ML20005H311
| ML20005H311 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/12/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20005H307 | List: |
| References | |
| NUDOCS 9001240410 | |
| Download: ML20005H311 (81) | |
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%, '> ENCLOSURE 1 PROPOSED TECHNICAL' SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS.1 AND 2 i DOCKET NOS. 50-327 AND 50-328 r '
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Unit 1 VII 3/4 1-11 . n 3/4 1-12 3/4 5-1
-3/4 5-7 3/4 5-11 3/4 5-12 3/4 5 B 3/4 1-3 B 3/4 5-2 B 3/4 5-3 Unit 2 VII 3/4 1-11 , 3/4 1-12 3/4 5-1 3/4 5-7 i 3/4 5-11 h 3/4 5 3/4 5-13 l~ B 3/4 1-3 B 3/4 5-2 B 3/4 5-3 s
l i. 1 l l 900124041o 900112 hDR ADOCK 05000327 PDC . 4 4
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4g * ~ i' \ l INDEX E , LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE _ REQUIREMENTS l
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(.; 4 SECTION, ; PAGE 3/4.5- EMERGENCY CORE COOL.'NP_ SYSTEMS (ECCS) g 3/4.5.1- ACCUMULATORS 1 ,- Cold Leg Injection Accumulators.......................... - 3/4 5-1 { F7 "^*d Ir.j;;tivo Acuumul ^v a......................... m , 3/4 0 gThb chan3e 3/4. 5.2 - ECCS SUBSYSTEMS .T grea6er than or equal to 350'F. . .. . . 3/4 5-5
^ - w.5' .;
avg [Cubmitted 3/4.5.3 /ECCS SUBSYSTEMS.- T avg less than 350'F.................... in1T5 3/4 5-9 ob e - 3/4.5.4 20"^.'! INICTIO" SYSTEtt ' OELare o f)- ..
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-3/4 5 *
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3/4.5.5 . i, 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 m Containment Leakage....................................... 3/4 6-2 Containment Air Locks..................................... 3/4 6-7
~ Internal Pressure......................................... 3/4 6-9 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
-Containment Ventilation System............................ 3/4 6-15 o
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................. 3/4 6-16 l- lR73 L Lower Containment Vent Coo 1ers............................ 3/4 6-16b i L R120 SEQUOYAH - UNIT 1 VII Amendment No. 67,63 116 l# l
~ June 1, 1989
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. REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN 1
LIMITING CONDITION FOR'0PERATION 3.1.2.5 As a minimum, one of the following borated water-sources shall be OPERA 8LE: 1 i 1 a. A boric acid storage system and associated heat tracing with: 1. A minimum contained borated water volume of 2175 gallons,
- 2. Between'20,000 and 22,500 ppe of boron, and i
- 3. A minimum solution temperature of 145'F. '
- b. i The refueling water storage tank with:
- 1. .A minimum contained borated water v of 35,443 gallons,
- 2. A minimum boron concentration of ppe, and
- 3. A minimum solution temperature of 60*F.
h g- APPLICABILITY: MODES'S and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1. 2. 5 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
1. Verifying the boron concentration of the water. .
- 2. Verifying the contained borated water volume, and 3.
Verifying the boric acid storage tank solution temperature when it is the source of borated water, b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water.
. < - SEQUOYAH - UNIT 1 q 3/4 1-11 s
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REACTIVITY' CONTROL SYSTEMS [>' i
~B0 RATED WATER SOURCES OPERATING. '
LIMITING CONDITION FOR OPERATION
;
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3.1.2.6- As'a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:i y, a; .A boric-~ acid storage system and associated heat t 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:
- 1. A' contained borated water volume of between 370,000 and 375,000
- gallons
- 2. Between and ppm of boron,
- 3. A minimum solution temperature of 60'F, and )
4 A maximum solution temperature of '105'F. APPLICABILITY: MODES 1, 2, 3 and 4. s ACTION: L a. With the boric acid storage system inoperable and being used as one
of the-above required borated water sources, restore the storage i system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN 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 SHUTDOWN within the next 30 hours.
~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 hours'and in COLD SHUTDOWN within the following 30 hours.
l-i o c. l' SEQUOYAH - UNIT 1 3/4 1-12 -- 1 1 l' )
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_ 3/4. 5' EMERGENCY CORE COOLING SYSTEMS'(ECCS) 3/4. 5.'l-- ACCUMULATORS - COLD LEG INJECTION ACCUMULATORS t f: LIMITING CONDITION FOR OPERATION i 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open,
. 12 e ith _b. co tai borated water volume of between and g lons of l OHI remval,and "
1460'chkon .has ' c. Between and ppm of boron and 'i been 'cubmHfad y e d. A nitrogen cover pressure of between nd t i, i g. '
. M 'LO APPLICABILITY: MODES 1, 2 and 3.*-
ACTION: a.
.With one cold leg injection accumulator inoperable, except as a result o f of a closed. isolation valve, restore the inoperable accumulator to i- OPERABLE status within one hour or be in at least HOT STANOBY within the next'6 hours and in HOT SHUTDOWN within the following 6 hours.
b. With one cold leg injection accumulator inoperable due to the isola-tion-valve being closed, either immediately open the isolation valve or.be the in HOT next STANDBY within one hour and be in HOT SHUTDOWN within 12 hours.
. c.#
3 With one pressure or water level channel. inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at-leastwithin HOTtheSTANDBY R128-SHUTDOWN following 6within hours,the next 6 hours and in HOT d.# With more than one channel (pressure or water level) inoperable per accumulator, inoperable. immediately declare the affected accumulator (s)
- Pressurizer pressure above 1000 psig.
,. #Cycle Actions e andoutage.
4 refueling d are in effect until the restart of Unit 2 from the Unit 2 R128 d SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 124 August 11, 1989
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*t nEMERGENCY' CORE COOLING SYSTEMS (ECCS)-
SURVEILLANCE REQUIREMENTS (Continued)
. ..... . t d. hv- W.- s7, .,2. -
- 2. Verifying that each of.the following pumps s' tart automatically
. ,,up.on,rf.ceipt of a safety injection signal:
a) Centrifuga1Echarging pump - b) ' Safety injection pump 4 c) Residual heat removal pump. .
- 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:
- 1. Centrifuga1' charging pump Greater than or equal to 2400 psig-
~
- 2. Safety Injection pump Greater than or equal to 1407 psig 13.. Residual heat: removal pump Greater than or equal to 165 psig
- g. By verifying the correct position of each mechanical stop for the C following Emergency Core Cooling System throttle valves:
- 1. Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
- 2. At least once per 18 months.
Charging Throttle Valvesn Injection . Safety Injection Safety Injection Hot PtArnp leg Throttle Valves Cold Leo Throttle Valves Valve Number Valve Number Valve Number
- 1. 63 - 582 1. 63 - 550 1. 63-542
- 2. 63 - 583 2. 63 - 552 2. 63-544 l 3. 63 - 584 3. 63 - 554 3. 63-546
- 4. 63 - 585 4. 63 - 556 4. 63-548
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, SEQUOYAH - UNI'T 1 3/4 5-7
v EMERGENCY CORE COOLING SYSTEMS (ECCS)
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3/4.5.4 00';;;, ;;4;;;;;]Y:;;;;; DELETED gg hBORONINJECTION ~ ~ { ff. s LIMITI NDITION FOR OPE ION 1 1 1 1 6e
.5.4.1 The bor njectiontanks be OPERABLE wit -
- a. 'A- nimum contained ated water vol f 900 gallons, Between 20,000 d 22,500 ppm of b on, and
- c. A minimum lution temperatu of 145'F. -
APPLICABILITY- H0 DES 1, 2 and 3. ACTION: -
-Wi+ the boron inject . tank inoperable, thin 1 hour or b n HOT STANDBY and tore the tank to O ABLE status-to IL delta k/k ated to a SHUTD0 RGIN equivalent
- 200'F within the 6 hours; restor status.withi he next 7 days or in HOT SHUTDOWN w e tank to OPERA 8
;, in the next 12 ho .
SURVEILLANCE RE EMENTS 1 s' 4.5.4 The boron injec n tar;k shall be d nstrated OPERA by: '
- a. Verifyin he contained bora water volume days, east once per 7
- b. erifyi'ng the boron ncentration of once per 7 days, d water in the tan t least
- c. Verifying t water temperat at least once per hours, t
L :/ ', W SEQUOYAH - UNIT 1 3/4 5-11 l - _ . _ _ . . _ _
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F^ EMERGENCY CORE COOLING SYSTEMS (ECCS) De,leb ( [-HEATTRACING LIMIT CONDITION FOR OP ATION 3.5.4.2< At lea two independent chan s of heat tracing all be OPERABLE for the boron njection tank and fo he heat traced por ons of the associ d flow paths APPLI ILITY: MODES 1, 2 3. N: With only one cha el of heat tracing either the boro njection tank or the heat trace ortion of an asso ted flow path OP BLE, operation ma continue for p to 30 days 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 STANDBY within hours and in HOT DOWN with the following 6 h rs. s SURVE NCE REQUIREMEN 4.2 Eac eat tracing chan for the boron inj ion tank and assoc eC flow path all be demonstra OPERABLE: At least once er 31 days by ener ing each heat traci channel, and I
- b. At I t once per 24 hour 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 unti establishment of e librium temperat 5 within the tank.
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- s. i SEQUOYAH - UNIT 1 3/4 5-12 \
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EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.5 ' REFUELING WATER STORAGE TANK i LIMITING CONDITION FOR OPERATION
..3.5.5 ,
The refueling water storage tank (RWST) shall be OPERABLE with: 1 a.
" A contained borated water volume of between 370,000 and 375,000 gallons, b.
2 7 A boron concentration of between and ppm of boron,
-i
- c. A minimum solution temperature of 60'F, and R16 E
- d. A maximum solution' temperature of 105'F.
' APPLICABILITY: -MODES 1, 2, 3 and 4.
ACTION: . w With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be-in following at'least HOT, STANDBY within 6 hours and in COLD SHUTDOWN within the 30 hours. N. 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 tj
- 2. Verifying the boron concentration of the water.
b. At least once per 24 hours by verifying the RWS1 temperature, l-l'.- W MAR 251S82
%d. SEQUOYAH - UNIT 1 3/4 5-13 Amendment No. 12
4- e ls l REACTIVITY CONTROL SYSTEMS BASES ggallons of 20,000 ppm t. orated water from the boric acid
" 100 gallons ofm n m borated water from the refueling water storage tank.
J 82.,062. With the RCS temper,ature below 200 F, one injection system is acceptable
.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. SHUTDOWN 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 of ppm borated water from the refueling water storage tank. g The contained water volume limits include all'owance 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. l
-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 SHUTDOWN 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
g . 4 r EMERGENCY CORE COOLING SYSTEMS BASES-r ' 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 ensure OPERABILITY of each
- component ensures that at a minimum, the ass'.rptions 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 acceptable- -
' level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses, s
3/4.5.4 BORON INJECTION SYSTEM kQk 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 can be sed byfinadverten em cooldown. R cooldown pressurization; oss-of-coolant dent or a ste ine rupture. . The limits njectiontankmi 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 usuab not ecause of tank harge line loca or other physic characteristics The OPERABIL of the redundan eat tracing chann
-the boron in associated with ion system ensur at the solubilit the boron solut n will be m ained above the ubility limit of F at 21000 ppm b on.
/
u i S l SEQUOYAH - UNIT 1 B 3/4 5-2
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EMERGENCY CORE COOLING SYSTEMS l BASES _3/4.5.5 REFUELING WATER' STORAGE TANK i 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 concentr.stion ensure that i
- 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 RWST and the RCS water volumes-with all
; control rods inserted except for the most reactive control assembly. These i assumptions are consistent with the LOCA analyses.
