ML20082M803
| ML20082M803 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/28/1991 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20082M796 | List: |
| References | |
| NUDOCS 9109060019 | |
| Download: ML20082M803 (43) | |
Text
{{#Wiki_filter:- , , c i l ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
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LIST OF AFFECTED PAGES Unit 1 3/4 1-10 3/4 5-4 B3/4 1-3 Unit 2 3/4 1-10 3/4 5-4 B3/4 1-3 9109060019 010923 PDR ADOCK 05000327 P PDR
l s REACTIVITY f.ONTROL SYSTEMS CHARGINC PUMP 5 - OPERATING IMITING CON 0lTION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one charging "* oder 0 ~RG E, restore at least two charging pumas to OPERABLE status withi ft be in at least HOT STAND 5Y and borated to a SH'JT00W 6 ho - MARG Ct4T4444 1N ee,uiya v-two-cha ,nM(1eas ety-ptap54F4ft_jlt31L20Kwit.h_in _deIta k/' RABimtetvs-witfiin-tN tlutati 8ND nc ;t ' day'r- OL NUTO'OTN wi thTnh' e nFD0'h5Urr.N- "'] ' s SURVEILLANCE REOUIREMENis 4.1.2.4 At least two charging pumps shall be demonstrated O'E?ASLE by b2!$ + verifying, that on recirculation flow, each pu.,) cevelops a dischar;e pressure of greater than or equal to 2400 psig when tested pursuant to Specification 4.0.5. N SEQUOYAH - UNIT 1 3/4 1-10 tr.e n::me n t N o . I
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-MERGENCY CORE COOLING SYSTEMS (ECCS) -
I 3/4.5.2 ECCS SUBSYSTEMS - T Greater Than or Ecual to 350 F . _ LIMITING CONDITION F')R OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABi.E centrifugal charging pump,
- b. One OPERABLE safety injection pump,
- c. One OPERABLE residual heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and e.
An OPERABLE flow path capable of taking suction from the reiueling water storage tank on a safety injection signal and automatically transferring phase of operation. suction to the contai g sum pd g the g ig [3, ggy 5yDoc hmP kSIDur1L NOlT APPLICABILITY: MODES 1' 2 and 3. b tith ArJct% Os Ftov V'ATH 'IrgoptitnaitiTy ACTION: - p "- '
- a. WithoneECCSsubsysteminocerable(restoretheinoperablesubsystem N #'
to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTD0'aN within the following 6
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- 14 ETtt J jf C. In the event N CS is actuated and injects water into the Reactor
/ Coolant System, a REPORTABLE EVENT shall be prepared and submitted to the Commission pursuant to Specification 6.6.1. This report shall includ? a description of the circumstances of the actuittion and the f-/ totd accumulated actuation cycles to cate. The current value of the usage factor for each affected cafety injection no: le shall be provided in this report whenever its value exceeds 0.70.
SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated CPERABLE: a. At least once per 12 hours by verifying that the following valves are in the indicated cositions with power to the valve operators remoyed:
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1 REACTIVITY CONTROL SYSTEMS i BASES 6042 gallons of 20,000 ppm borated water from the boric acid storage tanks or R144 82,082 gallons of 2500 ppm borated water from the refueling water stora.ge tar.k. 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 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 af ter xenon decay and cooldown f rom 200 F to 140 F. This condition requires either 835 gallons of 20,000 ppm borated water gg44 from the boric acid storage tanks or 9,690 gallons of 2500 ppm borated water from the refueling water storage tank. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. dl$MA 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 within containment after a LOCA. This pH band minimi:es the evolution of fodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one baron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. p s - n -- y~ h 3/4.1.3 MOVABLE CONTROL ASSEMBLIES , The specifications of this section ensure that (1) acceptaDie power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accides. analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod a1iqnmep.o-4nar11gnp . ~ ~ . ~~N Fon Tse RenuiREM MU OF THE @M ATione 6pTEm Few PATt4 5 (5PECl M C A T70M5 312..I Am o 3 l.2.2.) THE C H ARG roG Pa m P5 R u REF E R.ErJt ED fort Funo PATH t t A R I F i t ATlo n) 0"ty Avo Dc NoT A F F ctT Ftow PATH OPER.6 6 t t ITY. S Rtcr uie cme oTs Foe ta A ec ica Pu m e OP ER Ro l L i TY Au f1p ms sn Se m nTRf p/ h 5tEt1 F # C RT lo os 3. l. 2, 3 /?c o 3. /. 2. '/. - As
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Jcenument No. 100 SEQUCYAH - UNIT 1 B 3/4 1-3 Revised 03/13/37 Mo'/ j j 1,0,0a
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' REACTIVITY CONTROL SYSTEMS
[ =l CHARGING-PUMPS - OPERATING .,(- - ,I ,
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LIMii1NG CONDIT:0N FOR OPERATION i h.l.2,4 At least two charging pumps shall be OPERABLE. R27
' APPLICABILITY: MODES 1, 2, 3 and 4. l i
ACTION: ; With only'one charging A LE, 'cestore at least two charging pumps to ( OPERABLE status withi - 72 ' . , ? be in at least HOT STANDBY and borated to a l u , . _,_,- _! " N b, , " E. .m _._)!,N,yN, N. .,, _M, ,m
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f f l 3 I i SURVEILLANCE REQUIREMENTS- ; b w. 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each punp develops a discharge pressure ) of greater than or equal to 2400 psig when tested pursuant to Specification ! 4.0.5. ! i l 4 l, t i t f v i
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' EMERGENCY CORE'C00 LING SYSTEMS, -:
(. q 3/4.5.2 2ECCS SUBSYSTEMS 1- T,yg Greater Than or Equal to 350'F {
-LIMITING CONDITION'FOR OPERATION l
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. t h.5.2 Two-independent emergency core cooling system (ECCS) subsystems shall {
be OPERABLE with each-subsystem comprised of: 0 1
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE, safety-injection pymp, j i
c , One OPERABLE residual heat removal heat exchanger, j 1 d.. One OPERABLE residual heat removal pump and , , I
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically j transferring suction to the containment sump during the recirculation !
phase of operation. ' Dut To S 6 ray 'LV.TEcrio r0 Nm e Rt sipudt- t APPLICABILITY: _ MODES 1, 2 and 3. - iltwr Remov At be Rmous #en, t /cmus 1 katr &cHucen Ce, Fuw Pnm Iammury, l
' ACTION: -
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- a. With one ECCS subsystem inoperable, restore the inoperable subsystem i to OPERABLE status within 72 hours or be in at least HOT STANDBY within ;
the n t 6 hours and in HOT SHUTDOWN within the following 6 hours. ! occs T 3 ' b. HER L l Mg -;
- In' tie eiFriT~the ECCS is actuated and injects water inte the Reactor Coolant. System,_a REPORTABLE EVENT shall be prepared and submitted R28 j
, , to the Commission pursuant to Specification 6.6.1. TMs report shall ; include a description of the circumstances ot- the act.ation and the } total accumulated actuation cycles to date. The current value of i the usage factor _for each affected safety. injection nozzle shall be F provided in this Special Report whenever its value exceeds 0.70. l SURVEILLANCE REQUIREMENTS !
<4.5.2 Each ECCS subsystem shall be demonstrated ,0PERABLE: !
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- a. At least-once per 12 hours by verifying that the followir.g valves are in the indicated positions with power to the valve operators removed: .!
