Final Environmental Statement (FES) as i

modified by the Staff's testimony to the Atomic Safety and Licensing l

Board, supplements to the FES. environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2.

Result in a significant change in effluents or power levels.

3.

Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.

t V

i 1

i l

C

.r---

...w.

_-c.- -.-m

-r--

i 4

.a ENCLOSURE 3 l

PROPOSED TECENICAL SPECIFICATION CRANGE t

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-26)

DETERMINATION OF NO SIGNIFICANT RAZARDS CONSIDERATIONS I

(

i b

9 e

(,-'

l l

n i

=-

s

-i l

ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the proposed TS change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will nots (1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

i The deact1vation of the boron injection tank affects the steamline i

break transients with respect to core integrity and mass and energy release to containment. With the assumption that the boron injection tank remains installed without heat tracing.and with boric acid concentration reduced to zero ppm, analyses show that the departure from nucleate boiling design basis is met and no consequential fuel j

failures are anticipated. Additionally, temperatures and pressures reached in containment would fall below the containment design limits. Therefore, no significant increase in the probability or consequences of a previously analyzed accident would occur.

(2) Create the possibility of a new or different kind of accident from

)

any previously analyzed.

The boron injection tank is a component of the safety injection system whose sole function is to provide concentrated boric acid to the reactor coolant to mitigate the consequences of postulated steamline break analysis. The deactivation of the boron injection tank will therefore affect the steamline break transients, but it will not create the possibility of a new or different type of j

accident.

(3) Involve a significant reduction in a margin of safety.

The analyses performed for the deactivation of the boron injection I

tank indicate that the departure from nucleate boiling design basis continues to be met. Additionally, the temperatures and pressures reached in containment would fall below the containment design limits. Since the design bases contain the required margins of l

safety, no significant reductions in margins of safety will occur.

i l

_+

n o

Pi l.

l I

ENCLOSURE 4 1

3 f'

Final Safety Analysis Report 1

Chapter 15 i

Analyses Expected Changes i

l i

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4 SQN.5 DCN No.mo'%

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)

Page 4

The steam release as a consequence of this accident results in an initial C.

increase in st'eam flow which decreases during the accident.as the steam pressure falls. The energy removal from the Reactor Coolant System (RCS)

I causes.a reduction of r wlant temperature and pressure.

In the presence of a negative moderato temperature coefficient,-the cooldown results in a reduction of core shutdown margin.

i The analysis is rf r demonstrata lhA1_gt, fpo,QgwDgJdtirion is i

satisfied:

ssuming a stuck r ust'er' control assembly and a'singW'ttv'4e. as j

fail r i

the Engineered Safety Features,t'Or: M

5: n: rder-t h0"# "

/ _iWity. after reacter trip for a steam release equivalent to the i

spurious opening, with failure to close, of the largest of any single steam dump,. relief or safety valve, w h wt denn buis wm be. gt The following systems provide the necessary protection against an accidental depressurization of the main steam system.

l. ' Safety Injection System actuation fpom any of the following:

a.

Two-out-of-three low pressurizer pressure, b.

High differential pressure signals between stean lines, j

i 2.

The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal.

(

3.

Redundant isolation of the main feedwater lines: Sustained high feedwater flow would cause additional cooldown. 'Therefore, in

. addition to the normal control action which will close the main feedwater valves following reactor trip, a safety injection signal will rapidly close all feedwater control valves, trip the main feeddater pumps, and close the feedwater isolation valves.

15.2.13.2- ' Analysis of Effects and Consecuences Method of Analysis The following analyses of a secondary system steam release are performed for this section.

'Rw4t

5 8W9 i.orrmN 4

1.

A full plant digital com'puter simulation, "*J:.lReference code,'

l :

to e rmine R_CS_t g b; e%1os+ ton bNBR degn bags K mtt, a

o 2.

n :ndy:M to determine that the reactor deet net r:t;r :riti:0-5' The following conditions are assumed to exist at the time of a secondary system break accident.

5 1

15.2-39 COC4/0115F

-e

-e sw,, - -,

~e e

e-

,,,w,

,--,-n-,

r y

~r r

. ~ -

SQN-5 1

1.

End of life shutdown margin at no load, equilibrium xenon

.s conditions, and with the most reactive assembly stuck in its fully A

withdrawn position. Operation of rod cluster control assembly banks 0

during core'burnup is restricted in such a way that addition of positive reactivity in a secondary system break. accident will not lN lead to a more adverse condition than the case analyzed.

zg 2.

A, negative moderator coefficient corresponding to the end of life 8 n.

I rodded core with the most reactive rod cluster control assembly in j

the fully withdrawn position.

The variation of the coefficient with i

temperature and pressure is included.

The Keff versus temperature at 1000 psi corresponding to the negative moderator temperature coefficient used plus the Doppler temperature effect, is shown in Figure 15.2.13-1.

i 3,

Minimum capability for injection of high concentration boric acid solution corresponding to the most restrictive single failure in the Safety Injection System.

The injection curve assumed is shown in Figure 15.2.13-2.- This corresponds to the flow delivered by one ch rging um 11verin_o gifull contents h cold e

kevloe os l

ce as eeT taken Tor the Tow onc ntration or c acid which swn l

must be swept from the safety injection lines downstream of the RwsT j

Mject'e-te" 'te'et'e- "2 er prior to the delivery of high i

be-ea 1

concentration boric acid (NhM9 ppm) to the reactor coolant loops.

j e

M 50 4.

The case studied is an initial total steam fiow of 228 lbs7second at i

~

1015 psia from all steam generators with offsite power available, y

This is the maximum capacity of any single steam dump.or safety valve.

Initial hot shutdown conditions at time zero are assumed since this represerts the most pessimistic initial, condition.

l Should the. reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when power level reaches a trip point.

Following a trip at power the RCS contains more stored energy than at no load, the average coolant temperature is higher than at no load and-there-is-appreciable-energy stored in the fuel.

Thus,-4he-additional stored energy is removed via the cooldown caused by the steam line break before the no load conditions of RCS are reached. Af ter the additional stored energy is removed,

'coofdown' proceeds in the same manner as in the analysis which 5

' assumes no load condition at time zero.

However, since the initial steam generator water inventory is greatest at no load, the j

magnitude and duration of the RCS cooldown are less for steam line breaks occurring at power.

r l

5.

In computing the steam flow the Moody Curve for fL/D = 0 is used.

e 1

15,2-40 COC4/0115F

~

.l r

.i d-.

.

WC4W

c SON i

6.

Perfect,motsture separation in the steam generator is assumed.

4 C-4 7.

The upper head injection system (UHI) 15 simulated.

As stated in D

WCAP-8185 the significant effect of UHI is to retard the pressure

{

secrease of the RCS.

This in turn, reduces the flow of borated water from the Safety Injection System.

The potentially detrimental E

effect is compensated by boration provided by the UHI.

Results i

The results presented are a conservative indication of the events which would occur assuming a secondary system steam release since it is postulated that all of the conditions described above occur simul-taneously.

Figure-15.2'13-3 shows the transients arising as the result of a steam l

release having an initial steam flow of 228 lbs/second at 1015 psia with I

steam release frcm one safety valve.

The assumed steam release is the maximum capacity of any single steam dump or safety valve. Safety

)

Injection is conservatively assumed to be initiated by low pressurizer 1

pressure although steam line differential pressure would provide a more g'gD"5 c

d d

suff1cten't ne ative reactivity t M

-nuter =!! 5:h2

j n h

't!::!!t;.

e Tiracuvhy irnTlent for the ciseinown~Traigure

.l more severe than that of a failed steam generator safety or relief valve which is terminated by steam line differential pressure or a

(

failed condenser dump valve which is terminated by low pressurizer pressure and level. The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the sient occurs over a period of f ant tl Iw e

o 9e W m W o m b H 1Mt. a b e i h t b H a lue,

15.2.13.3 Conclusions The h pagg"y

, _. Z', "O

2 l" ' ' l ! "Z :" ' ' ' ' ' ' ' ' ' '!'