;
The contained water volume limit includes an allowance for water not s-usable because of tank discharge _ line location or other physical characteristics,
.The limits on contained water volume and boron concentration of the RWST 3 also ensure a pH value of between 7.5 and 9.5'for the solution recirculated within containment after a LOCA. This pri band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on-g mechanical systems and components.
} 00
/Wai+ionalIp the OPERABlUTY of +he /M3T~ as pari efthe ECCS ensures ihai svPNcien+ negaHve reacfivd35 i 8 injec+ed in fo the- sore lo counteroc+ any posiVive increase irl reac+ivHy caused by R66 cy61em
\ cooldown . L .- l: 1 l' (
.(
SEQUOYAH - UNIT 1 B 3/4 5-3 Revised 08/18/87 1
s . 4 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
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SECTION PAGE
-3/4.5- EMERGENCY _C_0RE COOLING SYSTEMS 3/4.5.1 -ACCUMULATORS.
i ColdLegInjectionAccumulators........................... 3/4'5-1 L ThisE che r (6= M=: !e = e = s = e t = .......................... =5D
- was e3/4.5.2 ECCS' SUBSYSTEMS - T avg greater than or equal to 350'F..... 3/4 5-5 i M nd ed- 3/4.5.3 less than 350 ECCS SUBSYSTEMS'- T avg F.................... 3/4 5-9 in T:5
' c.hong 8 3/4.5.4 00"0M UL'ECTION SYSTEM O Et.ETE O j 89-Z.5. __._.g_ ,m , , , , .,
__ , , _ ...m. ..................................... si , s ia cot TOCiQ.............................................. Es s i '2 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 1 Containment Leakage....................................... 3/4 6-2 Containment Air Locks..................................... 3/4 6-7 Internal Pressure......................................... 3/4 6-9 , 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 Containment Ventilation System............................ 3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS , Containment Spray System.................................. 3/4 6-16 Lower Containment Vent Coo 1ers. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-16b R61 April 4, 1988 SEQUOYAH - UNIT 2 VII AmendmentNo.//,61 a -
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e,041 , y c- .- , REACTIVITY CONTROL SYSTEMS I V IBORATED WATER SOURCE - SNUTDOWN ;
' LIMITING CONDITION'FOR OPERATION l l
. -3.1.2.5 As'a minimum, one of the following borated water sources shall be OPERABLE:
- a. A boric acid storage system and at least one associated h'est tracing system with:
1
- 1. A minimum contained borated water vo.lume of 2175 gallons. .j
- 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:
- 1. A minimum contained borated water of 35,443 gallons, S
- 2. A minimum boron concentration of ppm, and A minimum solution temperature of 60*F.
3.
, y .
APPLICABILITY: MODES 5 and 6. ACTION: i
-With no borated water source OPERABLE, suspend all operations involving CORE ,
ALTERATIONS or positive reactivity changes. '
'SURVEILLNNCE REQUIREMENTS 4.1.2.5 The above required borated water source chall be demonstrated OPERABLE:
- a. At least once per 7 days by: ,
- 1. Verifying the boron concentrttion of the water, l
- 2. Verifying the contained borated water volume, and
- 3. Verifying the boric acid storage tar.k solution temperature when it is the source of borated water,
- b. 'At least once per 24 hours by verifying the RWST temperature when it is the source of borated water.
- l. .
(, V e 7 '~( - SEQUOYAH - UNIT 2 3/4 1-11 l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - ~ ~ - -- ._ = _. -
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3 o , E REACTIVITY CONTROL SYSTEMS-BORATED WATER SOURCES - OPERATING - S, , m . LIMITING' CONDITION FOR OPERATION 3.1.2J6 As' a minimum, the following borated water source (s) shall be OPERABLE as' required t'y Specification 3.1.2.2: ,
- a. A boric' acid storage system a,nd at least one associated heat tracing ,
system with:
- 1. A minimum contained borated water volume'o gallons,
~
- 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:
- 1. A contained borated water volume of between 370,000 and
-375,000 ns, t ! 2. Between and-iHW ppm of boron, and 3; A minimum solution temperature of-60'F. ?
- 4. A maximum solution temperature of 105*F.
APPLICABILITY: MODE 5 1, 2 3 and 4. ACTIf_N:
- a. :With the boric acid storage system inoperable and being used as one I of~the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT -
STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN
. equivalent to at least-l% 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 hours. t
- 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 hours and in COLD SHUTDOWN within the following 30 hours. SEQUOYAH - UNIT 2 3/4 1-12 -
v mg ' : :n m .. V[': l;' ,, y m. V 3/4.5 EMERGENCY CORE-C00L'ING SYSTEMS n 3/4.5.1 ACCUMULATORS. P ij COLD LEG-INJECTION ACCUMULATORS LIMITI'NG CONDITION FOR OPERATION h
'3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
a'.- The isolation valve open,
, ),.Q ** h. A contained borated water volume of between and allons of um removal, eng borated water,
[.iv5* fica fieri-- has Between and ppm of boron. and iTSbeen chen$e 89-IS. submined
. , 'A nitrogenh [c.pressure of between cover and psig.
y- APPLICABILITY: MODES 1, 2 and 3.* ACTION:
- a. With one cold leg injection accumulator. inoperable, except as a
( result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With one cold leg-injection accumulator inoperable. due to the
. isolation 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 hours.
[' c.#' With one pressure or water level channel inoperable per accumulator, return the inoperable channel to 0PERABLE status within 30 days or E Rll3 be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
' d.# With more than one channel (pressure or water level) inoperable per accumulator, immediately declare the affected accumulator (s)
- l. inoperable, o " Pressurizer pressure above 1000 psig.
# Actions
' Cycle c and doutage.
4 refueling are in effect until the restart of Unit 2 from the Unit 2 R113 L SEQUOYAH - UNIT 2 3/4 5-1 l Amendment No. 113 August 11, 1989
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's s EMERGENCY CORE COOLING SYSTEMS ,
~
SURVEILLANCE REOUIREMENTS (' Continued) Yi 2.- Verifying that each of the following pumps start automatically upon receipt of a safety injection signal:. a)L ' Centr Ougal charging pump q b)~ Safety injection pump p , c) Residual. heat removal pump
- 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: -
- 1. Centrifugal charging pump Greater than or equal to 2400 psig 4
- 2. Safety Injection pump Greater than or equal to 1407 psig
- 3. Residual heat' removal pump Greater than or equal to 165 psig
- g. .By' verifying the correct position of each mechanical stop for the following ECCS throttle valves:.
.l . Within 4 hours following completion of each valve stroking ,
.. operation or maintenance.on the valve wnen the ECCS subsysteins are required to be OPERABLE.
i x ! 2. 'At least once per 18 months.. Chogin3 --Oe .. Injection Safety Injection Cold Safety Injection Hot cpomp Throttle Valves Leo Throttle Valves Lea Throttle Valves i 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 <
I SEQUOYAH - UNIT 2 3/4 5-7
p, - c < ., . 2 .a c- ..- - -EMERGENCY CORE COOLING SYSTEMS' 2 3/4.5.4 6 DELETE D j O '[
.BORONINJECTIONTANK/-
LIMITING COND FOR' OPERATION ,
- 1 1 b' 3.5 .l_ The boron inject n tank shall be OPER with:
'a. A minimum ntained borated wate olume of 900 gallo ,
- b. Ab n concentration of b een 20,000 and 22, ppm, and' ;
c .- minimum solution t erature of 145'F. AP CABILITY: ' MODES 1 nd 3. TION: c,- ~~
. With the boron njection tank inop able, restore the tank o OPERABLE status within I ho or be in HOT STAN anc borated to a SHU WN MARGIN equivalent to 1% de k/k at 200*F wit n the next 6 hours; re ore the tank to OPERAB statu ithin the next 7 s or be in HOT SHUTO within the next 12 h s.
J SURVEIL CE REQUIREMENTS
- - 1 1 _.
;
4.5.4.1 The_ boron jection tank shall demonstrated OPER by:
, .a. Ve ying the contained rated water volume least once per days, ,
, Verifying the ron concentration the water in the ta at least once per 7 ys, and
- c. Ver ing the water temp ture at least once p 24 hours.
D (' SEQUOYAH - UNIT 2 3/4 5-11 e
=*. *-
7
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it
- EMERGENCY' CORE COOLING SYSTEMS I
De,l ete. L
", HFAT TRACING O '
LIMIT!NG CONDI . N FOR OPERATION , , 1 , 1 I3.5. 2 At least two-ind ndent channels of h t tracing shall be RABLE f the boron injection ank and for the hea raced portions of associ-ated flow paths. APPLICABILITY: DES 1, 2 and 3. ACTION: With ly one channel of t_ tracing on eithe he boron injectio ank or on t heat traced portio f an associated f1 path OPERABLE, op tion may ontinue for up to 3 days provided the + k and flow path t eratures are veriffeo to be gr er than or equal t 45'F at least one per 8 hours;
'otherwise, be i at least HOT STAND within 6 hours an in HOT SHUTOOWN within the f , owing 6 hours.
\
SURVEILLAN REQUIREMENTS
/ / / ,
4 . 4. 2 Each heat cing enannel for t' boron injection t and associat flow path shall b demonstrated OPERA .:
- a. A east once per 31 s by energizing ea heat tracing nnel, nd .
. At least once r 24 hours by veri ing the tank a flow path temperatur to be graater than r equal to 145* . The tank temperat e shall be determi by measuremen The flow path tempe ute shall be deter ned by either surement or recirc -
ti -flow until estabi ment of equilib um temperatures wit n the nk.
^~
.v SEQUOYAH - UNIT 2 3/4 5-12
, ga EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK
}. t "
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 ppm of boron, R2
- c. A minimum. solution. temperature of 60'F, and
- d. A maximum solution temperature of 105'F.
. APPLICABILITY: ' MODES'1, 2, 3 and 4. '
ACTION: With the RWST-inoperable, restore-the-tank to OPERABLE status within 1 hour or be following-30 in at leasthours.. HOT STANDBY within 6 hours' and in COLD SHUTDOWN within the l 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 hours by verifying the RWST temperature.
l i=
- g. Amendment 2 I
. -- SEQUOYAH - UNIT 2 3/4 5-13 9/15/81
i,-
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- " ~. - j
,3 REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) prsvide 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 capability recuirement occurs at EOL from full power equilibrium xenon 604 conditions anc requires"-H9B gallons of 20,000 ppm borated water from the : boric acid storage tanks or 0',100 gallons of ppm borated water from the refueling water storags tank- 2500 Egg With the RCS temperature below 200'F, one injection system is acceptable 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 changes in the event the single injection system becomes inoperable. The boron capability required below 200 F is sufficient to provide a SHUTDOWN 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 of.2000 ppm borated water from the refueling water storage tank.
g
/ ~ 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
' 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 iodine andsystems mechanical minimizes the effect of chloride and caustic stress corrosion on and components.
The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. . 1 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-teined,-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 l i I r (L ! 1 SEQUOYAH - UNIT 2 B 3/4 1-3 Revised 08/18/87 Bases Change m__ _ _ __ _ Q _ - :- - - -
M:, q;
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3 .:
' EMERGENCY CORE COOLING SYSTEMS (4
- BASES t
e 'ECCS SUBSYSTEMS (Continued) i The. Surveillance Requirements provided to ensure 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 maintainpd 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 acceptable 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 Ddh The OPE LITY of the oninjection- tem as part of e ECCS ensures that suff ent negative ctivity is in ted i.nto the c to counteract any p ive increase reactivity ca d by RCS syste ooldown. RCS down ca e caused by i vertent depres ization, a los f-coolant accid or a {% eam line rupt . ( The
- its on injecti tank minimum c ained volume and oron concen -
tion e re that the a mptions used in e steam line bre analysis ar met The contained ter volume limi neludes an allow e for water ot L ble because o ank discharge li location or oth physical ch cteristics. The 0 . ABILITY of tlie r undant heat traci channels as ciated with p'. 'the'bor injectionsyste nsure that the so ilit oron solutio will e maintained abov he solubility li of135yofth Fa 1,000 ppm bor . .