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,s REACTIVITY CONTROL SYSTEMS i
BASES BORATION SYSTEMS / Continued)
. provide a SHUTD0 N MARGIN from expected operating conditions of 1.6% delta k/k
, after xenon decay and cooldown to 200*F. The maximu..i expected borttien capability requirement occurs at EOL from full power equilibrium xenon conditions and reNires 6042 gallons of 20,000 ppm borsted water from the g13 2 boric acid storage tanks or 82,082 gallons of 2500 ppm borated water from the refueling water storage tank. , 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 2500 ppm borated water R1; j from the refueling water storage tank. j M The contained water volume limits include allowance for water not available because of discharge line location and other physical j 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
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iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) limit the potential effec.ts 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. _ 1 V Fo A TH e Re em in.sme ers OF Tne Boe Arson; 5ysTrM Fww & rus (.5pcc/nu/rous 3 1.1.1 A n>o 3.1. 2. 2 ) TH e tu s ac, i m c Pu m ps 6 te Rep egoucen roa ruv As C LAks r s C A Tiom OnLf hvo Do TQor A F F CCT Fi.sco Pai,1 OPER ABluTy' REG u term wr.5 For Cu nRG/n/c Pu me OPCR A B it. tty kr Apore.sse Se9A RATuy 6f S P ccI ricA rtoms 3.1,2. 3 tien 3. I. 2. '/. p. W A. ^- SEQUOYAH - UNIT 2 B 3/4 1-3 Amendment No.131 Revised 08/18/87 OGT 20 030
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ENCLOSURE-2
-PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1-.AND 2:
- DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-91-03)
DESCRIPTION AND JUSTIFICATION FOR EXTEN"0N OF CHARGING PUMP ALLOWED OUTAGE TIME I f
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-Tennessee Valley Authority: proposes to modify the Sequoyah Nuclear Plant 4 (SQN) Units 1 and 2 Technical Specifications'(TSs) to revise- )
Sections 3.1.2.4 and 3.5.2 to extend the allowed outage-time (A0T) for'the i charging pumps. A bases clarification is included for reactivity control > systems to clarify that operability.of the charging pumps is addressed by l TS 3.1.2.3 and 3.1.2.4, not by the flow path specifications requirements. _) v t _ Reason for Change i SQN recently experienced a rotating element failure of one centrifugal [ charging pump,(CCP). The repair duration for this failure exceeded the i present 72-hour A0T in the-TSs'but was completed within the proposed 7-day - l A0T. Unit shutdown was' required by TSs, which were complied with to
-accommodate the CCP repairs. Failures of high head injection pumps that '
require greater than 72 hours to be repairs have been observed in the -[ nuclear industry. ' Unit: shutdown necessitated by such repairs is not [ justified if a longer A0T will not significantly impact nuclear safety.- l For=these reasons, SQN is pursuing a TS change to implement a 7-day A0T .!
-for inoperability of one CCP. This will increase plant availability as ;
opposed to performing unnecessary plant shutdowns. ! , - - 1 Justification for Change. ! f The nuclear safety impact of going from a 3-day A0T to a 7-day A0T for one [ inoperable CCP 'has been evaluated by Westinghouse Electric Corporation and I documented in Westinghouse Letter TVA-91-120 to TVA dated April 23, 1991 l (Enclosure 4). This Westinghouse evaluation has utilized a probabilistic r risk analysis'(PRA) approach to support extending the CCP A0T. .For this ! PRA the maximum A0T of 3 days or 7 days was used in place of an average - l out!-of-service time as used in;the SQN Individual _ Plant _ Evaluation Report ! I base case. issued for publication through the Nuclear Management and' !
-Resources Council on August 9, 1988, to Industry Degraded Core Rulemaking, !
for cora damage f requency analysis. ! i 5 This evaluation began by determining all accident sequences where the !
-charging pumps are utilized to prevent core damage. This determination [
-identified these accidents to be the small break loss of coolant accident {
(LOCA), medium break LOCA, transient events. requiring two pressurizer j i power-operated relief valves to open with-the failure of'feedwater j
-systems, loss of component cooling water, and high pressure j recirculation. For main steamline break, f eedline break,- or steam .t generator. tube rupture events, the CCP requirements are the same as for ;
small break LOCA initially; and.possible long-term requirements would be [ similar to the transient event. . With the identification of events i requiring CCPs, unavailabilities.were determined for one CCP out of i service. This unavailability, along with the availability of each train i of support systems, was used to determine the unavailabilities for high l pressure injection associated with small and medium break LOCAs as well as ! for loss of component cooling water and for high-pressure recirculation [ f or all recirculation required events. + l i
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After thelunavailabilities for onofCCP out of service were determined,fthe ;; effect;of 3- and 7-day A0Ts_was calculated and factored in.to provide-the i unavailabilities for one CCP _out of service for_3 daysiand: 7_ days. These l
'unavailabilities were utilized to quantify.the event trees to_ determine j the affect on core damage frequency. -It was calculated that-core damage :
f requency increases of 1.3SE-08/ year fot fsmall break LOCA sequences, ! 9.50E-09/ year for medium break LOCA sequences, and 7.80E-09/ year for loss of component cooling water sequences combine for c total increase of fr 3.11E-08/ year. The total core damage frequency for all sequences is I 2.93E-05/ year for_the base case in-the IPE and is the same for the 3-day
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and the 7-day CCP A0T f requencies. Since the 3.llE-08/ year increase in core damage frequency for a'7-day CCP A0T is less than 0.11 percent of , overall coreidamage frequency. the effact is-insignificant and the-2.93E-05/ year frequency-term is not changed ~by.this increase. Therefore, _- i the increase in A0T for one CCP out of service to 7 days does not i significantly increase the core damage f requency and is acceptable for -{ implementation at SQN. .j Environmental I Impact Evaluation f .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 not: _1 . Result in a significant it. crease in any adverse environmental _ impact ~
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previously evaluated in the Final-Environmental Statement (FES) as ! modified by the staff's testimony to the Atomic Safety and Licensing j Board. supplements to the FES, environmental impact appraisals, or ! decisions of'the Atomic Safety and Licensing Board. j e
- 2. Result in a significant change in effluents or power levels. [
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_ f 3.- Result in matters not previously reviewed in the licensing Lbasis f or f SQN that may have a significant' environmental impact. !
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.- ' ~ ENC.0SURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEOUG'tAH NUCLEAR FIANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-91-03) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION
. s s ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consequences of an accident prev!ously evaluated.
The centrifugal charging pumps (CCPs) provide emergency core cooling and reactor boration contrnl f unctions. These functions provide accident mitigation functions as well as nakeup for the reactor coolant system (RCS). However, these functions are not considered the initiation of any postulated design basis accident. Therefore, this change to the CCF A0T will not increase the probability of an accident. As shown in Westinghouse Electric Corporation Letter TVA-91-120 to TVA dated April 23, 199L (Enclosure 4), the increase of 3.11E-08/ year in core damage frequency for the CCP allowed outage time (A0T) being increased from a 3-day A0T to a 7-day ACT is insignificant. The probabilistic risk analysis (PRA) methodology used in this Westinghouse evaluation applied the CCP unavailabilities for a 7-day A0T to the event fault trees to determine the core damage frequency effect. The 3.llE-0S/ year frequency increase is only 0.11 percent of the 2.93E-05/ year overall core damage f requency for accident sequences. The increase is insignificant, and therefore increasing the A0T for CCPs to 7 days does not significantly increase the consequences of an accident.