^

l

____,<,..m a.

m.

ImYS 'b'AfaItchIU$i'+e'IIMt JaWio"ce5'si[v5hil dWbe m 16e core oe AcS occorJ U

e

a

' "" Wr i y *At *S A474: 1%s dowws w fe.ase hewon var Benee k p.em yyy 8h 15.2. *.i iG uiiHretivu vi Ca s er-+d A.G ur n u ~urs c r i ch Spurious SIS operation at power could be cadsed by operator error or a.

L false electrical actuating signal. A spurious signal in any of the

'following channels could cause this incident.

M8 [$

l ll 15.2-41 COC4/0115F

..,. A.<. s i.L.

j

SG-b j

1 1.

High containment pressure g gymsnAl 1

Po0* -

2.

Low pressurizer pressure

{,'li 3.

High steam line differential pressure 4.. High steam line flow coincident with either low average coolant temperature or low steam line pressure.

Following the actuation signal, the suction of the centrifugal charging g

pumps is divertid_fram the volumelontrol-if ak 10 thogint water _

storage tank.f The vaTvTs'TsolatingW4ecoa. Injectuon tank 4444 ~fr'om l

the~ chargin'g pumps and L 1e:ve; ';o';t',n; th; OIT fr;- the injection l

header then automatically open.

The charging pumps then f:::: M ; M y p d e. N S T I

l N sensea4 n4+d ppm) ber!

i: M teht M '-^- t% BIT, through the [ /

n line and in':o the cold legs of each looDJThe header and on pumps aTso*TfaaE automancally Dut' provide no flow when(R*j#

se as ety infec the inw 5

the RCS is at normal pressure.

The passive injection sy(ttem andeta t% ^

MI NNU l had mte-O ss ge;M: -a finw

t >ae-a DCE pr;;;at-k #

'"*D

- -....;n: f,. N M g Y d N M Yi,;7 2..y 7;m o. in a reactor trip followed bf a turbine eM)@$

trip.

However, it cannot be assumed that any single fault that actuates the SIS will also produce a reactor trip.

Therefore, two different courses of events are considered.

Case A Trip occurs at the same time spurious injection starts Case' 8 The reactor protection system produces a trip later in the

(" '

A transient.

For Case A the operator should determine if the spurious signal was transient or steady state in nature, i.e., an occasional occurrence or a l

definite fault.

The operator must also determine if the safetf injection l

system must be defeated for repair.

For the former case the operator would stop the safety injection and bring the plant to the hot shutdown I

conditions.

If the safety injection system must be disabled for repair, boration should continue through the normal boration mode and the plant

brought to cold shutdown.

For Case B the reactor protection system does not produce an immediate trip and the reactor experiences a negative reactivity excursion causing a decrease in reactor power.

The power unbalance causes a drop in T..,

and consequent coolant shrinkage.

Pressurizer pressure and level drop.

Load will decrease due to the effect of reduced steam pressure on load after the electro-hydraulic governor fully opens the turbine throttle valve.

If automatic rod control is used, these effects will be lessened un.tll the rods have moved out of the core.

The transient is eventually terminated by the reactor protection system low pressure trip or by manual trip.

1 15.2-42 COC4/0115F l

by~M The time'to trip is affected by initial operating conditions including 5

core burnup history which affects initial boron concentration, rate of C

change of boron' concentration, Doppler and moderator coefficients.

r Recovery from this incident for case B is made in the~ same manner described for case A.

The only difference is the lower T., and

)

pressure associated with the power unbalance during the transient.

The p

time at which reactor trip occurs is of no concern for this accident. At l

lower loads coolant contraction will be slower, resulting in a longer k :'

l time.to trip.

p l

15.2.14.2 Analysis of Effects and Consecuences Method of Analysis The spurious operation of the SIS system is analyzed by employing the detailed digital computer program LOFTRAN (Reference 4).

The code simulates the neutron kinetics, Reactor Coolant System, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator.

steam generator safety valves, and the effect of the safety injection

+

system.

The program computes pertinent plant variables including temperatures, pressures, and power level.

Because of the power and temperature reduction during the transient, operating conditions do not approach the core limits. Analysis of s

several cases shows the results are relatively independent of time to trip.

A transient is presented representing conditions at beginning of core i

life.

Results at end of life are similar except that moderator feedback f

[

I effects result in a slower transient.

~

1 The assumptions are:

1.

Initial Operating Conditions - the initial reactor power'and Reactor Coolant System temperatures are assumed at their maximum values consistent with the steady state full power operation including allowances for calibration and instrument errors.

~

2.

Moderator and Doppler Coefficients of Reactivity - A low beginning

~

of'llfe moderator temperature coefficient was used.

A low absolute value Doppler power coefficient was assumed.

3.

Reactor Control - The reactor was assumed to be in manual control.

FF 4.

Pressurizer Heaters - Pressurizer heaters were assumed to be C

nonoperable in order to increase the rate of pressure drop.

e gg.3.);

m --. ~. m - -

- wNe sw i

S.

%m Injection - At time zero two charging pumps inject 20 do*PM i

/

borated er into the cold legs of each loop. E. m t @ k M 4 h

7 S m b.* *, % w a % 4 ps, m.- m a.- u ;

ur x 5

(

r 15.2-43 COC4/0115F L

i l

. ~...

54Ps-2

~

TAetE 15.1.2-2 15heet 2)

(Continued)

StaeanR1r OF INITIAL E0181Y10181 Age CdMPUTER CtK1

INITIAL N555 REACTIVITY COEFFICIENT 5 THE#ttet P0wtR OUTPui ASSUMED A55ts(0 MODERATOR'"M00ERATOR

COMPUTER TEMPERATURE DENSITY fAULIS CODES UTILIZED f AU*f) fAwam(ul DOPPLER (21 titiT)

CONDITION II (Continued) l

' Loss of Normal 8tK0UT N4

88 4 3577 i

Feedwater toss of Off-5ite BLK00T l

Power to the un len 3423 Plant Avalliaries (Plant Blacheet)

Encessive Heat MnRVEL 8.43 tower e and 3423 Removal Due to Feedwater System Malfunctions L

Excessive toad LOFTRAN 5 and 0.43 tower ~

3423 Increase i

Accidental Depres-LOFTRAN S

t9pper 3423 surization of the Reactor Coolant 4

Syste*

WISE KEviSE Accidental Depres-

-M48YH-Function of

-3,4-pcm/PF 0

surization of the LoFTRAN nederator

-23 (5=bcritical)

Main Steam System Density i

See Swbsection I

{~

(rigure 15.2.i3-1) 15.2.13 4

s a

Inadvertent Operation LOFTRAN 9

tower 3423 lo8 j

of ECCS During o z Power Operation 79 1

a k

.u g

t$63F/COC4

,h 1

e

~<.

_7 0

t.

L_.

.\\

N f

50N-4

~4.,,.

TABLE 15.1.2-2 (Sheet 4)

'- l ',

(Centlawed).

MeetARY tr INIf tat enssalf tons Asa CastruTER CODES *.

t I, l.