- 3/4.5.5 REFUELING WATER STORAGE TANK 1
l The OPERABILITY 'of the refueling water storage tank (RWST) as part of i l- 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 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 A L SEQUOYAH - UNIT 2 8 3/4 5-2 L l
; q. 4
;
EMERGENCY CORE COOLING SYSTEMS-BASES l
'l REFUELING WATER STORAGE TANK (Continued) l 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 BR
- also ensure a pH value of.between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of b- iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
. Add -
Adddien aI13 ,1be OPERABILITY oP 4he RNST' as part of 4h e ECC3 ensares tha-l evN?cjerv{ ne a+ive reac+ivol is injec+ed in% +he core to coun rad on3 pcsi 've, increase in reac+ivi43 caused by RCS sys+em cooldorvo.
.y i
L l SEQUOYAH - UNIT 2 B 3/4 5-3 Revised 08/18/87
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T ENCLOSURE 2
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PROPOSED TECHNICAL SPECIFICATION CHANGE
- i SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2~ ,
DOCKET NOS. 50-327 AND 50-328
,(TVA-SQN-TS-89-26) s v . s b
, DESCRIPTION AND JUSTIFICATION FOR ,
BORON INJECTION TANK DEACTIVATION' . l .
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ENCLOSURE 2 [
~ Description of Change-
. , -Tennessee Valley' Authority proposes to modify the Sequoyah' Nuclear Plant U
(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 Y the boron concentration of the refueling water storage tank will be changedLi n.LCO 3.1.2.6. In Surveillance Requirement 4.5.2.g.2, the 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 injection system are being deleted. LCO 3.5.1.1 will be revised with a-
-new boron concentration for the cold leg injection accumulators, and1 k , LCOL3.5.5 will be revised with a new boron concentration for the refueling water storage tank.
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 maintain the tank and associated piping at a sufficiently high temperature , 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 heaters, and the potential for-low-temperature alarm actuation and heater burnout exists. Violation of the TS on concentration in the boron L 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 injection tank fluid into the reactor coolant system) time consuming and costly. These potential difficulties unfavorably affecting plant availability, R operability, and maintainability can be drastically reduced in severity or eliminated by the boron injection tank deactivation.
I' Q n a;, , _' r
] i J
y -2 it, n - Justification for' Change :
. l The only accident = analyses that'are significantly affected by boron )
e reduction, boron injection. tank removal, or bypassing are the steamline l b ' break transients. .These transients are affected with' respect to both core l r integrity and mass and energy release to containment. i The'fo11owing steamline break cases were considered-in the core integrity 2 analysis for SQN l (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 i double-ended rupture of a steampire downstream of the flow restrictor (1.4 j 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 analyzed;: uniform refers to an equal blowdown from all four steam 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 requirements 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 'I failures are anticipated. Th'e 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
.11mits' 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.
For 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 cases 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.
i
(F:
. ' ?
.r. s The mass and energy analysis considered two cases: (1) large or l
- 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.
C ' Assuming the boron injection tank-remains installed, without heat tracing, and with the boric acid ~ concentration reduced to zero ppm, the itemperatures and pressures reached in the small 'areak calculations fall ,
.below the containment design limits. e i / Ir. creasing the refueling water' storage tank boron concentration is !
proposed to address the future need (beyond Cycle' 4) for a boron
-concentration increase, which was identified when the Cycle 4 reload safety. evaluations were performed. In fact, the Unit 2 Cycle 4 reload safety evaluation stipulated that the boron injection tank needed to remain in operation 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 the fuel cycles-(in the process of goits 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 j
60 to 68). ! 1' ( In-performing this_ evaluation, the strategy employed was to select the L L highest baron concentration possible that would acconsnodate the removal of L the. boron injection tank (approximately 55 ppm), accommodate 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 l: maximum margin available for future fuel reloads. ' The evaluations performed to support boron injection tank deactivation accommcdate the effects from the following modifications planned for the ; Cycle 4 outages'for each unit:
- 1. Resistance temperature detector bypass elimination
- 2. Eagle 21 digital protection system implementation
- 3. Upper head injection removal
- 4. Vantage 5. Hybrid fuel implementation 5.' Use of new steamline break protection
- 6. Reactor trip on steam flow / feed flow mismatch elimination l
.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 cold ~1eg accumulators. This increase is necessary to meet the boron requirements in the postaccident sump. Also, to meet the increased boron requirements associated with future core reloads, the volume of the boric L acid storage system will increase.
1
-4 y [nvironmental Impact Evaluation
_ 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 it, crease in any adverse environmental impact "_ previously evaluated in the Final Environmental Statement (TES) as modified by the Staff's testimony to the Atomic Safety and Licensing Board, supplements to the TES, 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. }-
i r ME I E km - l0 _ == , Er , 4I s d[ C
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ENCLOSURE 3 PROPOSED TECENICAL SPECIFICATION CHANGE 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 F 1 J, . I l I
14 h l J ENCLOSURE 3 1 Significant Hazards Evaluation t i TVA har evaluated the proposed TS change and has determined that it does not represent a significant hasards consideration based on criteria , established in 10 CTR 50.92(c). Operation of SQN in accordance with the proposed amendment will nott I (1) Involve a significant increase in the probability or consequences of
-an accident previously evaluated.
The deactivation of the boron injection tank affects the steamline break transients with respect to core integrity and mass and energy l release to containment. With the assumption that the boron injection tank remains installed without heat tracing and with boric acid > concentration reduced to sero ppm, analyses show that the departure , from nucleate boiling design basis is met and no consequential fuel 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 analysed. The boron injection tank is a component of the safety injection system whose sole function is to provide concentrated beric 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 accident. (3) Involve a algnificant rzduction in a margin of safecy. I The analyses parformed for the deactivetion of the'coron injection tank indicate that the departure from nucleate boiling design basis continues to be met. Additionally, the temperatures and pressures reached to contairnent wr,ulo f all below the containment desi6 n limi tr.. Sinae the design baces contain the. required margins of safety, no significant reductiene in margina of safety will occur.
i
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- l. ENCLOSURE 4-Final Safety Analysis Report Chapter 15 Analyses Expected Changes p
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Y I l-l y , - - . , -. _ _________________m__m ____ _
I DCN No.mo6M A ! SQNWS 15 2./3.l Page
\
l 3 The steam release as a consequence of this accident results in an initial increase in st'eam flow whicle decreases during the accident. as the steam l
.C. pressure falls. The energy removal from the Reactor Coolant System (RCS) causes a reduction of coolant temperature and pressure. In the presence ;
of a negative moderator temperature coefficient, the cooldown results in ) a reduction of core shutdown margin. ! i The analysis is rf r de nstrate _that_thelollowj;gLgrJttrion is
~~
satisfied: ssuming a stuck r uster c'ontrol' assembly and a' sing lttvist. as
'h'# "
f 11 r i the Engineered Safety Features,t5:r: d 5: n: 7:t r- t: *
/'Mty after reactor trip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief or safety valve. w h hwt desgn b=.sts wu\ he. mt*
= _ - __ _ w ;
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 steam lines. ;
- 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 feedsater pumps, and close the feedwater isolation valves.
l l $.L 1'J .2 &nalysi u f Effects ami Conseaueuefy, M od of tnalystys. The following antlyses of a secondary system steam release are performed ' for this section. e mat. p os:rms 4 *t 5 40
- 1. A full plant digital com'puter sin.ulation. ,Rf fereace 4 coda '
togtermine PCS temperature ""'"*** and prestun M U . d E SA k sts Ir. m e.t.% .ggd wo l SNIAt!on 1 2. n ne'y:M to determine that the reactor e:: et reter :ritic;;,
% %s,~
The following conditions are assumed to exist at the time of a secondary system break accident. l- ., 15.2-39 COC4/Oll5F . W W __m_ _ _ _ _ _ _ _ - - -- _ _ - - - -
g ' SQN-5
- 1. I:nd of life shutdown margin at no load, equilibrium menon j conditions, and with the most reactive assembly stuck in its fully p withdrawn position. Operation of rod cluster control assembly banks u -
during core burnup is restricted in such a way that addition of positive reactivity in a secondary system break. accident will not lead to a more adverse condition than the case analyzed.'
- N
{-
. Z h
- 2. A, negative moderator coefficient corresponding to the end of Ilfe rodded core with the most reactive rod cluster control assembly in 8C the fully withdrawn position. The variation of the coefficient with temperature and pressure is included. The Keff versus temperature
~ at 1000 pst corresponding to the negative moderator temperature coefficient used plus the Doppler temperature effect, is shown in Figure 15.2.13-1.
3, Minimum capability for injection of high concentration boric acid solutton corresponding to the most restrictive single fatture.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 chariing ump livering gifull contents the_ cold
~
e Kevloe c4 cl e as eth taken Tor the Iow onc ntratlon torlc acid which EWn must be swept from the safety injection lines downstream of the RwsT. M en ' eject'e- tM 'te'et'e= ve' vet prior to the delivery of high concentration boric acid _-(g ppm) to the reactor coolant loops.
= w-
- 4. The case studied is an initial total steam flow of 228 lbs/second at 1015 psia from all steam generators with offsite power available. -
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.reacter be just critical or operating at power at the time of a stem roiekse, the re6ctor w)l1 te tripped by the normal overpower prMeetton when power icvel resches a trio point.
. Following a trip at power the RCS contains m re stored energy than ,
at no load, the average ecolaret tempertture is higlier thar. at no load anG-there-is-apprerdubie energy stored in the fuel. l Thusc -the-additional stored energy' is removed via the cooldown
- -
cour.ed by the steam line break before the no load conditions of RCS L .are reached. After the addition &l stored energy is removed, cooIdevn procee'is in the same war 3:gr as la the analytty which
- assumes no load condition at f.me zoro. However, sinte the initial a 5
steam generater water inventory is greatest at no load, the magnitude and duration of the RCS cooldown are less for steam line
- breaks occurring at power.
l'
- 5. In computing the steam flow the Moody Curve for fL/0 = 0 is used. .