- 2. Create the possibility of a new or dif ferent kind of accident from any previously analyzed.
Since the CCPs provide RJ4 makeup and accident mitigation functions, the generation of or chat - .n an accident type by the CCPs being out of service is not credible. Therefore, an increase for the CCP to a 7-day A0T will not create the possibility of a new or different kind of accident.
- 3. Involve a significant reduction in a margin of safety.
An increase from 3 days to 7 da*s f for the CCP A0T does net change any of the functions or the conditions that will require the operation of the CCPs. The only impact is the time allowed for one CCP to be out of service without requiring unit shutdown. While the CCPs are in service, they will operate under normal and emergency conditions without change as assumed in the safety analysis. The change to a 7-day CCP ACT will result in an insignificant increase in core damage frequency. Therefore, this change does not involve a significant reduction in any margin of safety.
0, . . . . . . . . . . . - . . . g ENCLCSURE 4 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND $0-32S (TVA-SQN-TS-91-03) PROEABILISTIC RISK ANALYSIS FOR INCREASING THE ALLOWED OUTAGE TIME FCR CENTRIFUCAL CHARGING PUMPS
,- AF;RL 3 9917.18i56 2 FROtt 'L ICE NSlHG ' -TO TOR _SEQUOY .
PAGE.002 i
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$84p,,,u.3m Westinghouse Energy Systems , ',
Electric Corporation j i 1 i I Mr. P. G. Trudel- TVA-91-120L
- Project Engineer NS-0PLS-0PL-11-9)-241 !
Tennessee. Valley Authority April 23, 1991 J P.- 0.-Box 2000 Ref: 1 TVA RO 998067 i Soddy Daisy, TN 37379 2 W G.O. C0545205- -j 3 TVA-91 ll3 l 4 TVA-91-ll2 ; Tennessee Valley Authority-Sequoyah Nuclear Plants-Units 1 and 2 : CCP A0T Extension to Seven Days . ! Technical Specification Change Justification -i Revision 1 '
Dear Mr. Trudel:
In accordance with your-request, the attached justification to support a_ _ permanent Technical Specification change for extending the. Centrifugal- Charging i Pump (CCP) Allowable Outage Time (A0T) from the current 72-hour time--interval to ~ j 7 days has been revised to incorporate TVA's comments. This justification is ; based upon the application of the~Probabilistic Risk Analysis (PRA) methodology. l
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.lf you have any questions, please do not hesitate to contact us. l 4
Very truly yours, ; r i gB.'J.Garry, Manager ; I TVA Sequoyah Project : l Domestic Projects Department j LVT/lg- l Attachment i
-cc: D. M. Lafever _[
-R. G.-Davis J k
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,- ARR R3 '91 10:57 FRoll L ICENS!!4G TO TVA SEQUOY PAGE,003 f
Page 1 of 29 Attachment To TVA 91-120 Tennessee Valley Authority Sequoyah_ Units 1 & 2 CCP Allowable Outage Time Extension To 7 Days ' Justification for Permanent Technical Specification Change Revision 1,-April 23, 1991 l 1,0 INTRODUCTION The current Tennessee Valley Authority's Sequoyah Nuclear Plant Units 1 and 2 , Technical Specifications allow a centrifugal charging pump (CCP) to be out of . service for 72 hours. -This justification documents the Probabilistic Risk Analysts (PRA) approach to support extending the Allowable Outage Time (A0T) i for the CCP from the current 3 day A0T to a 7 day A0T. The applicable marked-up Technical Specification sections are also provided. 1,1 Technical Specification Bases Explanation and Justification Relaxation The primary function of the Safety injection System (SIS) is to provide emergency core coeling (ECC) in the event of a toss-of-Coolant Accident (t0CA) resulting from a break in the primary Reactor Coolant System (RCS) or to ; provide emergency boration in the event of a Steamline Break accident resulting from a break in the Secondary Steam System. > The SIS is designed to provide borated water to the RCS to insure that the ! applicable criteria are satisfied assuming that both a loss of offsite power
- and single failure may occur coincident with a postulated accident.
The primary function of the reactivity control system (flow paths) is to , provide plant shutdown capability. The bases for this system is tied to GDC-26, For 10 CFR part 50, Appendix A, GDC-26 requires "Two independent , reactivity control systems of different design principles shall be provided. ; One of the systems shall use control rods.,,". The second system "shall be capable of reliably controlling the rate of reactivity changes from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core suberitical under cold conditions." , Per the Standard Technical S)ecifications (STS), if a component is found to be - inoperable, it will be possible in most cases to effect repairs and restore the ! system to full operability within a relatively short time. For a single component to be inoperable does not negate the ability of the system to perform i its function but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures. The i roquirement to place the reactor initially in the hot shutdown condition l provides for reduction of the decay heat from the fuel and consequently a i reduction of cooling requirements after a postulated loss-of-coolant accident. This also permitted improved access for repairs in some cases. The limiting times to repair are based on two considerations: , i
- 1. Assuring high reliability that the safeguard system will function ;
properly if required to do so, and, [ t
- 2. Allowances of sufficient time to ef fect repairs using safe and proper
- procedures, l
* . Page 2 of 29 l-Historically, failure to complete repairs within the specified time of going to l the HOT SHUTDOWN condition is considered indicative of a requirement for major o maintenance, and therefore, in such a case the reactor is to be put into the l COLD SHUTDOWN condition.
E ! There is a technical approach that can be used to justify extending the allowed l outage time (A0T) for the ECCS fluid system from 72 hours to 7 days. The probablistic risk assessment arguments (PRA) can demonstrate that the safeguard system will still maintain a high level of reliability to function if required to do so. Since it can be demonstrated that the A0T required by the STS does not sllow sufficient time to complete maintenance procedures that are important for plant continued operation, the relaxed A0T will provide time for more thorough and effective equipment maintenance, This increase, in turn, will improve plant safety and increase the plant availability by reducing unnecessary plant shutdown. As an example of precedent, Byron Station has performed this relaxation. Diablo Canyon is now in the process of making these Technical Specification changes. Per Technical Specifications 3.1.2.4 and 3.5.2 at least one charging pump must i be operable or the plant is placed into Tecii Spec 3.0.3. Technical Specification 3.1.2.2 addresses operable flow paths for emergency boration. The Technical Specifications for operability of the CCP are 3.1.2.3 and 3.1.2.4. Therefore no relaxation is needed for 3.1.2.1 and 3.1.2.2 to ensure flow path operability. I 1.2 Probabilistic Risk Analysis The Sequoyah Individual plant Evaluation (IPE) reports a base case core damage frequency analysis for Sequoyah Nuclear Plant Units 1 and 2. An average A0T is l used for components in this analysis as they are not expected to always be out j of service for the maximum allowed time. For the analyses reported in this document, an average out of service time is not used. Instead, the out of service time for the charging pumps is assigned the upper bound of allowable time (3 days or 7 days). l l a l
.++ TOTAL P A 3E . 002 ++
, AFR.24 '91 18: 49 F R 0f1 LICENSING TO TVA SEQUOY PAGE.CO2 Page 3 of 29 2.0
SUMMARY
The following steps were performed to evaluate the impact of changing the A0T for Sequoyah units 1 and 2:
- 1. Determine all sequences where the charging pumps are called upon to prevent core damage,
- 2. Requantify the fault trees used in step 1 with one charging pump removed,
- 3. Develop equations to incorporate changing A0T times,
- 4. Quantify the event trees with an A01 of 3 days,
- 5. Quantify the event trees with an A0T of 7 days.