INITIAL stss5 e...

l REACTIVITY COEFFICIENTS THERnnt P0wFF SUTPUT j., '

Assunto Assunto i

' MODERATOR'" MODERATOR"'

COMPUTER

- TEMPERATURE DENSITY

[ j' Y ','.* FAutTS CODES UTILIZED fAk/*F1 (Ak/am/cc)

DDPPLERf21 litifi

,'s CONDITION IV (Continued)

. avtsf vesc 4

Major secondary

+wRvEt-

'TNC Fonction of

-3.3.pcm/F S

Piederater-

, y,q (Critical) system pipe r*P-t c FTP.:.

tore up to and Density See -

Including double-15.4.2 (Figure' 15.4.2-1) ended rupture (Rupture of a 5tese Pipe)

Steam Generator M

MA NA M

3577 Tebe Rupture 8

Upper,

2396 and 3423 Single Reactor PHOENIX. LOFIRAN Coolant Pump THINC,,FACIRAN Locked Rotor Fuel Handilog HA NA NA 3577

+*

i Accident e

-1 pcs/*F 90L -

Consisent 8 and 3423 Rupture of a Cen-TWisetLE. FACTRAN I

trol Rod Mechanism LE0pARO

-26 pcm/*F BOL.

with lower limit s %

Housing (RCCA figure 15.1.6-1 e

Ejection) 4 4

j t Notes:

a s.

~

(i) Only one is used in an analysis i.e. either moderator temperature er moderator density coef ficient.

4 c2 %

(2) Reference Figure 15.1.6-1 O

,p t

f, D

9%3F/COC4 4

.,n,

~ _,, _ _ -,,,

,.c...

_ _ _ _ - _ _,. - _ _ _ _ = _. _ _.. ____-__:._..__.

(

l 6937 97

'R C.

epEc6

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k I.06 h

R0 POWER 1000 PSI thD Of LIFE A000tD coat witu ont ace 1.05 4'

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, Figure 15.2.13-1 Vorlation of KEFF wie in m e m g 9 e

.--,->,<.w

-.n...

s

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

w--

i 59N DCN No.*C33

Page l

1.06 i

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l ZERO POWER. 1000 PSIA I.05 END 0F LIFE RODOED Coat WITN out RCCA i

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1 1. 0 16 i

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1 250 300 350 160 0

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CORE AVERAGE TDFERATURE ('F)

)

l

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Figure 15.2.13-1 variacion of Kerg with core Temperature j

e l

D

[

S

,__.,..-_-_,_.,,_,,,,,.__,_-__,___,,_,.-,,,,__,.,.,-,,.._..r.___....,

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400

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200

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0 0 [ 100 200/300 400j500 600 700 800

/

/ SAFETY INJECTI,0N FLOW (GPM)

Figure 15.2.13-2 Safety injection Curve 1

C-O O

.u

,,+...n--

.-,y r.

,,.w--

~.

.y.

,~r.

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1600 i

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1000 i

800 I

600

..e 1

400 u.

200 y

.~

I 0

4 1

0 100 200 300 400 500 600 700' 800l SAFETYINJECTIONFLOW(GPM) t 1

ar i

L L

Figure 15.2.13-2 Safety. Injection Curve I

l

4000

'2h Patssy(iZtt twills 60 stCONDS 7

20

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-5.0 c

100 200 300 400 500 TIME (SECONDS)

/*

Figure 15.2.13-3 Transient Response for a Stoom line Break Equivalent

('

2 8 L /See et 1015 PSIA with Outside Power

+ - - - -

.-,r

,,,--..c,.-m,-.

,,c.,-

w g-,

y w.n

-ew--w.,w.

~ - -

K=-,'"

. ; eeWD tw tiWi SQN

\\

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g l~

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1

+

p b

1 22500 - -

20000 4:

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.11500 - -

,15000 - -

15

.12500 Tw

.10000

" l:: k.

gE

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.07500

.05000 a

o

[

.02500

e. 0

!!00.0

^

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g 2000.0 - -

1750.0 a

a 1500.0-g E 1250.0 -

f W

1000.00 - -

750.00-20.00 600.90

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pf 500.00 e -

8 W

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= 58

$ g 400.00 -

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300.00 - -

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g goog.co.

C. A

=

Q 1000.0 I -f000.0 - -

-2500.0 O.0 100 200 300 400 500 600 TIME (Sec)

F.3m /f,7./3-3 f!0'= ;

TRANSIENT RESPONSE FOR A STEAMLINE BREAK EQUIVALENT tw f,fg TO 228 LSS./SEC. AT 1015 PSIA WITH OUTSIDE POWER e

AVAILABLE.

' b..

s..

c SQN.3 gg M.Y\\

(

TALEj,,5 E

ntinued)

TIME Sl7.f NCE OF EVENTS FOR CONOTIION !! EVENTS Accident Event Time (Sec.)

i Excessive 1.oad Increase l.

Manual Reactor 10% step load increave 0

i Control (BOL)

Equilibrium conditions reached (approximate times only) 200 2.

Manual Rea: tor Control (EOL) 10% step load increase 0

1

~

Equillbrtum conditions reached (approximate times only) 50 3.

Automatic Reactor Cor. trol (BOL) 10% step load increase O

Equilibrium conditions reached (3) 4 Automatic Reactor Control (EOL) 10% step load increase 0

Equilibrium conditions reached (approximate times only) 50 Accidental depressurization of the' Reactor Coolant System Inadvertent Opening of one RCS l

Safety Valve 0

l Reactor Trip 29.3 Minimum DNBR occurs 31.5 1

Accidental deprevsurization of the Main Steam Safety System Inadvertent Opening of one main steam safety or relief valve 0

ggyise.

Pressurizer Emotles MO-ig7 20.^00 pp; boron reaches cc AE Q______

M9 257 UHI initiation time 4M. 2,61 (3) Old not reach equilibrium within the time' scale of Figure 15.2.11-2 3

Revised by Amendment 3 d

COC4/0723F O

e6 e

l --

o

.y c((bb SQN-3 OC

(

TABLE 15.2-1 (Sheet 7)

(Continued)

TIME SE0VENCE OF EVENTS FOR l

CONDITION 11 EVENTS Accided Event Time (Sec.)

i Inadverteni Operation of ECCS during Power Operation Charging pumps begin injecting borated water 0

Low pressure trip point reached 64-Rods begin to drop 66

/

Condit/onIVevent tsele%

od Major Secondar System Replee w Pip Ruptu e

,f, v. s.

Case 4 team line r ptures 0

i Criticalit attained 18 Pressurtz r empty 15 l

20,000 m boron reac s loops 20 UHI 1 lation time 16 1.

Case b Stea line rupture O

Cri cality attalded 14 /

Pr ssurizer emp 17/

,000 ppm bor reache loops 27-HI initiatto time 5.5 v

i i

Case c.

Steam line uptures 0

Criticall attain 21 i

Pres'urt r empty 16 s

20,000 pm boron reaches loop 30 UHI 1 lation ime 17 l.

Cas d Ste line ru tures Cr icality ttained 7

/

essurizer mpty 18 0.000 ppm oron reache loops 32 UHI Inlti lon time 39 Revised Amendme 3

\\

3 i

t L

COC4/0723F O

9 t

~,

- ~

I

CN NOI"'G?

a~

I SON-6 pago__

1 Fast-acting isolation valves are provided in each steam line that will f( ;

fully close within 10 seconds of a large break in the steam line.

For breaks: downstream of the isolation valves, closure of all valves would t

p completely terminate the blowdown.

For any break, in any location, no L

more than one steam generator would blowdown even if one of the isolation b

-valves falls to close. A description of steam line isolation is included e

P in Chapter 10.

Steam flow is measured by monitoring dynamic head in nozzles inside the steam pipes. - The nozzles which are of considerably smaller diameter than the main steam pipe ar3 located inside the containment near the. Steam generators and also serve to llait the maximum steam flow for any break further downstream.

15.4.2.1.2 ' Analysis of Effects and Consecuences Method'of Anal's1s-v The analysis' of the steam pipe rupture' has been performed to determi'le:

1.