1 15.2-40 COC4/0115F l m o D
.g. .
t 3 The mass and energy analysis consideret two cases: (1) large or 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, and with the boric acid concentration reduced to sero ppm, the temperatures and pressures reached in the small break calculations fall below the containment design limits. Increasing the refueling water storage tank boron concentration is proposed to address the future need (beyond Cycle 4) for a boron . concentration increase, which was identified when the Cycle 4 reload safety evaluations were performed. In fact, the Unit 2 Cycle 4 reload safety evaluation stipulated that the boron injection tank needed to remain in operation 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 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 accormadate the removal of the boron injection tank (approximately 55 ppm), acccamodate 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 avaluations performed to support boron injection tank deactivation .
accommodate the effects f rom the fo210wir.g modifications planned for the Cycle 4 outages for each unitt
- 1. Resistance temperature detector bypass elimination
- 2. Eagle 21 digital protection system implementation
- 3. Upper head injection removat
- 4. -Vantage 5 Eybrid fuel implementation
- 5. Use. of new steamline break protection x 6. Reactor trip oc steam flow / feed flov mismatch elimination
. In saranary, plant specific analyses have been performed for SQN's steaulino 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 maxin*m boron concentrations for both the refueling water storage tanks and the cold leg accumulators. This increase is necessary to meet the boron l requirements in the postaccident sump. Also, to meet the increased boron requirements associated with future core reloads, the volume of the boric i acid storage system will increase. j 4
1 l
, . t Environmental Impact Evaluation 1
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 environmental impact previously evaluated in the Final Environmental Statement (FES) as
!. modified by the Staff's testimony to the Atomic Safety and Licensing :
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.
i 3. Result in netters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact. I k 6 P { l i ll
- e. ,
I
+
fg.; O , ( 4
; I ENCLOSURE 3 ,
;
l PROPOSED TECENICAL SPECIFICATION CHANGE t SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-26) f DETERMINATION OF NO SIGNIFICANT EAZARDS CONSIDERATIONS l a 1 P t A p
~o i V, .
s
;- ENCLOSURE 3 ) ,
Significant Hazards Evaluation
;
TVA has evaluated the proposed TS change and has determined that it does not represent a significant hazards consideration besed on criteria established in 10 CTR 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. The deactivation of the boron injection tank affects the steamline 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 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 coolent to mitigate the consequences of postulated steamlint break analysis. The.deactivatien of the borou injection f.ank will.tbetefore affect the steamline break trantients, but it will net create the possibility of a new or cifferent type of accident.
(3) Involve a significant reduction in a margin of cafety. The analyses perferntd for the deactivation of the Snron injection tank inCicate that the departute from nucleats boffing design basis continues to be met. Additionnily, the temperatures t.ud pressures reached fu containment would fall below the containment design limits, Since the design bases contain the required margins of safety, no significant reductions in margins of safety will occur. l
-- ,y -
.o :
- 4.; ,.k^
t t ENCLOSURE 4 Final Safety Analysis Report
.,,. Chapter 15 !
M Analyses Expected Changes t k i
+
'?
v t t t t 9 i 9 6 E 4
.s 1
1 i h
, , _ _ . _ ,~ . _ . . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _
L ,- ; S0W-5 DCN No.mo'5W fg, g , a , j l Page . .
< The steam release as a consequence of this accident results in an initial :
increase in st'eam flow which decreases during th.e accident.as the steam '
. pressure falls. The energy removal from the Reactor Coolant System (RCS) causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin. .
The analysis is rf r demonstrate lhat_thelol_1herittrion is
~
i satisfied: ssuming a stuck to ust'er control' assembly and a' sing Rtvlse as
' h*"' "
f 11 r i the Engineered Safety Features,ther: et 5: : r:t r- t:
- 1
/* M 'ty after reactor trip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief or safety valve. g h be denn hsis wm be. mt+ ,
The following systems provide the necessary protection against an accidental depressurization of the main steam system.
- 1. Safety Injection System actuation from any of the following: ,
- a. Two-out-of-three low pressurizer pressure,
- b. High differential pressure signals between steam lines.
- 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 centrol action which will close the main feedwate' ydives following reactor trip a safety injection signal will rapidly close all feedwater control velves, trip the main L feedaater pumps, and close the feedwater isolation valve.s. .
4 15.2.13.2 Analysis of Cifects and Constauences Method c.f Analysis The following analyses of a secondary system steam release are performed ! for.this section. Rev%t
%5 40 orius 4 1
- 1. A full plant digital com'puter simulation,4%RVR Reference -N code, ' i to e rmine_RCS t g bNBR design boMS is mt.t. as MM) evaluodion
- 2. n :nd y: M to determine that the reactor d^et net r:t;r critic:'- 5 The following conditions are assumed to exist at the time of a secondary tystem break accident.
1 15.2-39 COC4/0115F . O W I g FIe 4 4g
, V SQN-5 j
.. _. ; i 3 6
- 1. End of life shutdown margin at.no load, equilibrium xenon i 1 i conditions, and with the most reactive assembly stuck in its fully r )
withdrawn position. Operation of rod cluster control assembly banks 0 - i during core burnup is restricted in such a way that addition of positive reactivity in a secondary system break accident will not { .. ; lead to a more adverse condition than the case analyzed. *
- N l
2._ A, negative moderator coefficient corresponding to the end of life zh 8a i rodded core with the most reactive rod cluster control assembly in 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. ;
' 3, Minimum capability for injection of high concentration boric acid ;
solution corresponding to the most restrictive single failure in the i Safety Injection System. The injection curve assumed is shown in Figure 15.2.13-2.- This corresponds to the flow delivered by one charging ump livering gifull contents the_ col d er kevige a5 , ce as eth taken Tor tne Tow onc ntration torlc acid which EWn : must be swept from the safety injection lines downstream of the RwsT 50 0" '-j?ct'ea t!" '!0'atie- "!'"?! prior to the delivery of high : 0 ppm) to the reactor coolant loops. # concentration
-- - - boric acid CO
^^I ndc __ _,___ _ _
- 4. The case studied is an initial total steam flow of 228 lbs/second at 1015 psia from all steam generators with offsite power available.
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. , Should the. reactor be just critical or operating at power at the l 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 -the-addittoral stored encrgy is removed via the cooldown caused by the steam line break before the no load conditions of RCS are rear.hed, After thz additional stored entrgy is removed,
'cooldown 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 magnitude and du, ration of the RCS cooldown are less for steam line breaks occurring at power.
- 5. In computing the steam flow the Moody Curve for fL/D = 0 is used.
1 15.2-40 COC4/Oll5F fn M *
$L g,
i SON .
- 6. Perfect ,motsture separation in the steam generator is assumed. 4 f
,-C m .
- 7. The upper head injection system (UHI) 1$ 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 I , effect is compensated by boration provided by the UHI, z 8 a.b' -
. Results ;
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 ; release having an initial steam flow of 228 lbs/second at 1015 psia with steam release from one safety valve. The assumed steam release is the ' maximum capacity of any single steam dum' p or safety valve. Safety Injection is conservatively assumed to be initiated by low pressurizer pressure although steam line differential pressure would provide a more c d d g' ilufficierit ne ative reactivity t 31&lrLAhtj^"tr y"'!):]=
- gD"E c'tt n !'t).
e Tesctm sy tranTlent for the case snowT) Tn7 igure
.. - more severe than that of a failed steam generator safety or relief valve which is terminated by steam line differential pressure or a ~
C 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 p metal.other than that of the fuel alements or the energy stored in the
! other steam generators. Since the sient occt:rs over a period of - aoove five minutes, the neglected t g s %% likely Devua ; signincant effect in slowing the e n e 'l - j "^QmWmom btGp. Aoe 1ht \leH value,
. 15.2.13.0 Conlusions -
- .~~ % A m .:
The' analysis W M cL t p :f[jg y gg y gj {{j g ,jg d"" N %Q A j #
- _ f.Q{ m
-=w
'l*m*ES 'tllafteIMMe'~+e'IIWGeWo'cWsiivMal dwee m 4 wre oe Re.S occsc )
. u e,n ,
Ahrn % 4.w<m w /s so n.wm rar B en.a % p.e .,v % cN'*ISt AS l 15.2. a. i ideniiHttriivn vi Cesca.,d Accidenu vescriph 8h
;
Spurious SIS operation at power could be caused by operator error or a . ' false electrical actuating signal. A spurious signal in any of the
'following channels could cause this incident.
Ca.i,c.a, ~ Q - g - > u l 15.2-41 C0C4/0115F .
. , . i e.A
SON-6
- o ;
- 1. High containment' pressure Qg g,ptn5 M l
- 2. Low pressurizer pressure
, f9'- .
)
- 3. High steam line differential pressure .'-
i 4. High steam line flow coincident with either low average coolant temperature or low steam line pressure. , 1 Following the actuation signal, the suction of the centrifugal charging 6 pumps is divertad_fr&m the volumelgontrolo tho refueltna water st_orace tank.f The vaTve^s isolating The h inject' on tank 44W- from the~ charging pumps and it.e ehe; i;;htin; tM *:T fra the injection i header then automatically open . The charging pumps then f:::: H;MyP d e. M57 N e+nsen444*d ppe) 50-!: Scid te'"tio- ' 0- t'e "P. through the [ / header and on line and inlo the cold legs of each loom.I ThF safetyin7ec on pumpraMi-hutomancaily tiu't' provide no flow when(Ref# se as the RCS is at normal pressure. The passive injection system and the low 5 he ut g; 3 9 y,,;M: 1 % finw at- y ==1 nc? p;;;M hsn ww'@ % 6G 3GUIM M.- . .;n:. meg',2%rs n; ; .%%iDIY.
"9 "'#** * *
;;;e o. in a reactor trip followed by a turbine
* #'" ' "* D 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.
Cate A Trip occurs at the same time spurious injection starts Case B The reactor protection system produces a trip later in the [' I transient. A 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 definite fault. The operator must also determire if the safety injection system Rust be defeated for repair. For the former case t.'te operator would stop the safety injection and bring the plant to the not shutdown conditions. If the safety injection system must be disabled for repair, boration shoJid centinue through the normal boration mode and the plant
- brou)ht 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 m 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-hydraulle 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. 15.2-42 COC4/Oll5F e e s ,e g 4
7 .. j
,, . u-::
The' time to' trip is affected by initla) operating conditions including 5 l
" core burnup history which affects initial boron concentration, rate of change of boron' concentration. Doppler and moderator coefficients.
[ I l f ; 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 % m
;
pressure associated with the power unbalance during the transient. The i { time at which reactor trip occurs is of no concern for this accident. At ) lower loads coolant contraction will be slower, resulting in a longer E time.to trip. , y 15.2.14.2 Analysis of Effects and Consecuences f Method of Analysis The spurious operation of the SIS system is analyzed by employing the dettiled 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 several cases.shows the results are relatively independent of time to trip. A transient is presented representing conditions at beginning of core life. Results at enri of life are similar except that moderator feedback i effects result,in a s'over transient. The assumptions art: L 1. Initial Operirting Conditions - the initial nactor power 'and Reactor Coolant System tergeratures art 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 Ilfe 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.
4 Pressurizer Heaters - Pressurizer heaters were assumed to be ' nonoperable in order to increase the rate of pressure drop. ema 3 m . - --- % _ _ - ,m % m sw.3
- 5. Coren Injection At time gero two charging pumps inject 20 0 c o Prm /
borated er into the cold legs of each loop. E w.n tr % k SlT G
%.ma & % m ,% apa.
W h a % \.miet . > m.~ m a.- a , 15.2-43 COC4/Oll5F 9 .D 4 9 4
J.- 4 Q . SCN-E TANLE 15.1.1-2 (Sheet 2) (Continued) StamARY OF INITIAL ColEILIONS Ano ConruTER CODES * - INITIAL N555 } REACffv!TY COEFF!CIENTS THEstat POWER OUTPUT ASSUMED ASSUMED . 8 N00ERATOR"'M00ERATOR ' , COMPUTER TEh.YRATURC DENSITY FAULTS CODES UTILIZED fAkf'Fl (Ah/am/cc1 DOPPLEnf21 tredT) CONDITION II (Continued) Loss of Normal StK0UT - -- sea
- NA 3577 Feedwater Loss of Off-Site BLkOUT - $1A pea 3423 Power to the -
i Plant Aemiliaries (Plant Blachowt) . ! E=cessive Heat MARVEL . - 8.43 to er e and 3423 1 Removal Ove to Feedwater System Malfunctions . 1 Escessive toad LOFTRAN - 8 and 0.43 tower - 3423 i Increase Accidental Depres- LOFTRAN - 0 Upper 3423 surization of the Reactor Coolant . System wtg EveSE Accidental Depres- -MAINE 4- - Function of -3.3-pcm/PF G surization of the to):TRAg Moderator -23 (Svi,cri tical) Main Steam System Density
- , See Schsection
, 15.2.13 (Figure 15.2.13-1)
Inadvertent Operation l of ECCS During t0ffRAN 0 Lower 3423 o8 c) z Po.er Operation
- 7 9
3
- b i w l- w ,
3 -
. 0%3F/C004 A l
,_ - , .. - ~m, , . . , . . , ._ .- . , . ,, . , , _ . . . . . , . _,__..____m _ _ _ _ _ _ _ . _ _ _ _ . ._
..c ._ .