- 6. Compare the core damage frequencies for an A0T of 3 days and 7 days.
The results of these analyses are shown in summary tables for each of the dominant sequences evaluated for a change in the A0T for charging pumps, by each initiating event and the dominant core damage sequences for three cases: an A0T of 3 days, an A0i of 7 days, and the base case used in Reference 1. The total core damage frequency from all initiating events is 2.93E-05/ year for each of the cases (from Tables 5 through 8). Some of the dominant sequences ; for the small LOCA, medium LOCA and loss of component cooling water initieting events increased. The change in each sequence is summarized in Table 4. The increase in core dainage frequency is due to an increase of J.38E-8/yr (small LOCA, failure of recirculation), 9.5E-9/yr (medium LOCA, failure of high pressure injection) and 7.8E-9 (failure of the charging pumps following a loss , of component cooling water). The sum of all sequences for each of these i initiating events are-4 Initiating Event A0T Core Damage Frequency Percent of Total Core (For Given initiating Damage Frequency ; (For Given initiating Event) Event) Small LOCA 3 days 1.09E-05/ year 37.2 :' Small LOCA 7 days 1.09E-05/ year 37.2 Medium LOCA 3 days 1.77E-06/ year 6.0 Medium LOCA 7 days 1.78E-06/ year 6.1 Loss of CCW 3 days 1.96E-08/ year 0.1 Loss of CCW 7 days 2.74E-08/ year 0.1 By changing the A0T from 3 days to 7 days, the increase in core damage frequency is 3.lE-08/ year. The total increase in total core damage frequency , is 0.11 percent ((3.1E-8/2.93E-5)x100]. This is not a significant change as it does not contribute to the overall core damage frequw.cy. Therefore, the results of the probabilistic analysis shown that changing the A0T for a charging pump from 3 days to 7 days does not impact plant risk. r
, APR. 24 '91 1S:50- FROM L! CENSING- TO TUA SEQUOY PAGE.003 Page 4 of 29 The analysis effort includes a Fault Tree Analysis (3.1), Unavailabilities for
-Event Tree Analyses (3.2), and a Calculation of A0T Times (3.3). Additionally, the following tables are included.
1 Results of ECCS Fault Tree Quantification 2 Unavailability of Top Events by Support State : 3 Unavailabilities for A0T of 3 Days and 7 Days t 4 Comparison of Dominant Sequences
$ Dominant Sequences. A0T 3 Days l 6 Dominant Sequences - A0T 7 Days :
7 Dominant Sequences Base Case ! 8 Core Damage Frequency Results by Initiating Event j l 3.0 ANALYSIS
\
'The success criteria (documented in Reference 1) credited the CCPs for :
prevention of core damage for each of the following initiating events: l t
- 1. The medium LOCA requires two out of four [two Safety Injection Pumps (SIPS) and two CCPs] high pressure injection pumps to prevent core !
damage.
- 2. The small LOCA requires one out of four (2 SIPS and 2 CCPs) high l pressure injection pumps to prevent core damage. }
3 The transient initiating events require the opening of two pressurizer power operated relief valves and one out of four (two SIPS and two ; CCPs) high pressure injection pumps to prevent core damage, if heat t removal via the feedwater systems fail. :
- 4. The loss of component' cooling water requires that one charging pump !
continue to provides reactor coolant pump seal cooling. .
- 5. High pressure recirculation is addressed on success of high pressure !
injection. One of the two residual heat removal pumps (RHRPs) and one ! out of four high pressure injection pumps are required to continue recirculation, i It should be noted that the steam generator tube rupture and main h steamline/feedline breaks were not addressed in Reference 1. However, the , success criteria for high-pressure injection would be the same as for the small- . LOCA for both of these events. ECCS recirculation is not initially required ! for either event, therefore, changing the A0T time for the charging pump would l l not impact either of these events. [ l i i [ I E 6 e - , , , - -- , - - . ,-e ., - , - - , -
_m _ _ _ _ _ _ _ _ ____ __ . . _ . _ _ - _ _ _ . .m_ e APR E3 '91 19:02 F R 011 L I C E t45114G TO TVA SEQUOY PAGE.007 Page 5 of 29 Eight support states are defined in Reference 1. The availability of the supporting systems (AC and DC power, essential raw cooling water, and component cooling water) define these support states. All support systems are available in support state I so that both trains of all systems depending on these support states are also available. Only 1 train of support systems are available in support state 2 and 5 so that only 1 train of each system depending on these support states are available. One train of essential raw cooling water is available but component cooling water is not available in support state 3 so that only the CCPs are available for reactor coolant pump seal cooling. Neither high pressure injection nor recirculation would be available in support states 4, 6, 7 and 8 because of the combinations of unavailabilities of the. supporting systems. 3.1 Fault Tree Analysis - The emergency core cooling system's high pressure injection and recirculation fault trees (Reference 2) were altered to simulate the unavailability of one CCP during its allowed outage. The unavailability of the CCP during its A0T was simulated by its removal from the fault tree models. It was assumed that i
-the pump would be properly isolated during the outage; therefore, the isolation ;
valves for the CCP in question were removed from the model as well. These fault trees were then requantified with one CCP removed. The original fault tree models and quantification output developed during the Sequoyah IPE are used as the base case, The original base case assumptions : were also used. These base case fault trees, quantification results, and assumptions were obtained from Reference 2. The names of the base case fault trees and their quantification results used in this analysis are presented in columns one and two, respectively, of Table 1. l The success criteria for the Emergency Core Cooling System (ECCS) high pressure ; injection and recirculation fault trees obtained from R6ference 1 are as follows: INJECTION MODE [ INITIATING EVElil ML(CESS CRITERIA MISSION TINF i Small LOCA 1 of 4 SIPS or CCPs 1/2 hour to 1 of 4 RCS Cold legs ! Medium LOCA 2 of 4 SIPS or CCPs 1/2 hour ! to 2 of 4 RCS Cold-Legs 4
~
TABLE I @ x RESULTS OF ECCS FAULT TREE QUANTIFICATION y' O T - FOR INCLUSION IN EVENT TREE N00ELS . $ g, _ w Col. 1 Col. 2 Col. 3 Col. 4 Col. 5 Col. 6 Col. 7 g UNAVAILABILITY CCfs FROM UNAVAILABILITY UNAVAILABILITY COLUPW 3 + COLUPm 3 + FAULT TREE FROM BASE CASE BASE CASE CCP 1A-A OUT CCP IB-B OUT COLUMN 4 COLUPW 5 ,. x 2 HPIS_S 9.6E-08 1.OE-04 3.3E-06 3.3E-06 1.0E-04 1.0E-04 r-HPI'J_SLA 1.5E-04 3.0E-09 1.5E-04* 1.4E-02 1.5E-04 1.4E-02 h HPIS__SLB 1.2E-04 3.0E-09 1.IE-02 1.2E-04* 1.IE-02 1 2E-04 HPIS_CEA 1.lE-02 - 1.IE-02* 1.0E+00 1.lE-02 I.0E+00 ! HPIS_CEB 1.lE-02 - 1.0E400 1.lE-02* 1.0E+00 I.lE-02 l j HPIS_M 1.3E-05 4.5E-05 2.5E-04 2.5E-04 3.00-04 3.0E-04 HPIS_MLA 2.5E-02 - 2.5E-02* 1.00+00 2.5E-02 1.0E+00 HPIS_MLB 2.IE-02 - 1.0E+00 2.lE-02 1.0E+00 2.lE-02 g l HPR 2.9E-04 3.3E-05 2.9E-04 2.9E-04 3.2E-04 3.2E-04 -* 5 HPR_LA 1.7E-02 - 1.7E-02* 2.4E-02 1.7E-02 2.4E-02 g c j HPR._LB 1.7E-02** - 2.4E-02 1.7E-02 2.4E-02 1.7E-C2 g
-c