The core heat flux and RCS temperature and prnsur.e r.p.tul.tino from s

Reference g code has bee,.iteanLg' The "M"!C. ; LoFTP.Af[ 8t<

coolsiopa f.o]Jowjag_the-n used.-

%8h The therma [and Iiydraulic behavior of the core following a steam line

]

~

ll 2.

L break. A detailed thermal and hydraulic digital-comouter calculation (THINC Code Paragraph 4.4.3.1) has been used to determine if DNB occurs for the core conditions computed in (1) above.

'6f.,

.(

The following conditions were assumed to exist at the time of a main steam line break accident.

1.

End of life shut.down margin at no load, equilibrium xenort conditions, and the most reactive assembly stuck in its fully

. withdrawn position: Operation:of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in the steam line break accident will not lead to a more u

adverse condition than the case analyzed.

2.

The negative moderator coefficient corresponding to the end of life rodded core.with the most reactive rod in the fully withdrawn position:..The variation of the coefficient with-temperature and pressure has been included. The effect of power generation in the core on overall reactivity is shown in Figure 15.4.2-1.

~,

The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculations.

Further, it was conservatively assumed that'the core power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod. To verify the conservatism of 15.4-16 0117F/C0C4 en 9

/

4 q

1

..-. w -

L SON-6 093 Io'J"W

.i dry----.-

_m.

(

'this method, the. reactivity as well as the power distribution was v

L l

checked for the statepoints shown on Table 15.4.2-1.

These core S

analyses considered the Doppler reactivity from the high fuel f

temperature near the stucs RCCA, moderator feedback from the high 5

water enthalpy near the stuck RCCA, power redistribution and O

nonuniform core inlet temperature effects.

For cases in which steam generation occurs in the high flux regions of the core, the effect of 5

vold' formation.was also included. It was determined that the reactivity employed in the kinetics analysis was'always larger than J

T the' reactivity calculation including the above local effects for all 4

.I L

statepoints in Table 15.4.2-1.

This result verified conservatism,

~

L 1.e., underprediction of negative reactivity feedback from power generation, t

.)

V 9 evisc

}

3.

Minimum capability for injection of high concentration boric acidA5 shown ?

(approximately 2^ ^^^ ppm) solution corresponding to the most,/

1 restrjetivesingI@11urein esft in ection systeyTfie L.

In]ectlon curve used s own in igure 1.

1 -2.

s corresponds-T to< the flow delivered by No credit has been takeit'To_g its.fulflow 4tose t

e.n.g,quBD-.dal.i. veri n y

)

concentratTo,n.eg head,e,r /ac utr n

r th'e low 0.1he cold 1 r

ti6rfc acid which must be swept from the safety

- -- N swn -

l

. injection lines downstream of the t= : !:j::ti;; t::t i:ckt4en AwS'T 1

[

! :: prior to the delivery of high concentration boric acid to the reactor co_olant. loops.

r y

s.

L 4

Four combinations of break sizes and initial plant conditions have U:-

been considered in determining the core power RCS transients:

-a. -Complete severance of a pipe outside the containment (downstream L

of the steam flow measuring nozzle) wtth the plant initially at 7

no load conditions, full reactor coolant flow with offsite power available.

1 b.

Complete severance of a pipe inside the containment at the outlet of the steam generator (upstream of the steam' flow measuring g'

. nozzle) with the plant initially at no load conditions with offsite power available.

Case (a) above with loss of offsite power.D..,, -~ ~

.....,.,.-.w=~--~-----.

elt::: n with tu s e

c.

fety ' j::t!:n :!; 2!

Loss of offsite powerY[Q

'-iti:ti:n of th:

results in coolant pump coastdown.

o with

-d.

Case.(b)- above with the loss o.f offsite power.54 mutt:::: :

- u m.

, u.

. a.. m. u. n _ a,.. a

u....... m.

For a steamline break inside containment,'with a failure of an MSIV in another steamline to close, the steam generator connected to the 4

R MSIV will' continue to release steam through any lines or valves that

)

may be open downstream of the MSIVs or upstream of the failed MSIV.

Normally, there are open lines to the main steam reheaters, turbine s

gland seals, main feedwater pumps, and possibly the turbine-dr'lven i

auxillary feedwater pump (steam for the, auxiliary feedwater pump is drawn from two steamlines upstream of the MSIVs). During the 1

15.4-17 0117F/COC4 a

I

.i-DCN No. moisW SON-1 Poge

~ t

.steamlinf break, steam flow to the main feedwater pump turbines and

[Q the main steam reheaters will be terminated. The flow to the main feedwater pump turbines is terminated by stop valves which actuate l

automatically on receipt of a safety injection signal. The flow to the reheaters becomes negligibly small'because the reheaters are the condensing type. Main steam flow which condenses the reheat steam E

ceases when the high pressure turbine stop valves close, and the reheaters effectively become a water trap. The remaining steam flow-amounts to about 20,000 lbs/hr., or less than.2 percent of nominal steam flow.- In order to encompass any additional steam release through unidentified lines and drains, and also to. noticeably perturb the steambreak results, this additional steam release was i

conservatively assumed to be mote than 100,000 lbs/hr. Even with 3

this high value for additional steam release, the steambreak analysis results were not significantly affected.

The greatest deviation calculated was less than 0.02 percent in the oeak core neat finr.

j Since the steanline rupture causea the reactor coolant systes to cooldown, there wonid be,no' reason (or signal) for the

(;

power-operated relief valves to open.

These are fail-l L

closed valves.

Therefore, any postulated nalfunction of a 4

l power-operated relief valve snat be considered.an independent 1 failure-and inconsistent with a coincident-failure-anywhere-else (MSIVs).

The-case,of spurious opening of a power-

...v

.s.n.N.- Q,4. b.v -An n..fr i.1 rw i s.r., e.).4 r t.s. s t e a a l-L a e break wi th -

following a large steamline break with subsequent closure of all l

MSIVs'would be less severe than the steamline break' case reported.in

- the FSAR. _ The spurious opening of a secondary system valve, such as f'

a power-operated relief valve, is considered separately and ieported

('

in Section 15.2.13.

The analy'es presented do not consider additional steam blowdown from s

i either of these sources. Steam released from open lines and drains 1:

on the secondary. piping does not significantly affect the analysis results..and the failure of a power-operated relief valve is reported separately..

'S.

Power peaking factors corres'ponding to one stuck RCCA and nonuniform core inlet coolant' temperatures are determined at-end of core life.

The coldest core inlet ~ temperatures

are assumed to occur in the l'

sector with the stuck rod. The power peaking factors account for-the

-effect of the local vold in the region of the stuck control assembly during the return to power phase following the steam'11ne break.

This void in conjunction with the large negative moderator coefficient partially offsets the effect of the stuck ~ assembly. The power' peaking factors depend upon the core power, temperature, W, and, thus, are different for each case @ h and flow y

pressure

~~

w ~~

um The4 values = :d f;,7 thre; of the four steamline break accidents _ ~

analyzed are given in Table 15.4.2-1.

The............. selected

h%3 m ;

on the basis of hot channel factors, core power, and reactor cool pressure. Th; farth ;ese 15 le55 5e e;e (wieu we W D6R.

e core bise aa. shown f

15.4-18 0117F/COC4 O

O w

k..-,=-,.

=

3

.-v

l A

x*

' DCN nom'*

[

SON Page l:

~'

gf Mec l,

parameters esed for each of the +heee cases' correspond'to values SM E

determined from the respective transient analysis. F!n ti;; pint;

/

1 All'the cases e % ve assume initial hot shutdown conditions at time r

ero since this represents the most_ oetsimistic initial c dition

[

Shoulc tne reactor be just critical or o at pow r_a e time of a steam line break, the-reactor will be tripped by the normal overpower protection system when power level reaches a trip point.