= n ~4 e* . .-
-e ,
e .
~\ .
SQas-4
;< . .
TaSLE 15.1.2-2 (Sket 4)
-l ,
(Continued)
, itsemay of 1NtitAL fGMDITIONS Aas COMpuTEE CGDES * .
I, . . ,
.fMITIAL NS$$ i <. *. i
- HDCTIVITY C0ffFICIENTS THEmmt POWE$ GUfruf t
., A55UMrb ASSUMED
- N00ERATORh0GCRATOR"'
j' COMPUTER - TEMPERATURE DENSIff. fAk/'F1 1AL/en/ccl 90PPLERf21 feedfl FAutf5 CODES UTILIZEB
't s**tsE vise ~
- CONDITION IV (Continued)
Function of
- 4.3. pcse/F G
- Major secondary N HINC y,q (Critical) system pipe rvP- Moderator-LcFTRAN - Density See tore up to and
. including double- 15.4.2 (Figure -
ended rupture 15.4.2-1) * (Rupture of a Steam , Pipe) NA NA NA E 3577 Steam Generator
.. Tube Ropture 0 Upper , 2396 and 3423 i , Single Reactor PH0ENIX. LOFTRAN ,
Coolant Pump THINC. TACIRAN - I l ,. Locked Rotor .e l NA NA 3577 Feel Handling
- NA Accident I
, I I
-l ,ecm/'F BOL - Consisent 8 and 3423 Ampture of a Con- TWINKLE. FACTRAN I
LEOPARD -26 pcu/*F 80L - with lower
. trol Rod Mechanism limit shown ,
Housing (RECA -Figure 15.1.6-1 '
- Ejection) .
Notes: 4 a e Only one is used in an analysis i.e. either moderator temperature er moderator density coef ficient. yg l' it) O
.Or lo Zs (2) Reference Figure 15.1.6-1 -
I l . . . 4
- JE I* 5 l 0563F/COC4 ,
l ,. M i q W .
- 'w er ** tr _
#' s*-- *v er sN *r ir-= W W 1 "h"'.4 vv= m~ w 'e -.soov--+- - - - -ww -w-- .,-
- e v e- g m-- -
c l
. 6937 97 T i
. (; WIEeue 6 i
l.06 / l k z& - 8 s; , R0 POWER. 1000 PSI th0 Of LIFE RODDED 1 . l.Os - cDat witu Out Rcc Stutt IULL OUT l
. i 1.04 -
= 1.03 5
.G f .02 -
< g .
y1.01 - T
- l. - .
r i 0.99 - s 0.98 2d 30o y(0 we 500 ss0 - RE AVERAGF TEMPERA '('F) (
., Fe'gure 15.2.13-1 Variation of Kgyp with Core Temreurs l .
1 - t *-~ , - - ~ - - - - - . . _ _ . - - - - - - . .
.- ;
. .j
$$N _ _ . l DCN No.MD'03
- i Page -
l r I.06 l Itno PowEn. iooo esta 1.05 - E"' LIFE 8880E8 ! CORE WITN ONE RCCA STUCK FULL OUT .3 i 1.04 -
;
,- ' m 1.03 . . .-
5. C. [ - E 1.02 - t t . d -
;: - '
. y1.01 -
s l.00 -
- i a s i
0.99 -- 0.98 i ! ! 250 300 350 400 450 500 550 CORE AVERAGE TEMPERATURE ('F) Figure 15.2.13-1 Variacion of K,ff vich core Temperature er e e e
- e
. ~ .s. - - . -- -,-.--..---v. . - - . , _ . . - - . . - _ _ . , - - . . , , . , - - - . -
= _ . _. - . . _ . _ .
'4
.' .*(',f'l'J.'t8* " I' Page. -
- .
_ / / 1800 - 4 1600 -
/
O ~
/
/
Z
!a ,- - / ,i
/
(- (Eg i0m . .
/
),/" . .
3 - 200 - f l l i l # f l 0 0 100 200 300 WO 500 600 706 800 SAFETY INJECTI, FLOW (GPM) Figure 15.2.13-2 Safety injection Curve e
* ;
. 1 59N iP!
' 2600 b cc W
s k i-2400 i E J z i 8c[.f 3 2200 j t 2000' - t,
- 1800 !
1600 . . l 5 ! E
~
w 1400 - R m
.y' 1200 -
E l 1000. - t 800 1 i 1
- 600
i'n - i 400 200 > 1 O O 100 200 300 400 500 600 700 800 i SAFETY INJECTION FLOW (GPM)
, 1 i . Figure 15.2.23-1 Safety. Injection Curve <
l
.-___.m...m,, , , -__.,,-,,_-3
1C e pht6 ,
/ -7635-35 b .
==0
/ /. / I*
.s
/ l/ I / l I
\~
1 , 5 Pit $$piZtt (Wilts AT 160 SECONDS p/ . 4
- _ 20,000 PPN 90 ROM Rt NES LOOPS AT 197 afCONDS . _
l , i 2000 L '
<t / .
M 00 - l 7
) / )
0
\
600
/ / / l -
g 20 / . (- l' e, MO
, v ,00 -
< I -
> " / /' / / L g, / / / / 4
- ! N
-5 -
/ '
/
/ l l l l /
.. -5.0 _
0 100 200 300 400 500 TlHE(SECONDS)
/
-Figure 15.2.13-3 Transient Response for o Steam Line Break Equivalent to 228 Lbs
' (' Avelleble/5ec et 1015 PSIA with Outside Power l
'?
~
e
. . . . 3 si%9 aQN
- m. ..
. . \
\
\ .
j
.25000 , y !
'C E c
22500 < - 20000 - -
.11500 - -
b
+
$r g
i- .15000 < - .. N
*% .12500 < - .. S Iw .10000 < - .. .
Og g *i
.07500 - - ..
l L Qg
.w .05000 < - .-
. 02500 < a 0.0 f .
500.0 ' ~ 8250.0 - . g 2000.0 - - ..
, 3g 1750.0 - - .-
1500.0 - - - gE ino.O- - _- 1000.00 - - W ... 750.00 - - 500.00 l - 800.00 ' g $50.00 -% *
. .';
W 500.00 - - C 450.00 - ' W
< 58 ,
$ g 400.00 350.00 -- -
u - 300.00 - - . . .-
. n0. -- 00 , . .- ..
( i
. NI .
2000.0 - - - .. . E 1000.00 - - .. C O. O. - 2* i G< .toso.o - l w .
= -2000. 0 - -
- 5 00.0 O.0 100 200 300 400 500 600 TIME (Sec)
F,3x /r,2./3-3 f!C'JnE :
- TRANSIENT RESPONSE FOR A STEAMLINE BREAK EQUIVALENT n ,f,fg TO 228 LSS./SEC. AT 1015 PSIA WITH OUTSIDE POWER AVAILABLE.
t : i
- g v .. ..- ....:.. ..
i SON.3 L
- m. ' CON h0' H TABLE 15.2-1 (Sheet 6) p TIME SEOUENCE OF EVENTS FOR CONDITION II EVENTS I
L. Accident Event Time (Sec.) M : Excessive Load Increase
- 1. Manual Reactor 10% step load increase O f Control (BOL)'
Equilibrium conditions reached 1 (approximate times only) 200
~2. . Manual Reactor Control (EOL) 10% step load increase 0 l
Equil#brium csndttions reached (approximate times only) 50 .
- 3. Automatic Reactor '
Control (BOL) 10% step load increase O Equilibrium condittens reached (3) (
- 4. Automatic Reactor- .
Control (EOL) 10% step load increase 0 i Equilibrium conditions reached (approximate times only) 50 .
. 1 Accidental depressurization of the Reactor Coolant System Inadvertent Opening of one RCS :
Safety Valve 0 Reactor Trip 29.3 Minimum DNBR occurs 31.5 Accidental depressurization of the Main Steam Safety System. Inadvertent Opening of on.e main steam safety or relief valve 0 ggvise Pressurizer Empties +60 . it7
. h,; boron reaches coke
{^.0^0 o -n_ _ _ _ __ _ t99- 2 57 UHI initiation time ee4 2.61 (3) Old not reach equilibrium within the time' scale of Figure 15.2.11-2 - ' L Revised by Amendment 3 ' 3 C0C4/0723F
g . . SON-3 1 A g0.PUI TABLE 15.2-1 (Sheet 7)- i (Continued)
;-
TIME SEQUENCE OF EVENTS FOR CONDITION 11 EVENTS Accident Event Time (Sec.) 1 Inadvertent Operation of ECCS dur)ng Power Operation Charging pumps begin injecting L borated water 0 Low pressure trip point reached 64 < Rods begin to drop 65
)
/ \
/
. Major Secondar$ Sy5 tem Pip Rupture CondittenIVevent[
j Mie% ud EtplMd W( ff. v.1 -i j
. Case team line ptures 0 i Criticallt attained 18 Pressurizpr empty 15 I
20,000 ppm boron reac. s loops UHI in tlation time 20 16 [- Case b Stea line rupture O Criticality attained 14 / I
. Prssurizerempy 17/
000 ppm bor reache loops 2f HI initiatto time 5.5 bI Case c Steam line uptures 0 Criticall attain 21 Pres'uri s r empty 16 20,000 pm boron reaches loop 30 UHI in tlation ime 17
- 1. Cas d Sten line ru tures
,- . C teality ttained 7 essurizer moty . 18
/
0,000 ppm oron reache loops 32 UHI initi lon time 39 Revised Amendme 3 e , , 3 COC4/0723F
- G
e ' CCN Ho ?
^
SQN-6 {pogo l Fast-acting isolation valves are provided in each steam line that will
' 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
(. completely terminate the blowdown. For any break, in any location, no more than one steam generator would blowdown even if one of the isolation valves falls to close. A description of steam line isolation is included in Chapter 10, 4 Steam flow is measured by monitoring dynamic head in nozzles inside the steam pipes. The nozzles which are of considerably smaller diameter than 1 the _ main steam pipe are located inside the containment near the steam generators and also serve to limit the maximum steam flow for any break further downstream. 15.4.2.1.2 Analysis of Effects tnd Consecuences
, Bethod of Analysis
- The analysts of the steam pipe rupture'his been oerformed to determine:
- 1. .t.udtino frog ThegoolfioM1.owjJig core heat.flun andt e.RCS temperature steam.g' The and prnsurg"/.J'!R-* WTs.A <
Reference g code '4 She
- 2. The thermal and hy'draulle behavior of the con e following a steam line e break, A detailed thermal and hydraulic algital-computer calculation
'(THINC Code, Pa~agraph 4.4.3.1) has been used to determine if DNB 6( !
occurs for the core conditions computed in (1) above. (
- 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, ecullibrium xenon 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 s reactivity in the steam line break accident will not lead to a more ,
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 I steam generator and those associated with the remaining sector were t
, 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 1 conditions cause underprediction of the reactivity feedback in the ' high power region near the stuck rod. To verify the conservatism of l E l 15.4-16 0117F/COC4 1
- c
,< b )
SQN-6 D[NT[h W h.h55 m==
"] .
this methid, the reactivity as well as the power distribution was ' C checked for the statepoints shown on Table 15.4.2-1. .These core analyses considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high , water enthalpy near the stuck RCCA. power redistribution and l nonuniform core inlet temperature effects. For cases in which steam i generation occurs in the high flux regions of the core, the effect of I void' formation was also included. It was determined that the l
, reactivity employed in the kinetics analysis was always larger than- j the' reactivity calculation including the above local effects for all l4 1 ;
statepoints in Table 15.4.2-1. This result verified conservatism, I , i.e., underprediction of negative reactivity feedback from power : L - generation.