- Since outage of the specified pump wou.d not alter the system unavailability if its train is assumed unavailable, these cases were not modeled; the unavailability is the same as in the base case. p Since single failures of the RHR system dominate the recirculation with one train unavailable scenario, h this case was not analyzed in the original IPE. The unavailability of HPR_LB is assumed to be equal to o that of HPR_lA. 8
.. AP.R 23 '91 19:04 FRON LICENSING TO TVA SEQUOY PAGE.009 Page 7 of 29 l 1 RECIRCULATION MODE : INITIATING EVENT 10rCESS CRITERIA 211El1Qt(_llfiE Small LOCA or 1 of 2 RHRPs to 1 of 4 231/2 hours Medium LOCA SIPS or CCPs to 2 of 4 RCS Cold legs Two cases were analyzed in this study. Case A involved the removal from ! service of CCP 1A A; Case B involved the removal from service of CCP IB-B. The appropriate CCP, and its respective isolation valves, were removed from the i fault trees listed in Table 1. These new fault trees were then re-quantified. Columns 4 and 5 of Table 1 summarize the quantification results of Cases A and # B, re pectively. Note that not all base cases changed. This is because some of the base cases already assumed entire train failures. For example, fault tree HPIS_SLA modeled the unavailability of the ECCS to mitigate a small LOCA . if train A electric )ower was unavailable. Hence, CCP 1A-A is already removed ' from the tree, and tie base case results will not change for Case A. The common cause calculations from the base case (see column 3 of Table 1) are used as a conservative estimate of common cause failure rates. Removing a pump , from service will actually decrease total system common cause dependency, i because the common cause normally associated with that pump and its complement will be removed. The total system unavailability, i.e., random plus common cause unavailability, for Case A and Case B is given in columns 6 and 7, respectively, of Table 1. 3.2 Unavailabilities for Event Tree Analyses The unavailabilities shown in Table 1 are used to quantify the event trees. These unavailabilities depend on the support state. As shown in Table 1, these unavailabilities are calculated for several conditions. The results of the fault tree analyses for HPIS_SLA, HPIS_SLB, HPIS_MLA, HPIS_MLB are used in ' support states 2 and 5. If a charging pump is out of service in support states ' 2 and 5. only one train of high pressure injection (HPI) or recirculation (HPR) would be available. There is a 0.5 probability that train A would be available and a 0.5 probability that train B would be available. Therefore, the -l unavailability of HPI or HPR is calculated as: , 0.5 (probability train A not available) + 0.5 (probability train 8 not . available). : a 9
~ _ .
.. APR es si isios rao,4 ticensino to run scouoy enet.oio i
-Page 8 of 29 If only one train is available, there is also a 0.5 probability that the associated charging pump is also available: j Train Available Probability CCP Out by Support State CCP_A CCP_B Train A 0.5 0.5 ,
Train B 0.5 0.5 ; 1 This is the probability that one pump is out (0.5) times the unavailability of ! that train given that the pump is out: , I 0.5x (CCP_A out + CCP_8 out) (train A not available) 0.5x (CCP_A out + CCP_B-out) (train B not availabic) ; These equations are multiplied that each train is for the by the probability of HP(0.5)I for the fault trees available. As an example, unavailability ' quantified for the small LOCA: , i Unavailability in Support States 2 and 5: l 0.5 x [0.5(1.5E-4 + 1.4E-2) + 0.5(1.lE-2 + 1.2E-4)) - 6.32E-3 If a charging pump is not out of service, the calculation is: > 0.5 x (1.5E-4 + 1.2E-4) = 1.35E-4 ' This same calcul6 tion was performed for the medium LOCA HPI and the loss of Component Cooling Water (CCW) event. . The failure probability of the event tree top events can be defined with three r unavailabilities for the high pressure injection (HPI) mode an HPI-S = unavailability calculated for small LOCA and transient events HPI-M = unavailability calculated for medium LOCA HPI-C - unavailability calculated for loss of CCW event - One unavailability is used in this analysis for the failure probability of the event tree top for high pressure recirculation (HPR) as: HPR - unavailability calculated for all events requiring recirculation. An operator error of 1.0E-3 is added to the unavailabilities shown in Table 1 to calculate the total unavailability of HPR.
. I D
1 l r
, Ar/R 23 '91 19:06 F ROtt L I CE N51t4G TO TVA SEQUOY PAGE.0li ,
Page 9 of 29 The results of these calculations are shown in Table 2. The base case ! unavailabilities are the same as used h. Reference 1 and represent the unavailabilities of the top events with no charging pumps out of service. , Note that no ECCS injection or recirculation is possible in support states 3, 4, 6, 7 or 8 and the unavailability is 1.0. Table 2 i Unavailability of Top Events by Support State ; Support State Base Case One CCP Out ; HPI-S (Small LOCA and Transients) 1 1.00E-04 1.00E-04 ; 2,5 1.37E-04 6.32E-03 ; 3,4,6,7,8 1.00 1.00 HPI-M (Hedium LOCA) 5.80E-05 ' 1 2.95E-04 2,5 2.30E-02 5.11E-01 3,9,6,7,8 1.00 1.00 HPI-C (Loss of CCW) ; 3 1.10E-02 5.05E-01 ; High Pressure Recirculation 1 1.32E-03 1.32E-3 2,5 1.80E-02 2.15E-2 3,4,6,7,8 1.00 1.00 M 3.3 Calculation of A0T Times ' This section provides the methodology to calculate the impact of changing the , A0T times for the CCPs. The total time during a year is designated as the time T. The time T can be partitioned into the time when only one CCP (To) is out of service or none of . the CCPs (Tn) are out of service. Both CCPs are not allowed to be out of : service at the same time. I i
, Af?R 2391- 19:06 FROM L ICEf1 Sit 4G TO TVA SEQUOY PAGE.014 Page 10 of 29 > t f i Tn To r Total Time T During the time interval Tn, both CCPs would be available, and the 5 unavailability 021 of each of the HP1,'nditions would be the random failure of 2 trains of HPI. During the time interval To, one CCP could be out of service for a time To and i-only one CCP would be available and the unavailability Qli of the HPI system would be the random failure of the remaining CCP. The total time T can be partitioned as: T - Tn + To If this equation is divided by the total time T: ; Tn + In = 1.0 I T T I Then- - To . The fraction of time (or probability) when both CCPs are ! T available (or that no CCPs are out of service). _ The unavailability of the HPI system is Q21. Tq - The fraction of time (or probability) when only one CCP is : T available (and one CCP is out of service). The unavailability of
- the CCP is Qli, f The total unavailability of each top event for HFI or HPR (QHP1) condition is:
- QHPi -
Tn x Q21 + 12 x Qli (Equation 1) T T i The unavailability of each of the event tree top events was quantified using i Equation 1 and the results are shown in Table 3. I
, AeR 83 '91 19:07 F R0l1 L I C E t4S il1G T TVA SEQUOY PAGE.013
'Page 11 of 29 r
Table 3 ; Unavailabiliti.s for A0T of 3 Days and 7 Days Unavailability ! HPli by Support State CCP out 3 days CCP out 7 days $ HPI-S 1 (Q21) 1.00E-04 1.00E-04 - 2,5 (Q11) 1.88E-04 2.56E-04 3,4,6,7,8 1.00E+00 1.00E+00 , HPI-H 1 (Q22) 5.99E-05 6.25E-05 > 2,5 (Q12) 2.70E-02 3.24E-02 3,4,6,7,8 1.00E+00 1.00E+00 HPI-C (Q31) 1.51E 02 2.05E 02 [ t HPR ! 1 (Qdl) 1.32E-03 1.32E-03 ' 2,5tQ42) 1.80E-02 1.81E-02 - 3,4,6,7,8 1.00E+00 1.00E400 - 1 Note: Each Q21 and Q11 was calculated with Equation 1. ! __ l i i e i i
- ?