Following a trip at power the RCS contains more stored energy than at no load 'the average coolant temperature is higher than at no load l.

'and there is appreciable energy stored in the fuel.

Thus, the

' additional stored energy is removed via.the cooldown caused by the steam line break before the no load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the

,Y additional stored energy has been removed..the cooldown and reactivity insertions proceed in the same manner as in the analysis

,~

which assumes no load condition at' time zero.

i l

However,- since the initial steam generator water inventory is j

l-greatest at no load, the magnitude and duration of the Reactor L

. Coolant System cooldown are less for steam line breaks' occurring at.

power.

6.

In computing the steam flow during a steam line-break, the Hoody Curve (Reference 22) for fL/0 = 0 is used.

4 7.

Perfect moisture separation in the steam generator is assumed. The assumption leads to conservative results since, in fact, considerable -

l water would be discharged.

Water carryover would reduce the i

magnitude of the-temperature decrease in the core and the ' pressure increase in the containment.

L 8.

The Upper Head Injection System (UHI) i s simulated. During a design l'

steamline break accident, the reactor coolant system (RCS) pressure h

decrease may be large enough to' actuate upper head injection (UHI).

The injection flow rate is a strong function of the RCS pressure--the flow being higher for a lower RCS pressure.

l.

The UHI flow rates are based on the following model.

The pressure-H drop (AP, Ibf/ft') across a component is given by 8

V AP = Ki o 2g.

1 15.4-19 Oll?F/COC4 e

O t

-wo-3~~-v-a

.m.

(

e SQN-4' DCN No.*f -

i Where:

Poge -

^

loss coefficient (dimensionless)

Ki

=

fluid' density (Ibm /ft')-

p'

=

fluid velocity (ft/second)

V

=

32.2 lbe-ft/lbf-second' ge

=

Multiplying the right-hand side of equation (1) by pA'/pA'

]

gives' AP =

K,o'V'A' 4

2p'g,A' m

)

I

(-

or using - e = pVA for mass low rate (Ibm /sec), the pressure drop becomes AP.= v' Kap 1

8 Where Ka 15 a' geometrical constant (lbm-f t'/lbf-sec ).

L Solving for a gives the following expression for the UHI system.

i flow rate:

I' i

w - K4 PAP-The pressure drop used in the model is the difference between the UHI-4 gas pressure and the RCS-pressure. The density over a-given time-step 11s assumed to be constant at the value corresponding to the i

pressure at the beginning of that time step.

The proportionality constant,.K, is an input to the-code.

H The expansion of nitrogen over a time step is assumed to be 1sentropic. The-ch uge in nitrogen volume is. calculated as:

(Rshe AS Vn: =Vaa.+ hat Where Vne. is the volume at the beginning of the timestep and E is

. the average flowrate calculated during the timestep. The pressure is then calculated from:

PnVK, = P.V?:.

Where y is !.4 for nitrogen

~

.15.4-20 01.17F/COC4 1

s

~

.. ~ -....

j 1

. _ _ __. __ _ ___ _ __ _ _ _ _ _ _ _ ___ _ _ _ _____ _ _ _ _ _ _ _ _ _ g

DCN thlityA SON-6 b" b ' ' -

s LoFTMt4

(:x.

Since "'a'!:

s not used in the analysis of LOCA, it does not have to M

kvwf.,

e e high UHI flow rates induced by the severe depressuri-

~

%5 ration of LOCA. The upper head of the reactor vessel remains full of Sfo @ p eoplad-water as.it receives flow from the UHI accumulator.

LoFTRAt0 In MARVR, he boron concentration and enthalpy are determined in the manner:

X...

(WXaot + X.,M..)/(Wat + M..)

Where X'can be replaced by either H or B., W is the accumulator

' flow rate, and M.y is the mass in the dead volume.

As. stated in WCAP 8185 the significant effect of UHI is to retard the pressure decrease of the RCS. This in turn reduces the flow of borated water from the Safety Injection System. This potentially detrimental effect is compensated for by the boration provided by the UHI.

8 The RCS depressurizes and cools down as heat is removed via the assumed ruptured steamline. Depending upon the relative rates of temperature and pressure decline, flashing may occur in the RCS at locations other than the pressurizer. In a plant without UHI, the-primary coolant system volume in which flashing will occur first is

'the upper head-of the reactor vessel. The temperature in this region tends to be higher than the temperature in other regions, which experience higher coolant flow rates.

Water in.the upper head of the Sequoyah reactor vessel does not flash during thersteamline rupture.

If a steamline 1s. assumed to rupture when the plant is in a hot shutdown condition and the coolant temperature in the reactor vessel upper head is assumed to remain at its initial no-load value (547*F), then the reactor coolant system would have-to depressurire to 1020 psia (saturation pressure at 547'F) from 2250 psia before any quality would be obseryed in the upper head. Meanwhile, the UHI system is conservatively assumed to add cold water.into the reactor vessel' upper head when the primary system pressure drops below-the conservatively assumed 1300 psia 6

,,'m h ece,s6 4

5't he ""*!E' ode calculates the mass and energy of the fluid in the uppe d based upon the incoming UHI flow and enthalpy, enthalpy and flow from the lower regi g

essel, and any heat

-input-from the vessel walls. p+ ;h. grjdel the UHI water In the flows into the node represent uw ei m w v of the' reactor vessel, where it mixes with the resident reactor coolant (see Figure 15.4.2-6).

The UHI system prevents flashing in the reactor coolant system during a steamline break by cooling the upper head region and by adding mass to the primary system, which retards the depressurization.

15.4-21 Oll7F/COC4 f

-l DCN No.e25,21!L' SQN-6 page Only'the' Pressurizer water flashes and this. void volume is easily

[f determined from reported plots of the pressurizer water volume Vl history.

Any heat input to the reactor coolant tends to retard the cooldown.

resulting from the steamilne rupture and thereby mitigate the adverse effects of the accident. If the core returns to power, it does not reach as high a power level as it would have reached if the heat

. input-were not accounted for. Heat addition does not significantly.

diminish the margin of subcooling since it retards the l

depressurization as well as the cooldown..

L Heat ' transfer from the hot walls to the fluid in the upper head and the pressurizer is very small. Both regions are outside the active circulation path'of the coolant. The pressurizer is filled with l;

saturated steam and water. The water remains at the saturation-enthalpy as it flows out of the pressurizer during the steambreak i

cooldown.. The water saturation-drops as the pressurizer empties.

The total temperature decline is about 50*F based upon the depressur-12ation during the outsurge. On the average, the temperature drop ls about 25'F and the heat transfer area is half of the initial area (at no load the water inventory is about 25 percent).

Therefore, the heat. transfer to water is small, due to the low AT and heat transfer area, and this heat input to water, however small, would be used for flashing anyway. This would produce more steam, which tends

.to retard the depr ization. Once the pressurizer is empty, heat

(

transfer from g alls to steam is very poor, y

l The heat trans e o the water from the metal in the rest of the

[

primary s'ystem is much greater than the heat contributed by the pressurizer walls.

When UHI is added, the water in the upper head is cooled and remains below the saturation temperature.

Heat transfer from the reactor-i.

-vessel head is greater in this case, but small compared to the cooling *effect of the UHI water, and the_ heat addition to the primary coolant from the steel walls in the other regions of the primary system which are also neglected in the FSAR analysis. When all these j

p heat sources are considered, the cooldown and consequences of the L

steambreak are significantly reduced.

L The maximum UNI flow rate will occur during a major loss of coolant accident and will amount to about 3000 lbs/second.

The UHI flow rate Lo calculations are described in NCAP-847939. The maximum UHI flow rate during a steamline break is a small fraction of the UHI flow rate j

during a LOCA, rarely exceeding 10 percent of the LOCA flow.