" --- Q evise- .l
- 3. Minimum capability for injection of high concentration borte acitiAs stown (approximately i^ ^^^ ppm) solution corresponding to the raest r' 1 1
(restrictive _ sing N 11ure Jine hft in action system / 76e inlictlon curve usTd s hown in figure I.1.. ).-T"TliTI corresponds
'to the ficw deitvered b hithug.41.tyering its full flow ;
olha. cold. itj head o credit has been takeiiTor 'th'e few ~ 4tvise ; concentratTon bT:Iri~c e,rac d which must be swept from the safety M bn j injection lines downstream of the 5: m: inj:: tic :::h i;;i;tica Aws7 i
- E::: prior to the delivery of high concentration boric acid to the ;
reactor, coolant _v loops, p e^% ,
- 4. Four combinations of break sizes and initial plant condttions have been considered in determining the core power RCS transients: -
- a. Complete severance of a pipe outside the containment (downstream of the steam flow measuring nozzle) with the plant initially at no load conditions, full reactor coolant flow with offsite power -
available. . .
- b. Complete severance of a pipe inside the containment at the outlet '
l of the steam generator (upstream of the steam flow measuring g nozzle) with the plant initially at no load conditions with l offsite power available.
. , . .. ~ ~ .~ - -- - ~ - -- ,. s . ,, . .
- c. Case (a) above with loss of offsite power M.:;.t::::e;
_y. with tM L -
- -itiati n Of th: ::f:ty 'nj::ti n ti;ni!
- Loss of offsite powerY'VibC results in coolant pump coastdown. .
- d. Case (b) above with the loss of offsite power,:ti'! tan:::: eith th: initi tica Of th: :sfety 'njectica ti;ael l-For a steamline break inside containment,'with a failure of an MSIV in another steamline to close, the steam generator connected to the
. MSIV will continue to release steam through any lines or valves thet may be open downstream of the MSIVs or upstream of the failed MSIV.
Normally, there are open lines to the main steam reheaters, turbine gland seals, main feedwater pumps, and possibly the turbine-driven auxiliary feedwater pump (steam for the. aux 111ary feedwater pump is drawn from two steamitnes upstream of the MSIVs). During the ly 15.4-17 0117F/COC4 e #
DCN No.100'5M SQN-1 Poge steamline break, steam flow to the main feedwater pump turbines and the main steam reheaters will be terminated. The flow to the c*in . feedwater pump turbines is terminated by stop valves which actuate 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 ceases when the high pressure turbine stop valves close, and the reheaters effectively become a water trap. The remaining steam flow amounts b 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 conservatively assumed to be more than 100,000 lbs/hr. Even with 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 Deak core near f1ur Since the stenaline rupture causea the reactor coolant system to cooldoen. there would be no reason (or signal) for the power-operated relief valves to open. These are fail- l closed valves. Therefore, any postulated salfunction of a power-operated relief valve must be considered an independent failure and inconsistent with a coincident failure anywhere else (MSIVs). The c a s e. of spurious opening of a power-
. ... w.n %- 44. .%v -@c n . .f,a l .1 v 4.r. . e . ,3.4 t s.s. s s e a a l i n e break vith {;
following a large steamline break with subsequent closure of all MSIVs would be less severe than the steamline break case reported in the FSAR. The spurious opening of a secondary system valve, such as .( a power-operated relief valve, i in Section 15.2.13. onsidered separately and reported ( The analyses presented do not ider additional steam blowdown from eld.er of these sources. S: ' leased from open lines and drains on the secondary piping dor significantly affect the analysis results, and the failure ver-operated relief valve is reported separately.
- 5. Power peaking factors corresponding 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 sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steam line 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, pressure q ~nd flow, and, thus, are different for each case
. ..m ~~~~~ m~-m ~
_.n . The4 values u::d for three of the four steamline break accidents analyzed are given in Table 15.4.2-1. The*tbr:: c ::: Orc selected * **3 %% on the basis of hot channel factors, core power, and reactor cool pressure. The fourth cese is iei5 levece cwisiive iv DNBR.r'T T core
/
15.4-18 0117F/COC4
. y _s,
~,
;
s... .- _ _ _ - ~ DCN No.Mo' F SQN Page u i ,A[ Me ss ; arameters esed for each of the theet cases correspond ~ to values- Uc4 ; y determined from the respective transient analysis. T b; ti;; pints - r: :::d fr :::t ::::. / n , j
All the-cases ebeve assume initial hot shutdown conditions at time -
ro since this represents the_most Dessimistic initial e diti_o_nj Shoul0 The reactor 5e just criticaFor o a pow r at the 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 . , 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 additional stored energy has-been removed, the cooldown and L:
reactivity insertions proceed in the same manner as in the analysis which assumes no load condition at' time zero. ,- . Dwever,'since the' initial steam generator water inventory is , 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/D - 0 is used, y , ,
i.
- 7. Perfect moisture separation in the steam generator is assumed. The assumption. leads to conservative results since, in fact, considerable water would be discharged. Water carryover would reduce the
- magnitude of the temperature decrease in the core and the pressure increase in the containment. .
- 8. The Upper Head Injection System (UHI) is simulated. During a design steamline break accident, the reactor coolant system (RCS) pressure
- 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. The UHI flow rates are based on the following model. The pressure
- drop (AP, Ibf/ft') across a component is given by 8
V
- 1. bP = K, o 29 I
15.4-19 Oll?F/COC4 e
- 4
.< y n'
- . s .{. t
- r 2
SQN-4^
- q. - ,
. DCN No.*f '
Where:: Page - - t o' . . Ki -= loss coefficient (dimensionless) g p'.= Ifluid density (Ibm /ft') . V = fluid velocity (ft/second) ,
- g. -- 32.2.lbe-ft/lbf-second' .
' Multiplying- the right-hand side of equation (1) by pA8 /pA' ,
. gives AP =
K.o*V8 A' 4 2P'9eA' .
.s or using = w = pVA for mass-low rate (1bm/sec), the pressure drop becomes AP = w' _ . _
Kap Where Ki is a' geometrical constant (ibm-f t'/lbf-sec'). Solving for w gives the following expression for the UHI system ,. flow rate: , w = K4 PAP The pressure drop used in the model is the difference between the UHI
. gas pressure and the RCS pressure. The density over a given time step is assumed to be constant at the value corresponding to the pressure at the beginning of that time-step. The proportionality constant, K, is an input to the code.
The expansion of nitrogen over a time step is assumed to be
. 1sentropic. The change in nitrogen volume is calculated as:
p Re/6C AS Vn: Vaa. + At , Where Var. 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: P VI - P.VKr. Where y is 1.4 for nitrogen
, .15.4-20 -
~01.17FfC0C4 .
~ *
~
. . _ . , , U *[
go *
;c i[.y'f '
j c ,' DCNtt, m m a
~ '
SQN -
} ,
Pogs '..
. LoFTnAt4
" ;'C Since "'O T' s not used in the analysis of_ LOCA, it does not have to
- 7evnt, e -13 me .high UHI flow rates induced by the severe- depressuri- :
s5 ration of LOCA.; The upper head of.the reactor vessel remains full of SW gbcaptad-water as it receives flow from the UHI accumulator, torTRAt/ + In' MMMik, he boron concentration and enthalpy are' determined in the f] manner:
' X.' , = (NX.At + X..M.,)/(Wat + H..)
Where.X-can be replaced by either H or B., W is the accumulator flow rate and M., is the mass in the dead volume.
. :As stated in NCAP 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~15 compensated for by the boration provided by the ,6 ,
. UNI.- '
i
.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 ,l l during the.steamline rupture. If a steamline is assumed to rupture . l when the plant is in a hot shutdown condition ~and the coolant l temperature in the reactor vessel upper head is assumed to remain at its. Initial no-load value (547'F), then the reactor coolant system L
.would have'to.depressurire to 1020 psia (saturation pressure at 547'F) from 2250 psia before any quality would.be observed in the
-upper head.~ Meanwhile, the UHI system is conservatively assumed to add cold water into the reactor vessel upper head when the primary 1, . system pressure drops below the conservatively assumed 1300 psia 6 set ol g ,v,,, o , y
,o 4 L' he "^"!!' ode calculates the mass and energy of the fluid in the H uppe d based upon the incoming UHI flow and enthalpy, enthalpy 1
and flow from the lower regt r _ofd g grdel), essel', and any heat
-input from the vessel walls. In the F.A u . the UHI water flows into the node represent upW-h6~of the' reactor L 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 Oll?F/COC4 e t
l t
, \
'4 DC;N No.r%dd5 I SQN-6 page, ,_,,_ ,
~
j
;j 3.
..Only' the* Pressurizer water flashes and this. void volume is easily-determined from reported plots of the pressurizer water volume [
C l history. Any heat-input 'to the reactor coolant tends to retard the cooldown resulting from the steamline 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 ' depressurization as well as the cooldown.
- Heat transfer from the-hot walls to the fluid in the upper head and the pressuriter is very small. Both regions are outside the active circulation path of the coolant. The pressurizer is filled with-saturated steam and water. The water remains at the saturation enthalpy as it-flows out of the pressurizer during the steambreak cooldown. The water saturation drops as the siressurizer empties.
The total temperature decline is about 50*F based upon the depressur-
. iration during the~outsurge. On the average, the temperature drop is about 25'T 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 zation. Once the pressurizer is empty, heat (
transfer from- g alls to steam is very poor. (7 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. ' 1
? , . .
- . When UHI is added, the water in the upper head is cooled and remains
! . below the saturation temperature. Heat transfer from the reactor . vessel hea'd is greater in this case, but small compared to the ' cooling'effect of the UHI water, and the heat addition to the primary l
' coolant from the steel walls in the other regions of the primary system which are also neglected in the FSAR analysis. When all these l heat' sources.are considered, the cooldown and consequences of the ]
.steambreak are significantly reduced. 1 The maximum UHI flow rate will occur during a major loss of coolant I e
accident and will amount to about 3000 lbs/second. The UHI flow rate ) L calculations are described in kCAP-847939. The maximum UHI flow rate during a steamline break is a small fraction of the UHI flow rate during a LOCA, rarely exceeding 10 percent of the LOCA flow. The
, average UHI' flow during the first 200 seconds of a steam break is about 50 to 60 Ibs/second. g
_A
/ l l The UBI fl ow rate is based upon the pressure l l
drop b e tw e en the UBI system and the reactor coolant system. I x ~ 15.4-22 011ff/COC4 l 1
~
.c . _ . _ _ _ _ _ _ _ _ _ . -. . _ _ _ _ _ _ _ _ _ _ _ -
~
~ - - - , - - - - -
ce s : gy 7 5;N-4
^
m
- g ,
g
- r- .