.. QPR.24 '91 19:51 F R0f1 L I C E NS I NG TO TVA SEQUOY PAGE.004 Page 12 of 29 4.0 RESULTS The unavailabilities shown in Table 3 were used to quantify the event-trees ! described in Reference 1. Table 4 shows the increase in core damage frequency ; calculated for each of the sequences which are impacted by changing the A0Ts ' for the charging pumps. The sequence number, core damage frequency calculated ; for a 7 day A0T and a 3 day A0T and the increase in core damage frequency by ' changing from a 3 day A0T to a 7 day A0T are given in the table. The total increase over all sequences is 3.11E-08/ year. The total core damage frequency was calculated for 3 cases: 3 day A0T, 7 day A0T and the base case. The total core damage frequency is 2.93E-05/ year for each of the cases. i Tables 5, 6, and 7 list the dominant sequences for each of the case's. The core i damage frequency, percent of core damage frequency, the plant damage state, ' event tree identifier and event tree sequence and failed top events (nodes) are
~
shown in these tables. Table 8 shows the core damage frecuency per year calculated for each initiating -! event. The total core damage frequency, parcent contribution, conditional , probability and the frequency of the initiating event are shown on this table ' for each initiating event evaluated in Reference 1. The results are shown for an A0T of 3 days, 7 days and the base case (with an average outage time of ; about 18 hcurs). The sum of all sequences for small LOCA and medium LOCA and ; loss of co.nponent cooling water are: ;
+
Initiating Event A0T Core Damage Frequency Percent of Total Co e Damage Frequency 7 Small LOCA 3 days 1.09E-05/ year 37.2 - Small LOCA 7 days 1.09E-05/ year 37.2
- Medium LOCA 3 days 1.77E-06/ year 6.0 Medium LOCA 7 days 1.78E-06/ year 6.1 +
toss of CCW 3 days 1.96E-08/ year 0.1 Loss of CCW 7 days 2.74E-08/ year 0.1 I t i t
.. - - _ . ._ ._ ________O
. AF R . I .1 '91 10:52 F%CM LICENS!!4G TO TVA SEQUOY PAGE.005 Page 13 of 29 I Table 4 Comparison of Dominant Sequences Core Damage frequency Sequenew Humber 7 day A0T 3 day A0T 'ncrease Small LOCA, f ailure of Recirculation 1 8.22E-06 8.2?E 06 0.00E+00 6 1.31E 06 1.31E 06 0.00E+00 7 1.19E 06 1.18E-06 1.00E 08 11 7.64E 07 7.64E 07 0.00i+00 21 2.79E 07 2.77E 07 2.00E 09 26 1.68E-07 1.67E-07 1.00E 09 32 1.21E-07 1.21E-07 0.00E+00 39 7.25E-08 7.21E 08 4.00E-10 46 3.95E-08 3.93E-08 2.00E-10 54 2.69E 08 2.68E 08 1.00E.10 64 1.70E 08 1.69E C8 1.00E-10 Total 1.22E 05 1.22E 05 1.38E 08 Small LOCA, Failure of High Pressure Injection 3 1.85E 06 1.85E-06 0.00E+00 10 7.95E 07 7.95E-07 0.00E+00 24 1.89E 07 1.89E-07 0.00E+00 30 1.38E 07 1.38E 07 0.00E+00 35 1.01E-07 1.01E 07 0.00E+00 44 4.45E 08 4.45E-08 0.00E+00 59 2.20E-08 2.20E 08 0.00E+00 Total 3.14E-06 3.14E 06 0.00E+00 Medium LOCA, Failure of Recirculation 8 9.39E-07 9.39E 07 0.00E+00 31 1.31E 07 1.31E 07 0.00E+00 27 8.73E-08 8.73E 08 0.00E,00 51 3.08E-08 3.08E 08 0.00E+00 Total 1.19E-06 1.19E 06 0.00E+00 Medium LOCA, Failure of High Pressure Injection 43/41 6.09E-08 5.14E 08 9.50E 09 Failure of Component Cooling Water 53 2.73E 08 1.95E-08 7.80E-09 Total increase in core damage frequency
- 1.38E 08 + 9.5E 09 +7.80E 3.11E-08/yoar 4
, AER J:3 '91 19:10 F RO11 L I C E NS l HG TO TVA SEGUOY PAGL.016 Page 14 of 29 i Table 5 Dominant Sequences A0T 3 Days Core Damage Percent Damage Event Troo Sequence Support State Frequency State . Number and failed Nodes 1 8.22E 06 28.1 SLC2 ET3 SLOC 2 1 R2 ; 2 1. <E 06 6.7 TEC2 ET6 LOSP 12 5 AF DAl ! t 3 1.85E 06 6.3 SE ETr r 26 7 HP OA CS l 4 1.75E-06 6.0 1EC2 ET7 STBL 22 '. 012 FH4 5 1.44E 06 4.9 TEC2 E15 1RAN 13 1 AF MF OA1 -[ i 6 1.31E 06 4.5 SLC2 E18 SLOC 2 1 R2 7 1.18E 06 4.0 SLC2 ET3 SLOC 2 2 R2 8 9.39E 07 3.2 ALC2 ET2 MLOC 2 1 R2 9 8.60E 07 2.9 Y E14 V SE 1 1
- 10 7.95E 07 2.7 SE ETu SLOC 26 4 HP OA CS !