The l

average UHI flow during the first 200 seconds of a steam break is about 50 to 60 lbs/second.

g

_s

?

The.UHI fl ow rate is based upon the pressere (l

drop between the UBI system and the reactor coolant spstem.

N J

L x

L 15.4-22 Ollff/COC4 4

-. -.-b r-.

.c

a 4r' h

V SQN-4 g

C v

. C' [,

The core flow rate = 1s a constant volumetric flow, and any void, if T

I s

-present, would affect the mass flow rate through changes in the average coolant density. When the reactor coolant pumps are running, p

the core flow rate exceeds the maximum UHI flow rate by more than a

- This is assuming the maximum UHI flow rate during a b 4.,

factor of 10.

LOCA. Compared to the average UHI flow during a steam break, the core flow rate is more than 600 times greater.

@+

.The upper head na-ymMtar6M85 W 1

nsia e pressure assumed in this analysis is psikj med n the an kvu V and'the actual UHI setpoint will be 100 psi.

65 5%Q 4

'Sens'1tivity studies were performtd for the Sequoyah Nuclear Plant to L

~?-

determine the effect of raising or lowering the UHI setcoint assumed in the_ steam line rupture analysis. A high UHI setpoint results in a relatively early actuation of the UHI system during the reactor

%4 W s t=', r4 - Mture.

sw DTitFrnuhn i.&w-@% 3.me m woron p s ewwc earm ogwn

. m eesa in..

..:.- di" -~~htnrtTor coolant system h

The UNI ad o

'depressurization:and therebyYeNee the safety injection w e

wpy delivered, due to the relatively higher backpressure.

net result O

is that slightly higher peak power levels are attaine following the i

return to criticality during_ a stear. line rupture cooldownsh A tod m P'?**-

i Therefore, the assumption that the UHI accumulator pressure is at the

l' tw h4 A end of the setpoint range is conservative.

(-

9

=~~=__m-_--_

32 Results T

The results presented are a conservative indication of the events which

.i would occur assuming a steam line rupture since it is postulated that all 5

of the conditions described above occur simultaneously.

t

~

Y' Core Power and Reactor Coolant System Transient 3

Figure 15.4.2-2 shows the RCS transient and core heat flux following a X

main steam pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial _ no load a

-condition (Case A).

The break assumed is,the largest break which can J

occur anywhere outside the containment either upstream or downstream of a

the isolation valves. Offsite power is assumed available such that full i

reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one steam generator. Should the core be critical l

at near zero power when the rupture occurs the initiation of safety injection by high differential pressure between any steam line and the 3

remaining steam lines or by high steam flow signals in coincidence with

,1 either low-low RCS temperature or low steam line pressure will trip the reactor.

Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steam lines by the high steam flow signals in coincidence with either low f.

RCS temperature or low steam line pressure. Even with the failure of one m

'y

.15.4-2'3 Oll.7 F /COC4

~

[

,[-

~

.c 7

~

.-l'

-~_...

'l

~i a

Y pcn No. n @

[

SQN-6 L

[

Poge (jl valve, relea'se l's limited to no more than'10 seconds-for the other steam generators while' the one steam generator blows down. The steam line-isolation valves are designed to be fully closed in less than 5 seconds f

after receipt of closure signal with no flow through them.

The steam flow on Figures 15.4.2-2 throust1,14A,2.49. erpt.s ste.aa.f-tow

.amJetrAts-onJ1,y.,/ In ;ddi t1:n,1

;:::r:t: : Revat, j wee-+Hemed t:.!::h:r;: thr:e;b th: 57::5 fer th: 't--t 10 :::end:.

6 Sl%Q

> Tha

-l

?!!"--t'^- that !!' :t::: ;: : :t:r: 51:::::: 'er th:

t

/

10 5:00nd: !: ::::: v:tiv: utt', 7::p::t t th: : r; :::tiVity tr:::!;nt j and th:.:::.:nd ::;rgy 7:! :::.

Th: 10 :::end v:! : :;r:::rt: : very i

!c^; t!-- d:!:y ':r t5: :?;n:! ;:::r:t5:n, tr:::-* tt:1, r ::!;t :nd 5255-e'nt :!eter: " th: :t::r'in: ?:0!:t!:n v:!v::.

Th- !en; t're l $.

d:!:y :::er:: : :::::rv:t'v:!y !:r;: :::r;y r:!:::.

o Sh:r ::::t d:r' ; ;e::!b!: :!d! ; '- th: re::ter :: ! ant :y:ter and '!ce j

-Mcck:;;, ; ! n; t'm: d !:y 'er :10:er Of the :t te!Lne !:0!:t!:n va!ve:\\

is act aeres sarily ceaservative.

In order te che d fer pe::!b!: ve! din; j 6'

H in the pei= ry c^^1:at rytte=, a stes=14 ae break ette! t!On ett performed '

i ia which the i 0! tien va!ve: e:r: :::cred : : !::: tu: ::::nd: 'r:th:r

(

thia ~10 :^-^-dt)~2fter the br ak Me vetding, 2nd th:r:f: : r.: f!:w

_ /b!echt; rete!t:d.

The degree Of tubcaa!'n;

'a the pr' e y :^^!!=t i tyster 91: net :!;n!::nt"y Off::::d. Th: :el;;1 ted :cwe teet fis

/

teaded te be !c"e*

The !ea; t'-- d-!!y '!0 : cend:) !: ered 5:::e:: th:j q

add!ttent! ear: -e!

e ha: a b!;;er ^#fect cet:!de the MSSS than th:

marly~ uelve c1eruee ti== het

^a the prie ry :^^!:nt t:rper:ter:.

As shown in Figure-15.4.2-the core attains criticality with the rod-j cluster control assemblies inserted (with the design shut Wse one stuck assembly) before boron solution of approximatel 0.0% pm enters ;the RCS from the Safety Injection System.- The dela ts of the time to receive and actuate-the safety injection signal and the time to completely open valve trains in the safety injection lines. The safety--Injection pumps are then ready to deliver flow._ At this stage a further delay time is incurred before boron solution can be injected to 6

the RCS due to low concentration solution being. swept from the safety Injection lines. A peak core power well'below the nominal full power y

value is attained.

gg l

l950 )

1 The calculation assumes the 20,000 pm boric acid is mixed with, and g

diluted by the' water flowing e RCS prior to entering the reactor t:

-core.

The concentration after mixing' depends upon the relative flow rates in the RCS and in the Safety Injection System. The variation of mass flow rate in the RCS due to water density changes is included in the t

l-calculation as is the variation of flow rate'from the Safety Injection System, UHI and the accumulator due to changes in the Reactor Coolant

. System pressure.

The Safety Injection System flow calculation includes L

the line losses in the system as well as the pump head curve.

15.4-24 Oll7F/COC4

,--,,..-.m.

r.,

we

--v r e

+e

--e e r-w -,

' - - * - * ' = -

c..

I CH No. M M l

O SON-6 Page The accumulat' ors provide an additional sop tbwet+r-#t+r-4he gY l

V RQfprestwre-derones,Jo,)a,1 y 4Waf The.nt:gr:t:d flow rat: Of N 3

470t;d Ot:r f7;; th: Of:ty hj::ti:n syst:e for each of the four casejs Shom analyzed is shown 1n_ Fig _ure_15.4.2-7. y coec. becen ce<w*cerierg

=

Figure 15.4.2-3 shows Case B. a steam line rupture.at the exit of a steam generator (upstream of the flow measuring nozzles) at no load.

The 6

sequence of events is similar to that described above for the rupture outside the containment except that criticality is attained earlier due to more rapid cooldown and a higher peak core average power is attained.