The core flow rate is a constant volumetric flow, and any void, if g b
'@" present, would affect the mass flow. rate through changes in the- t .
l 1 p F < -average coolant density. When the reactor. coolant pumps are running, the. core flow rate exceeds the maximum UHI flow rate by more than a Z
'c F factor. of 10. This is assuming the maximum UNI flow rate during a b Q.h; LOCA.- Compared to the average UHI flow during a steam break, the i core flow rate is more than 600 times greater. .,
b' The upper head na yc="htat-pens 6::tm 85 D andi19AE ns10 e pressure assumed in this analysis is , psik., t omed the an and the actual UNI setpoint will~be 100 psi, va e, 3 6s%o f Sens'itivity studies were performed for the Sequoyah Nuclear Plant-to determine the effect-of raising or lowering the UHI setpoint assumed ( in the steam line rupture analysis. A high UHI setpoint results in a relatively early actuation of the UHI system during the reactor %g '
;
WpreTir0M2trfterr oevsHLtly,th0-:td W egture. sQ m eesa m A . tewnmy new serion n -m our. boron The' UHIhei" Whor cool depressurization and therebyYeddce the safety. injection w e t system wpy h delivered,~due to the relatively higher backpressure, net result is- that- slightly higher. peak power levels are attaine following the .
;
return to criticality during a steam line rupture cooldownsth A lowd WT-f"$S*** Therefore. the assumption that the UHI accumulator pressure is at the ' low bl.9h end of' the setpoint range is conservative. C-
,7 =- _ w % _
Results The 're.sults presented are a conservative indication of the events which
~
would occur assuming a steam line rupture since it is postulated that all of the conditions described above occur simultaneously. . Core Power and Reactor Coolant System Transient , Figure 155.2-2showstheRCStransientandcoreheatfluxfollowinga main' steam pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial no load condition-(Case A). The break assumed is ,the largest break which can
' occur anywhere outside the containment either upstream or downstream of the isolation valves. Offsite power is assumed available such that full reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one steam generator. Should the core be critical at near zero power when the rupture occurs the initiation of safety injection by high differential pressure between any steam line and the
' remaining steam lines or by high steam flow signals in coincidence with 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 1 RCS temperature or low steam line pressure. Even with the failure of one. I
. i
.15 . 4-~ 2'3 _ - OI.17 F/COC4
. .. - l
;
~.r..~ ...: . .. L
f 'k
'c, c 4
._ c._,
f DON No..L%.919-
.~ SQN-6 pop
' i
(';
~
valve, relea'se is limlted to no more than 10 seconds for the other steam - generators while~ the one steam generator blows down. The steam Ilne
. isolation valves'are designed to be fully closed in less than 5 seconds .
after receipt of closure signal with no flow through-them. 4 The steam flow on Figures.15.4.2-2 throug4lfiA,QpegatA.stram f-low pler-only/ Ir Odd!t.0n, 2. : ::r ; ::r:t r: Ke vM., - i :::: :::rr d t: .t::h:r;; thr::;h th: 5 ::h 'Or'th r:t !O :::end:. 6
; n. ....... u. 6..
.o :3:._ ; :_:::_ 3,;;;= g_ 3 , _:: [Shwf)'
10 te: nd: !: :::::re:ttv: uith ::p::t t; th; :::: :::tivity tt:::i; t ) - > 2nd th: :::: :nd :n:r;y 7:!::::. Th; 10 ::: nd v:!;; r:p :::nt; ; v:ry 5
'^a; t% d:!:y ':r th: :!;n:! ;:::r:tt:n, t : :rt tt:!, r:::!;t :nd
- Ubtearent !!c:er: c' th: :te:-'in: !::!:tt:n :!v::. The len; t' : l6 ;
d:! ; :::er:: : :::::rv:ttv:!y 1:r;: :::r;y re!::::. , ( 2:r ::::!d:r n; p ::!b!: v:! din; '- th: r:::ter :--!:nt :y:ter :nd '!ee ; i
-b!0ck:; , !:n; ti : d:!:y for :!::ere ^< the :te r'ine !:e!:t!cn v !ve:\
is act aereira ily c^ ate uitive. Ia e* der te che:P #er per:!b!: ve!dia;- j $ ia the pei= ley ca^1 at rytte=, a iteirae bre:M :!?c!:tten es: performed i ia which the'! e!stien vi!v:: eer: :::en d t: :!::: t:: ::: nd: 'r th;r ( -
-thea 10 retend:) #ter the breth Me vetdtn;, :nd th:r:fe : : flow-t'
,- Ib!echt;0 rete!t:d. The de;ree e' tubcaa!'a; 'a the artrary ^^! ant $ - F 6 :ytter ya: net :!;nt::ntly O "::t;d. Th; ;;!;u!:ted cere heet fivx / L teaded te be !cre- The lea; t!-a de!ay '!O ::cend ) !: u::d 5::: :: th:j additical' est: -e!eate har a b!;;e e*fect eet:!de the 4555 th:n the -
=^rly v=1"e'cle:9-e t "a 51: ca the ar% ry ^^1:nt' t:rper:tur:.
As shown in Figure 15.4.2- the core attains criticality with the rod - ( , cluster control assemblies inserted (with the design shut vise
- l one stuck assembly) before boron solution of approximatel 0.0^^
enters the RCS from the Safety Injection System. The dela pm 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
-the.RCS due to. low concentration solution being swept from the safety 6
' injection lines. A peak core power well below the nominal full power value is attained. ,q
.The calculation assumes the 20,0^^ pm boric acid is mixed with, and g diluted by the water flowing e RCS prior to enterirtg the reactor 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 wt,ter density changes is included in the 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 calculat'lon includes the line losses in the system as well as the pump head curve.
15.4-24 0117F/COC4 r - - . - _ _ _ - . _ - _ _ _ - - . _ _ _ _ . _ _ _ - . _ - - - - . _ - _ _ - - - - _ - - - _ _ - . - _ - -
~. . - - . .- - - - _ _ - - . .. -- .
m
+ w ..
~'
gg Ob No. MOM Page The accumulat' ors. provide an additional sop 6*ehr after-4he g
.;
;
RQ peestwredegsag64MW. The .nt:gr t:d flow rat; of N Y.,
.or:t:d :t:r fr a. tt: S:f:ty ::j::ti:n y: tem for each of the four casejs Shown analyzed is shown inJigure 15.4.2-7. y coce becc^ cerw%t}er Figure 15.4.2-3'shows Case B. & 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 i parameters for cases c and d respectively which correspond to the cases g discussed ab-h additioQlou,JL4 fil.tt powor at the A @ m f F afety inlec on s time includes M $ generated. The Safety start the emergency Injection diesel generator System 4nedelay
- 1ev4f.,
o !? ::W: t 1 e R$ c cilft ev ter and the core power increase is slower than ShM A - , in the similar case with offsite power available. The ability of the ! 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 remains well below the nominal full power value. L; s I 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 power this heat is
.C . removed to'the atmosphere via the steam line safety valves which have been sized to cover this condition.
I Generic thermal and stress analyses and subsequent fracture mechanics i analyses of reactor vessels have been performed for 4-Loop plants. These l analyses were applied to a 4-Loop reactor vessel having material properties and end of life (40 years) accumulated fluench similar to the
.Sequoyah vesse.l. The fracture mechanics analysis utilized linear elastic fracture mechanics method in the evaluation of the reactor vessel 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 cooling of a steamline break the operator is instructed to use the intact steam generators for the purpose of removing decay heat L and plant stored energy. This is done by maintaining the stear 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 system
. relief valves. The operator is instructed to terminate auxiliary 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 Oll?F/COC4
2 Q ' .. '. g;.: , . ..
' DCN No. "IMW SQN-6 Pop q
. ( ! can- be met by- simple switch actions by the operators, i.e., closing .
o auxt_11ary feed discharge valves and stopping charging pumps and safety injection pumps. 'Thus, the required simple actions to Ilmit the cooldown l : 1 and depressurization can be easily recognized, planned and performed 1 within ten minutes. For the longer time requirements for decay heat-removal and plant cooldown the operator has time in the order of-hours to respond. t
- 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 be maintained safely at' hot standby conditions for extended periods of time, there is no safety ~ requirement which dictates rapid achievement of cold shutdown conditions.
, With only onsite power available, the plant can be maintained in a safe I hot standby condition using the intact steam generators by supplying feedwater with the auxiliary feedwater system, and venting steam through the secondary side, power-operated relief valves. The relief valves wili l be controlled;to gradually reduce pressure and temperature as the core ,
l residual heat decays. If the relief valves are .not available, the safety I
- valves will be.used for steam dump. In this case, the primary system L
_ pressure would be controlled such that adequate subcooling is i maintained.= Primary system temperature would be maintained at that value_ 6 ( L necessary to lift the steam generator safety valves as necessary*to match I; [
.A the decay heat from the core. This temperature would be approximately 553*F which corresponds to the lowest steam generator safety valve l
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.. 4 The sequence of even'ts is shown in Table 15.4.1-12. ' Marcin to Critical Heat Flux
- l' Past' experience in performing DNB analyses for steamline breaks for W 6 cores has shown that Case B (Inside break with offsiteg mwer u u lm ut wqrgeJhan-E-as%reak ut. tit.off.s.ih nmeri > TM;iadei;che
, M e-- by n i-i-'ng th:.:t: : ;;bts present d n TeMs.15.4.2-i. Cases A
[, ghn and B generally have very similar temperatures-and pressures, but Case l e ewens -to
, 'lHe Power lev Cen:r:!!y, en!y f r Of th: state pointe- . 4.2-1 Tre subjected to detailed nuclear and thermalpresented hydraulic in Ta analysis. For C:::
B, the p0 bt r!th th: 't;h::t perer !:v:1 !! ini!yzad. _awna,4anen hse inA4rstad +hte antn+ 4e
!"ca atit
+ha ann wh4rh util menhsh1u hsua iEileEA55 USSR. 5f555! tid 5[54t5eftsbiabedidae "cEaEUI5gh^Iai
'dep -dtng-on the :end!ti^as) !! ina!yzed.
A A coQete SCT of the. Steamhne. brmk transient' stattpoints 4rt rcVIWo to dNMr@ A e, N5h liml+ig Conch-flor), h/c [/N/;7w% b uite QS
# Shun 15.4-27 0117F/COC4 l .
,s .; .
;;s ' '
( ,
'DCN No.PloiSW
-SQN-6 :
Poge- . , y _ t M ' y;y g;;. g'si.;ng; g_ ., ;3 g _) .; _;;;
"h:!y t: 5::: th: ! n::t O't" i; th: p: rt with th: hign::t ;;;;r"'0wN . 3 r:0! .~ U:: !!;, thi: ;;!-t ?: th: 02: etth the hiaktet powar at wi+h
( C :: ". etther th: pr:::dte; : ::::::d' ; p t-t ': :!:: :::!y::1. i 9 00!d in; ^' the -!-t: : !y::d rete!t <- Du""' :- ' . 20, add ' +i ra a'( p !:t: :y 5: :::!y::: t: ' :;r: th t th: pet-t 9!tb t5: -i-i r- 0""" 2 l tr-d?tt0: h:: 5::r 13:!y::d. j
.a -
. The' points- an d for this application had4DNBRM greater than 1.30.
~
us, iiL is conclude a e. mum unis for ae s
- than 1.30. gg ve a..s shoun The maximum linear heat rate for the most limiting steambreak case i
presented in the FSAR was-less than"M kW/ft, which is less than the linear heat rate which results in fuel melting. Ner: i:em+n : 43:= -) i fallgen marhantem meevistad wt+h &hte am m ir itamme has+ / _ _ _ ~ 15.4.2.2'-Maior RUDture of a-Main Feedwater Ploe 8 15.4.2.2.1. Identification of Causes and Accident Descriotion ' 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 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 preclude the subsequent addition of auxtilary feedwater to the affected L steam generator. (A break upstream of the feedline check valve would (> affect the Nuclear Steam Supply System only as a loss of feedwater. This L case is. covered by the evaluation in Subsection 15.2.8.) Depending upon the size of the break and the plant operating gonditions 'c at the= time of the b'reak, the break could cause either a RCS cooldown (by excessive energy discharge through the break), or a RCS heatup. l Potential RCS cooldown resulting from a secondary pipe rupture is i evaluated in Paragraph 15.4.2.1, " Rupture of a Main Steam Line." 6 I Therefore, only the RCS heatup effects are evaluated for a feedline rupture. l L A feedline' rupture reduces the ability to remove heat generated by the core from the RCS because of the following reasons:
- 1. Feedwater to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase prior to reactor trip;
- 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 e
~ ^ ^
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f , DCN NO,MorSn -..1 A s
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TABLE 15.4.1-12 (Sheet 1) TIME SEQUENCE OF EVENTS FOR CONDITION IV EVENTS [[[h q
. ,7 Wrw Time (Sec)
Accident / Event .) (
- i. / .