11 7.64E-07 26 SLCl ET3 SLOC 3 1 R2 R3 12 6.59E 07 2.2 TE ET7 STBL 26 d EH2 EH4 EH7 13 6.43E-07 2.2 TEC't ET9 ATWS 7 1 RTl OA3 , 14 6.24E-07 2.1 TEC2 ET7 STBL 46 8 AFT EH1 15 6.07E-07 2.1 TEC2 ET9 A1WS 20 1 RTl PL PR 16 4.73E 07 1.6 AEC2 ET2 MLOC 17 1 AC 4.28E 07 5 AF NF OAl 17 1.5 TEC2 ETS TRAN 13 18 3.16E-07 1.1 TEC2 ET9 ATWS 16 1 Ril PL OA3 ; 19 3.06E-07 1.0 TEC2 ET4 TRAN 12 1 AF OA1 20 2.93E-07 1.0 ALC2 ETI LLOC, 2 1 R1 ; 21 2.77E 07 .9 SLCl LT3 SLOC 3 2 R2 R3 22 2.07E 07 .7 SEC2 ET3 Stoc 3: 1 RT 23 1.94E 07 .7 TEC2 ET9 ATWS 28 2 Ril PL AM O
-- .,m e--.- - . - . - , , - . - . - -
, APR' 23 ' 91 19 10. F ROM L I CEi4$ 111G To TVA SEQUOY PAGE.017 Page 15 of 29 Table 5 (Continued) Dominant Sequences A0T 3 Days Core Damage Percent Damage Event Tree Sequence Support State frequency State Number and failed Nodes 24 1.89E 07 .6 SE ET8 SLOC 26 8 HP OA CS 25 1.78E-07 .6 AEC2 ET1 LLOC 11 1 AC 26 1.67C 07 .6 SLC2 ET8 SLOC 2 2 R2-27 1.66E-07 .6 TEC2 ET9 ATWS 28 1 RT1 PL- AM 28 1.560 07 .5 TE ET7 ST8L 50 8 AFT EH1 CH3 29 1.47E 07 .5 TEtt E10 LOSS 12 1 AF OA1 30 1.38E 07 .5 SEC2 ET3 SLOC 23 1 HP OA 31 1.31E 07 .4 ALC2 ET2 MLOC 2 2 R2 32 1.21E 07 .4 SLC1 ET8 SLOC 3 1 R2 R3 33 1.06E 07 .4 SE ET7 STOL 21 8 EH2 S2 CV 34 1.06E 07 .4 TLC 2 ET5 TRAN 4 1 AF NF R2 35 1.01E-07 .3 SE ET8 SLOC 26 6 HP OA CS 36 9.64E 08 .3 TE E16 LOSP 15 6 AF OA1 CS 37 8.73E 08 .3 ALC1 ET2 MLOC 3 1 R2 R3 38 7.45E 08 .3 TEC2 'ET7 STBL- 27 8 EH2 OA? 39 7.21E 08 .2 SLC2 ET8 SLOC 2 5 R2 40 6.38E 08 .2 TEC2 ET5 TRAN 13 2 AF NF '0A1 41 5.85E-08 .2 TLC 2 ETS TRAN 4 2 AF NF R2 42 5.12E-08 .2 AEC2 ETI LLOC 7 1 LP 43 S.14E 08 .2 AEC2 ET2 MLOC 13- 2 HP OA 44 4.45E 08 .2 SE ET8 SLOC 26 3 HP OA CS 45 4.42E 08 .2 TEC1 ET6 LOSP 13 5 AF OA1 R3 46 3.93E-08 .1 SLC1 ET8 SLOC 3 2 R2 R3
- APR 23 '91 19 11 f hort L ICEllsif4G TO TVA SEQUOY PAGE.010 Page 16 of 29 Table 5 (Continued)
Dominant Sequences - A0T 3 Days Core Damage Percent Damage Event Tree Sequence Support State Frequency State Number and failed Nodes 47 3.68E-08 .1 TE ET6 LOSP 15 5 AF OAl CS l 48 3.55E 08 .1 TE ET9 ATWS 31 6 RT1 PL AM CS 49 3.29E-08 .1 SEC2 LIB SLOC 31 1 RT ; I $0 3.26E-08 .1 ALCl ETI LLOC 3 1 R1 R3 51 3.08E-08 .1 ALCl ET2 MLOC 3 2 R2 R3 52 2.99E-08 .1 AEC2 ET) LLOC 7 LP 53 2.68E-08 .1 SL ET3 SLOC 6 2 CS R2 54 2.40E 08 .1 SE E13 ERCW l 4 55 2,37E 08 .1 AEC2 ET2 HLOC 21 1 RT [ 56 2.36E-08 .1 TEC2 ET4 TRAN 12 5 AF OA1 57 2.24E-08 .1 TLC 2 ET4 TRAN 3 1 AF R2
- 58 2.20E-08 .1 SEC2 ET8 SLOC 23 1 HP OA 59 2.07E 08 .1 SLWCl ET3 SLOC 4 1 DR 60 1.96E-08 .1 TEC2 ET9 ATWS 11 1 RTl AF l 61 1.95E-08 .1 ALC2 ETI LLOC 2 2 R1 62 1.95E 08 .1 SE E12 CCW 3 3 S2 63 1.87E 08 .1 TE ET7 STBL 31 8 EH2 OA2 EH5 1.69E-08 5 R2 R3 64 .1 SLC1 ET8 SLOC 3 65 1.61E 08 .1 TEC2 ET6 LOSP 12 1 AF OAl 2.93E 05/ year Total Core Damage frequency !
r i
r / . : ., , APR .23 '91 19:12 FROM LICENSING TO 104 SEQUQY FAGE.019 - a Page 17 of 29 ., Table 6 Dominant Sequences - A0T 7 Days { Core Damage Percent Damage Event Tree Sequence Support State frequency State Number and failed Nodes 1 8.22E-06 28.0 SLC2 ET3 SLOC 2 1 R2 2 1.97E-06 6.7 TEC2 E16 LOSP 12 ; AC OA1 3 1.85E 06 6.3 SE ET8 SLOC 26 7 HP OA CS 4 1.75E 06 6.0 TEC2 ET7 STBL 22 8 EH2 LH4 5 1.44E-06 4.9 1LC2 L15 1RAN 13 1 Af NF DAl 6 1.31E 06 4.5 SLC2 ET8 SLOC 2 1 R2 7 1.19E-06 4.1 SLC2 ET3 SLOC 2 2 R2 8 9.39E 07 3.2 ALC2 ET2 HLOC 2 1 R2 9 8.60E 07 2.9 V E14 Y-SE 1 1 10 7.95E-07 2.7 SE ET8 SLOC 26 4 HP OA CS 11 7.64E-07 2.6 SLCl ET3 SLOC 3 1 R2 R3 12 6.59E-07 2.2 TE ET7 STBL 26 8 EH2 EH4 EH7 13 6.43E-07 2.2 TEC2 EIS ATWS 7 1 RTl 0A3 14 6.24E-07 2.1 TEC2 ET7 STBL 46 3 AFT EH1 15 6.07E 07 2.1 TEC2 E19 ATWS 20 1 RTl PL PR 16 4.73E-07 1.6 AEC2 E12 ML6v 17 1 AC 17 4.28E-07 1.5 TEC2 ETS TRAN 13 5 AF NF OAl 18 3.16E-07 1.1 TEC2 ET9 ATWS 16 1 RTl PL OA3 19 3.06E 07 1.0 TEC2 E14 1RAN 12 1 AF OAl 20 2.93E 07 1.0 ALC2 ET) LLOC 2 1 R1 21 2.79E 07 1.0 SLCl ET3 SLOC 3 2 R2 R3 22 2.07E-07 .7 SEC2 ET3 SLOC 31 1 RT 23 1.94E-07 .7 TEC2 ET9 ATWS 28 2 RTl PL AM