Figures 15.4.2-4 and 15.4.2-5 show the responses of the salient parameters for cases c and d respectively which correspond to the cases g

i discussed a with additioqQ,) pts,p0MAJte pcwe6 time includes M

$ generated. The Safety Injection System delay Is de on 1 start the emergency diesel generator-4*4 Ws a.

s y IT c e

o A

in the similar case with offsite power available.

The ability of the

'l emptying steam generator to extract heat from the RCS is reduced by the decreased flow in the RCS.

For both these cases the peak core power l:

remains <well below the nominal full power value.

i l

It shou.ld-be 'noted that following a steam line break only one steam generator blows'down completely. Thus, the remaining steam generators are still'available for dissipation of decay heat after the initial

< (-

transient is'over. In the case of loss of offsite pcwer this heat is

(_

removed to the atmosphere via the' steam line safety valves which have been sized to cover this condition.

Genehicthermalandstressanalysesandsubsequentfracturemechanics analyses of. reactor vessels have been performed for 4-Loop plants.

These l

analyses were applied to a 4-Loop reactor vessel having material I

propertle's and end of life (40 years) accumulated fluence similar to the-Sequoyah vessel. The' fracture mechanics analysis uttilzed linear elastic' fracture mecha'nics method in the evaluation of the reactor vessel

-l Integrity.- The fracture mechanics analysis results show that the Reactor Vessel Integrity under large Steamline Break conditions would be maintained over the design life of the vessel.

For long term coolinglof a steamline break the operator is instructed to use the intact steam generators for the purpose of removing decay heat and plant stored energy.

This is done by maintaining the steam generator

)6-narrow-range span.

' Steam pressure from the steam generators is. relieved by the steam dump system, secondary system atmospheric safety. valves, or secondary systcm

. relief valves.

The operator is instructed to terminate aux 111ary feedwater flow to the faulted steam generator as soon as he determines which steam generator is-faulted. As soon as an Indicated water level returns to the pressurizer the operator is instructed to turn off the safety injection pumps and restrict the charging pumps as required.

15.4-25 Oll7F/COC4

o

..i

ilo,

-F OCN No. "WM SQN-6 i

pgp

!-?

can be met by simple switch actions by the operators, i.e., closing auxillary feed discharge' valves and stopping charging pumps and safety-l Injection pumps. - Thus, the required simple actions to limit the cooldown and depressurization can be easily recognized, planned and performed within ten minutes.

For the longer time requirements for decay heat

.1 removal and plant cooldowi the operator has time on the order of hours to respond.

i The worst case condition for long term cooling following a steam line break is loss of offsite power with failure of one emergency power train, since the' condition requires the greatest amount of operator action and-

-the longest time to achieve cold shutdown. However, since the plant can i

be maintained safely at hot standby conditions for extended periods of j

time, there is no safety requirement which dictates rapid achievement of cold shutdown conditions.

l With only onsite pcwer available, the plant can be maintained in a safe hot standby. condition using the intact steam generators by supplying 1

feedwater with the auxiliary feedwater system, and venting steam through the secondary side, power-operated relief valves. The relief-valves will i

be controlled to gradually reduce pressure and temperature as the core residual. heat decays.

If the relief valves are.not available, the safety

.' valves will be used for steam dump. In this case, the primary system pressure would be controlled such that adequate subcooling is maintained. Primary system temperature would be maintained at that value 6

L necessary to lift the steam generator safety valves as necessary to match the decay heat from the core.

This temperature would.be approximately H

553*F which corresponds to the lowest steam generat~or safety valve setpoint.of 1064 psig.

For either means of steam relief, the steam generator water level will be maintained within the span of the narrow range Indicators.

l The sequence of even'ts is shown in Table 15.4.1-12.

~

Marcin to Critical Heat Flux 1

i Past experience in performing DNB analyses for steamilne breaks for H 6

7 cores has shown that Case B (inside break with offsitgear m-- Am wqrja r M 6 Q - e 1,345jah alse he s[gh

,#seen Jhan4ts%'a; i

n by ex!? a th:. :t:d p;lnts presentiv Cases A

)

i.viv is.

.c-i-

'and 'B generally have 'very similar temperatures and -pressures, but Case B l

" 84*

hg m p

_ --m

Ae powerpe]

C:n:r:!!y. On!y f:= Of the state pointelpresented in Ta

. 4.'2-Fire subjected to detailed nuclear and therma hydraulic analysis.

F0r C:::

S, th: p0 Mt etth th: '!;h::t per:r !^"0! !! ina'yzad. !! ate part L

experteaca ha! indicated th!! pe!at !! tha ca-whi'h "

--^h'h'"

"a 1

the 10 ::t DMSR. In addit!cn, either the precedia; c secceedia; pe!at 4 depend!ng On the COndit!^a ) !! ana!y2ed.

l A A co-e\\e+e se-t og rhe. Meemha bmk musient snnpounts urt rehe:d p (

m 4e4emne -Ae. med Iw+g conclalon, 77j, b,re hew

-e-swn 15.4-27 Oll7F/COC4

/

p 1

'DCN Nc..Mo'M3 A

!QN-6 Poge R

F t C::: 0 n:id; br::t with 10:: cf off;it: p;;;r), th: ::!nt ::t N

~

L

'!k:!y t: hre: th: 10:::t " "R i: th:.p; int with th: 'igh::t p ver! u \\

L

-r:tte. L':2:!!y, tht: p !-t !: th: On: r! th the hiPest p~ar h U +h C :: ?.

!ther th: pr:: d!n; Or :;;::::::in; p tnt i: :::: :::!y::d.

Shee!d y of the pet-t: int!y: d re:0!t i-Duao' et-1.30. Edd!ticael \\

1 p0?-t:.:y 5: :::!y::d t

'9:: : th:t th: p !nt with th: -i i r- 0"E'

(

1 tr-d!tien 5:: 5::r :::!y::d.

j i

a The pointe-' analyzed for_ this application had4DNBRM greater than 1.30.

L us,1it is concluded.thar tiMTalsum uskTorWrekk is M than 1.30.

gg e Sheun -

1 v

The maximum, linear heat rate for the most limiting steambreak case s

presented in the FSAR was less than@O-kl4/ft, which is less than the)

I.

linear. heat rate which results in fuel melting. Ther: 1: n hn:r- )

'failura marhantem meenr4m+ad v4&h ehle an d 14anse ham + em+n

/

H 15.4.2.2 Maior Ruoture of a Main Feedwater Pipe L

15.4.2.2.1 Identification of Causes and Accident Desertotion j

- A major feedwater.line rupture is defined as a break in a feedwater pipe-large enough-to prevent the addition of sufficient feedwater to the steam generators to maintain shellside fluid inventory in the steam -

i generators.

If the break is postulated in a feedline between the check valve and the steam generator, fluid from the steam generator may also be discharged through the break. Further, a break in this location could f i

. preclude the subsequent addition of auxiliary feedwater to the affected

- ( -lI

, steam generator.- (A break ~ upstream of the feedline check valve would l

affect the Nuclear Steam Supply System only as a loss of feedwater.

This p

case is covered by the evaluation in Subsection 15.2.8.)-

I Depending upon the size of the break and-the plant operating gonditions -

i at the time of the b'reak,-the break could cause either a RCS cooldown (by I

- excessive energy discharge through.the break), or a RCS heatup.

Potential RCS cooldown~ resulting from a secondary pipe rupture'is 1

- evaluated in Paragraph.15.4.2.1, " Rupture of a Main Steam Line."

6 l-Therefore, only the RCS heatup effects.are evaluated for a feedline rupture.

A feedl'ine' rupture reduces the ability to remove heat generated by the core from the RCS bec'ause of the following reasons:

1.

Feedwater to the steam generators is reduced.

Since feedwater is subcooled, its loss may cause reactor coolant temperatures to r

increase prior to reactor tript 1

L 2.