Major' Secondary System P)pe Rupture /
- 1. Case a Steam line ruptures 0 i Criticality attained 18 ,
\ Pressurizer empty 15.
20,000 ppm boron reaches /,. loops 20 4 i s
) r .
L 2' Case b Steam line ruptures 0 '\
-// .
Criticality attained Pressurizer empty . 14 17 T
) '
,' 20,000 ppm boron reaches j 4
.( ' loops ,
' 21 (
.- \. .. .. . _,
- 3. ' Case c / Steam 1ine ruptures 0 -
-/ Criticality attained , 21 .
/
f' ,e Pressurizer empty.
-20,000 ppm boron reaches f'
,/ 16 j/ 4-
/
/
/ loops _..
30' s f . 4.
/ /' Steam line rupturus
/ .
t Case d / ,,. 0 Cr.itIcality attained , .' [~- 17 s f'
. Pressurizer empty / e' 19
/.
/- ,/20,000 ppm borop' reaches loops -
32
'4
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.x i
. j. 1 DCN No?'" 0 1
-4 ,Poge - tl
' Major. Secondary
. System Pipe Rupture ,
4
- 1. Case a ,
. Steam line ruptures 0.
Pressurizer empty I 13.4' ']
' UHI initiation time 21 .7 1 E Criticality attained 30.8 j 30.8 Y Boron reaches core -l ll l
, 2..' Case b' Steam line ruptures O Pressurizer empty 15.0-l s Criticality attained 19.8 .I UHI initiation time- 23.0 -)
Boron reaches core 31 .8 )
- 3.- Case c Steam line ruptures 0
' Pressurizer empty- 14.6
. UHI initiation time 24.1 Criticality attained 35.3
*i -
Boron reaches core 47.3
- 4. Case <d_ ,2_
Steam'line ruptures 0
-Pressurizer empty 16.5' Criticality attained 23.3 UHI initiation time 28.4. . 1 Boron reaches core 52.3 Accidental depressurization .
- of the Main Steam System .
. Inadvertent Opening of one q
main steam safety or relief valve O Pressurizer empties 161 Boron reaches core 227' UHI initiation time 237 Criticality attained 305
'. *** e . .
l
,. e ). , ; ,, g
c; .^. . s DCN No. MON A l Page l SQN +
~ne w che\m (cce,~~,mrm.4.2-1 TABLE 15 Shee,t' 1 )
cttd Rep ,
/ -
CORE P RAMETERS U D IN STEA,WBREAK DNB ANALYS{S' .' JI / ( .
's
/ / / Case a. Time Point' .c Parame er a
/ 1
/ 2 3 ,. 5 ,
/ / /'/ '
/ \
l actor Vessel / niet / '
'~
temperature ,tb' sector - ,
/
connected,to affected // '
/ ,' . a Steam Geperator., 'F / -
j436.9 434.6 ,/ 411.4 405.0 399.7 Reac,t VesselIni$t , [ , [ " temperature toire- ,- - . finingsector'*F ,
,/ 492.8 491.1 486.0. 481.3 475.9 m
,- / < , e RCS presstJre, psia _
1077.0 / 1049.8 1023.5 1143.0 l'1117.0 RCSow, f)/%. '100 100 100'/ 100 .. 100l / lriox,x g ge .9i 45 7.22 ,/ 6.83 6.32 41.0/
, Time, sec. 32.5 55.'O 6 5'.'O 75.0-
/ /
/
,e ..p .
/ .s' V .
L . G l . l l' t l l T COC4/0713F'
ev T.. m;
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IDCN No.Mim A _ ' o , , ,Page __.-- l-C; byg - . SQN 7
/
bg
/ .-~~,.,.
TABLE 15.4.2[1 (Sheet 2) ' (Continued) -
-- N '
,/ .
j . . CORE PARAMETERS USED IN STEAM BREAK DNB, ANALYSIS . . i
/ *
/ ,' ,
/ '
Case'b. Time Point ^ Parameter < 1 2 3 4 5 >
/ i r- ReactorV/ essel inlet / j . ') -
temperature to sector / , / connected to affected / \'
'F 382.6,* 368.4 363.5 355.1 351.4 ysteamgenerator,/ -
I
/ReactorVessel, inlet
- g temperature t.o re- -
maining sector, *F 521.8 505.1 497.9- 482.7 475.4
/ / z RCS presrure, psia / 1245.0 1107.8 1.070.6 984.7 954.7 s
/, / .
/
jo
.RCS f.l'w, % '
100 ,f 100 100- 100 100 > Heat flux, % 9.67 / 10.79 10,98 10.3 9.96
/ /
1
/ ime, T sec. ,
l 30.0 f 45.0 52.5 67.5 75.0
/
/ e f ,
i*
/* ,
l-l. l: - 1.. i l' COC4/0713P
T
; , g- - ' '
DCN No. AA E - Page-- 1
~,.-s , . . ~
SON
!L ' belet6m / / '
'g [f --- TABLE 15.4.2 (Cootinued)
(Sheet 3) ^-
.y 7- ..
p
-~ -
/ CORE PARAMETERS U ED IN STEAM BREA DNB ANALYSIS
.// / %,
c _/ - / -
* # / Case d. Time Poln .
[ Parameter 4 /' [/
,* 2 3 5 l
/
I ..
/ /
React Vessel inlet / ,"' temperature to sector '-' connected to affected eam generato
/
375.1 350.1 330.3 318.5
/
/
.x *F - 3,05.5 ,
E Reactor Vessel. Inlet' '
/' - /'
/
temperature <to re- .
- t. maining sector, 'F 529.7 528'.6 528.1 527.5' 526.7 s' . . / / 'l RCS prpssure, psta < 1524.0 J'348.6 1277.5 1256.7 1229.0
- i. 'RCS,Ilow,% 40.6' 32.2 27.0 24.2 .' 21.4 H Hdat flux, % - 5.8 6A3 6.19 5 . ,4 4.67 /'
l , .(/Time, sec.' //
/
5.0 35.0
/
5.0 52.5 62.5 /,
/
[ ! 1: l e
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4 I I 1 l
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( 4 Po9e _ p Tcthle 15'4i2'-I ,
- - 5 LIMITING CORE PARAMETERS USED IN STEAM BREAK i: ,
! . DNB ANALYSIS .
t c Case Inside break with power (case b) Reactor vessel inlet, 319.3'F'(Faulted.SGLoop) tamparature -. 414.2*F (Intact SG Loops) : RCS pressure 798.52 psia
.' RCS flow 100%(ofnominal) .
[ 3 Heat flux 17.60 (of nominal). Time 212.75 seconds 9
" enim .
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- IflTIAL FLOW 15 309Y L5/SEC FROM F AULTED /
j$ TEAM GENERI. TOR (fhD 9279 LB/SEC FROM THE j i. L ( , t
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[/ 250 - f' / f_ [
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1 [ LEGEND: , t E 200 I COR E HE AT F LUX ' / ./ l d 7 / =-- (PERCENT OF N0MlWA'L) STEAMRELEASE[
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'PR SURIZER' EMPTIES AT 15 ,SEC l
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/ $5 un ' IO ~
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g l l l l l 0 25 50 75 100 12[5150 175 200 TIME (SECONDS) t' [ Figure 15.4.2-2 Transient Response to Steam 1.ine Break Downstream )- . ( of Flow Measuring Nozzle with Safety injection and Off-Site Power (cose a) Revised by knendment 2 . . .*
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'5B 8/7/#1 0F FLOW MEASURING N0ZZLE WITH SAFETY INJECTION AND WITH OFF-SITE POWER (CASEA)
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" / ,/ /
2.5 /
' -
- 300 . '
/ INITI AL' STEAM / LOW 13 9798'L83/SEC FROM [
! -. l FAULTED STEAM GENERATOR ( AND 2982 LBS/SEC g
/ I FROM IMTA01 STEAM GENERATORS / '
250 p f' j/ LEGEND: .
.g. -
/
/ g '200 CORE NEAT UX (PERCENT OF NOMINAL)
{
- - == ST EA ELE'ASE ,
(PyCENT OF ION LBS/SEC) N meu b 150'
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'g 7 l FAULTED STEAM j 100 - N GENERATOR ONLY y5 o %
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'{
~
2000 PRES $URIZER EMPTIES 17 SECONDS
/
)
!. 1000 - 0 i i ! i ! 0 25 50 75 100 25 150 175 200
. TIME CONDS) 1 -
Figure 15.'4.2-3 Transient Response to Steam ine eo at Exit of Steam Generator with Safety injection and Off-Site Power (ease b) . Revised by Amendment 2 , 1 - k '
, ., _- . _ .- . . - ..~. .- - - - - .- .. - .. . . .- .
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l -W ;NTACT STEAM GENE"ATCR5 o u w IS .s0000 - - EW L $- ,@- - p TED ! TEAM GENERATOR STEAM FLC'd CNLY .-
- g. -
q, OF MNT NOMINAL STEAM FLOW) ,;
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e e e e e e C e C C C C Q C C* b C b C b C b C b Q C b C
= es N (Tt er V1 te N f
I TIME GEC) k F.m rs. + 2 -3 FIGURE 2 TRANSIENT RESPO!GE TO STEAMLINE BREAK AT EXIT OF
- e/r/F'/. STEAM GENERATOR WITri SAFETY INJECTICN AND WITH OFF-SITE PCVER (CASE B)
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!/ INITIAL FLOW IS 9798 LBS/SEC FRON FAULTED /
} STEAM GENERATOR'(AND 2982 LBS/SEC / FROH -
)
INTACT STEAN GENERATORS ,/ l [ -
/ LEGEND:
CORE MEAT FLUX /
/
, , 200 -t / (PERCENT OF MOMINAL) ,
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0 25 50 75' 100 125 150 /175 200
\
j[lHE (SECONDS) / [ Figure 15.4.2-5 Transient Response to Steem Line Brook at Exit of
- (.
Steam Generator with Sofety injection and Without Off-Site Power (caso d) Revised by Amendment 2 . y
'.;g' '
- 3 N CCN No.*?^
a#r,94_I
- ; . \ pag, _ .
- o. 1.0000 : ; ,
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TIME (SEC) p;p s5. +.2.-f TIGURE 4 TRANSIENT RESPONSE TO STEAMLINE BREAK AT EXIT OF STEAM GENERATOR WITH SAFETY INJECTION AND WITHOUT ! 0FF-SITE POWER (CASE 0) _ _ [ - .
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l/ L 'g 0 l l. L 0 50 100 150 200 TIME (SECOND ) Figure 15.4.2-7 Integrated Flow of Borated llater versus Time . l' . l, , C. . - ,%::'
n; [i :, : . 6 =. f # ,
'l W 5,'I61
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F;y e **f*2 ~7 - FIcunt S CORE BORON CONCENTRATION VERSUS TIME u e/7/89 w.
- . em . . n_ .. . . .
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