Liquid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip; 3.

The break may be large enough to prevent the addition of any main feedwater after trip.

(

~

15.4-28 0117F/COC4 1

e I'

e l- --L.-

s

. w-.

,n..

..n.v.,...,.p...n_w.

,s,e,.

m.

i DCN NO,dinA 1

i,.I9'%

]-

i SQN-4 j

TABLE 15.4.1-12 (Sheet 1)

TIME SE00ENCE OF EVENTS FOR D* EE

CONDITION IV EVENTS Wrw

,/

Accident -

Event Time (Sec).

)

/

3

~

Major Secondary System-Pipe Rupture-

/_

Steam line ruptures 0

j'

, 1 l.--Case a Criticality attained 18

/,

Pressurizer empty 15 20,000 ppm boron reaches 4

loops 20 j

f

.(

1 2'

Case b Steam line ruptures 0

\\

V Criticality attained 14

)

. l f

Pressurizer empty 17 20,000 ppm boron reaches

/

4 loops 21

' j

3. ' Case c

/ Steam line ruptures 0

l

/ Criticality attained 21

/4 f

/

Pressurizet empty

,/

16

/

,/

20,000 ppm boron reaches

-f

/

loops

/

30,'

/

,/

/

-Steam line ruptures,

,. 0 4.

Case d

/

Cr.iticality attained

/

17

[

/~

Pressurizer. empty /

19-

/

./

- /

/20,000 ppm borop' reaches-

,/

4

)

/

loops 32

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c 4

Major Secondary

. System' Pipe Rupture i

4.

1.

Case a

-Steam line ruptures 0

/.

s.

Pressurizer empty I.

13.4-ll

-l UHI initiation time 21.7.

,' d ~

L Criticality attained

' 30.8 --

- y Boron reaches core 30.8

[

g 2.

Case b Steam line ruptures 0

K l.

Pressurizer empty 15.0 q

Criticality attained 19.8, UHI initiation time 23.0

e;

.i Boron reaches core

-31.8

('O

'l

3.. Case c-Steam line ruptures 0

7 Q

j L

Pressurizer empty 14.6 UHI initiation time 24.1 y

Criticality attained 35.3 Boron reaches core 47.3

?

u+

4.

Case d,n Steam'11ne ruptures 0

Pressurizer empty 16.5' i

Criticality attained 23.3 3.-

UHI initiation time-28.4.

s v

l Baron reaches core 52.3

+

l-

.t O

's L

-Accidental depressurization lL of the Main Steam System e

w c

Inadvertent Opening of one main steam safety or relief valve 0

~

Pressurizer empties 161

[

Boron reaches core 227 UHI initiation time 237 Criticality attained 305 4

. ~ -.

..._J._,*

DCN No."'C# l

(

Page -

SQN

,-he\\ete

~/~ rw qq gep(zg TABLE 15.4.2-1 Sheef1)

,n -.- ~,- m n y _..

CORE P RAMETERS U D IN STEA BREAK DNB ANALYSIS' K

.j l

', Case a. Time Point'

/

4 5

Parame er

,r I

,e 2

3+

l actor Vessel / niet

/

temperature,tb' sector /

f' e

connected,to affected t Steam Geperator, 'F,.'

,f'436. 9 434.6,/ 411.4 405.0 399.7 Reac,t Vessel inf'et

,[

[

temperature to re-

/

ja'ining sect,or', 'F

,/

492.8 4,91.1 486.0 481.3 475.9.'

RCS pressd're, psia -

1143.0 1117.0 1077.0,/

1049.8 1023.5 100 100' 100 7

/, 6. 83 RCS flow,'%

100 100 6.37-('>setflux,%

'. 91

/7.45 7.22 f

. Time, sec.

[32.5

'41.0!

55.'O 6 5'.'O 75.0

/

e j

s' a

\\

9 L.

COC4/0713F'

..q i

e

, rk ' '

IDCN Ho S N

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Page _

a

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De,lty,

SQN

/

b--

j 7

8 TABLE 15.4.2'1 (Sheet 2)

(Continued)

--- s '\\

.C0EPARAMETERSUSEDINShEAMBREAKDNB.,INALYSIS

/

5

/'

Case'b Time Point l

. Parameter

/

I 2

3 4

5 j,

p

/

Reactor / essel inlet

/

)

V temperature to sector /

/

connected to affected

/

(

steam generator. */

382.6 368.4 363.5 355.1 351.4 l

/

' Reactor Vessel,lblet temperature't.o re-i maining sec, tor, 'F 521.8 505.1 497.9 482.7 475.4

/

RCS pressure, psia

/ 1245.0 1107.8 1070.6 984.7 954.7

-t

/

lo,/,

w.- 1 100-

/

RCS 100 100-100 103,

p

(

Heat flux, % 9.67 10.79-10,98 10.3 9.96'.

/

\\

/ Time, sec.

30.0 l5.0 52.5 67.5 75.0 f

/

l i

/

e l

II i

l

..(

COC4/0713F

~

" I -

'7 L.

DCN No.0* W A -

Page_

o j,

....,.. - ~.

,n, bekf.tdm / /

-~

./

^ /.

TABLE 15.4.2

-(Sheet 3)

(Cont'i nued) f'

._/

l

,/

- ~ ~.

/

CORE PARAMETERS USED IN STEAM BREA DNB ANALYSIS s'

/

s

-f

/

-[ Parameter,

/ Case d. Time Poln

/

1 3

4 5

/'.

j j2

/

React essel inlet./

[

-temperature to sector corrnected to affected

/

/ /. sfeam generator,.r*F 375.1 350.1 330.3 318.5' 305.5

- (

j'

,/

t

( ' Reactor Vessel inlet

/

/-

1 temperature,'to re-

/

f maining sector, 'F 529.7 528'.6 528,1 527.5

526.7 l-

/;

/

1524.0 348.6 1277.5 1256.7 1229.0 4.1RCSpr, essure, psia

. (.l RCSdlow,%

40.6 32.2 27.0 24.2 '.' 21.4 l,

't

- 5. ),/

He/

\\'

(/

4.67

/

at flux, % 5.8 643 6.19 y

/

Time, sec.

5.0 35.0 5.0 52.5 62.5

/

9

\\

G l

1 C0C4/0713F

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l

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LIMITING CORE PARAMETERS USED IN STEAM BREAK' ONS ANALYSIS Case Inside break with power (case b)

Reactor vessel inlet.

319.3'F'(Faulted.SGLoop) temperature

. 414.2'F (Intact 'SG Leops)

- RCS-pressure 798.52 psia.

RCS flow 100%(ofnominal)

Heat flux 17.60 (of nelninal)

Time

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Figure 15.4.2-2 Transient Response to Steam LIne Break Downstream

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s e'jplst L'

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TIME (SEC)

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Rye IS/4.2 TIC'J": 1 TRANSIENT RESPONSE TO STEAMLINE BREAK 00WNSTREAM w s/7/s1 0F FLOW MEASURING N0ZZLE WITH SAFETY INJECTION AND WITH OFF-SITE POWER (CASEA)

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ine eo et Exit of

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Figure 15;4.2-3 Transient Response to Steam Steam Generator with Safety injection and Off-Site Power (eme b)

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25 50 TIME (SECONDS) y

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Transient Response to Steam Line Break Downstream of Figure 15.4.2-4 Flow, Measuring Nozzle with Safety injection and i

Without Off-Site Power (case c) i Revised by Amendment 2

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Revised by Amendment 2

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TIME (SEC) p;p 15. + 2. -f TIGUR 4 TRANSIENT RESPONSE TO STEAMLINE BREAK AT EXIT OF STEAM GENERATOR WITH SAFETY INJECTION AND WITHOUT

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Ifater versus Time l

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