l .. #,C AGR E3 '91 19 13 F R011 L 1 CE115 IllG To TUR SEQUOY PAGE.020 L l Page 18 of 29 ; Table 6 (Continued) Dominant Sequences - A0T 7 Days j Core Damage Percent Damage Event Tree Sequence Support State Frequency State Number and failed Nodes i 24 1.89E-07 .6 SE E18 SLOC 26 8 HP OA CS 25 1.78E-07 .6 AEC2 ET1 LLOC 11 1 AC . 26 1.68E-07 .6 SLC2 ET8 SLOC 2 2 R2 27 1.66E-07 .6 TEC2 ET9 ATWS 28 1 Ril PL AM 28 1.56E 07 .5 TE ET7 STBL 50 8 AFT EH1 EH3 ! 29 1.47E 07 .5 TEC2 E10 LOSS 12 1 AF OA1 30 1.38E 07 .5 SEC2 ET3 SLOC 23 1 HP OA 31 1.31E 07 .4 ALC2 ET2 MLOC 2 2 R2 ! 32 1.21E-07 .4 SLC1 E18 SLOC 3 1 R2 R3 ! 33 1.06E-07 .4 SE ET7 STBL 21 8 EH2 S2 CU l 34 1.06E-07 .4 TLC 2 ET5 TRAN 4 1 AF NF R2 > 35 1.01E-07 .3 SE E18 SLOC 26 6 HP OA CS ! 36 9.64E-08 .3 TE ET6 LOSP 15 6 AF OA1 CS t 37 8.73E-08 .3 ALCl ET2 MLOC 3 1 R2 R3 38 7.45E 08 .3 TEC2 ET7 STBL 27 8 EH2 OA2 , 39 7.25E-08 .2 SLC2 ET8 SLOC 2 5 R2 l 40 6.38E 08 .2 TEC2 ET5 TRAN 13 2 AF NF OA1 - i 41 6.09E-08 .2 AEC2 ET2 MLOC 13 2 HP OA : 42 5.88E 08 .2 TLC 2 E15 TRAN 4 2 AF NF R2 ; 43 5.32E-08 .2 AEC2 ETI LLOC 7 1 LP 44 4.45E-08 .2 SE ET8 SLOC 26 3 HP OA CS ! 4v 4.42E-08 .2 TEC1 ET6 LOSP 13 5 AF OA1 R3 l 46 3.95E-08 .1 SLC1 ET8 SLOC 3 2 R2 R3 ;
- i
APR,A3 'si 9: 14 rA0ti L 1 C E t451 f 46 TO TUA $EhU0h PAGE. Obi Page 19 of 29 Table 6 (Continued) Dominant Sequences - A0T 7 Days Core Damage Percent Damage Event Tree Sequence Support State Frequency State Number and failed Nodes 47 3.68E-08 .1 TE ET6 LOSP 15 5 AF OAl CS 48 3.55E-08 .1 TE ET9 ATWS 31 6 RTl PL AM CS 49 3.29E-08 .1 SEC2 ET8 SLOC 31 1 RT , 50 3.26E 08 .1 ALCl E11 LLOC 3 1 R1 R3 , 51 3.08E 08 .1 ALCl E12 MLOC 3 2 R2 R3 52 2.99E-08 .1 AEC2 ET) LLOC 7 2 LP 53 2.73E-08 .1 SE E12 CCW 3 3 S2 54 2.69E-08 .1 St E13 SLOC 6 2 CS R2 55 2.40E 08 .1 SE E13 ERCW l 4 56 2.37E 08 .1 AEC2 ET2 HLOC 21 1 RT 57 2.36E-08 .1 TEC2 E14 1RAN 12 5 AF OAl 58 2.24E-08 .1 TLC 2 ET4 TRAN 3 1 AF R2 59 2.20E 08 .1 SEC2 ET8 SLOC 23 1 HP OA 60 2.07E-08 .1 SLWCl ET3 SLOC 4 1 DR 61 1.96E-08 .1 TEC2 ET9 ATWS 11 1 RTl AF 62 1.95E 08 .1 ALC2 ETI LLOC 2 2 R1 63 1.87E-08 .1 TE ET7 STBL 31 8 EH2 0A2 EH5 64 1.70E-08 .1 SLCl E18 SLOC 3 5 R2 R3 65 1.61E 08 .1 TE:2 iT6 LOSP 12 1 AF OAl 3 L, y
- 2. 93 E - 05/yr,s. Total Core Damage frequency
, APR.23 '91 19 14 F R 0ll L l r El4S il TO TVA SEQUOY PAGE.0E2 Page 20 of 29 i Table 7 ! Dominant Sequences - Base Case Core Damage Percent Damage Event Tree Sequence Support State Frequency State Number and failed Nodes 1 8.22E 06 28.1 SLC2 ET3 SLOC 2 1 R2 2 1.97E-06 6.7 TEC2 ET6 LOSP 12 5 AF OA1 3 1.85E 06 6.3 SE ET8 SLOC 26 7 HP OA CS 4 1.75E-06 6.0 TEC2 ET7 STBL 22 8 EH2 EH4 5 1.44E-06 4.9 TEC2 ET5 TRAN 13 1 AF NF OA1 ! 6 1.31E-06 4.5 SLC2 ET8 SLOC 2 1 R2 7 1.18E 06 4.0 SLC2 ET3 SLOC 2 2 R2 8 9.39E-07 3.2 ALC2 CT2 MLOC 2 1 R2 9 8.60E 07 2.9 Y E14 V-SE 1 1 r 10 7.95E 07 2.7 SE ET8 SLOC 26 4 HP OA CS : 11 7.64E 07 2.6 SLCl ET3 SLOC 3 1 R2 R3 12 6.59E 07 2.2 TE ET7 STBL 26 8 EH2 EH4 EH7 13 6.43E-07 2.2 TEC2 ET9 ATWS 7 1 RT1 OA3 14 6.24E 2.1 TEC2 ET7 STBL 46 8 AFT EH1 ! 15 6.07E-07 2.1 TEC2 ET9 ATWS 20 1 RTl PL PR ; 16 4.73E-07 1.6 AEC2 ET2 HLOC 17 1 AC - 17 4.28E-07 1.5 TEC2 ET5 TRAN 13 5 AF NF 0A1 r 18 3.16E-07 1.1 TEC2 ET9 ATWS 16 1 RTl PL 0A3 19 3.06E 07 1.0 TEC2 E14 TRAN 12 1 AF OA1 ; t 20 2.93E-07 1.0 ALC2 ET1 LLOC 2 1 R1 ; 21 2.77E 07 .9 SLC1 ET3 SLOC 3 2 R2 R3 22 2.07E-07 .7 SEC2 ET3 SLOC 31' 1 RT e I
. APR 23 91 19115 F R oll LICEt45 f tM TO TVA SEQUOY PAGE.023 Page 21 of 29 l Table 7(Continued) ! Dominant Sequences Base Case Core Damage Percent Damage Event Tree Sequence Support State ; Frequency State Number and failed Nodes i 23 1.94E-07 .7 TEC2 ET9 ATWS 28 2 RT1 PL AM 24 1.89E 07 .6 SE ET8 SLOC 26 8 HP OA CS 25 1.78E 07 .6 AEC2 CT1 LLOC 11 1 AC ! 26 1.67E 07 .6 SLC2 ET8 SLOC 2 2 R2 f 27 1.66E-07 .6 TEC2 ET9 ATWS 28 1 RTl PL AM i 28 1.56E-07 .5 IE E17 STBL 50 8 AFT EH1 EH3 29 1.47E-07 .5 TEC2 E10 LOSS 12 1 AF OA1 t 30 1.38E 07 .5 SEC2 ET3 SLOC 23 1 HP OA 31 1.32E-07 .5 ALC2 ET2 MLOC 2 2 R2 32 1.21E-07 .4 SLC1 ET8 SLOC 3 1 R2 R3 ( 33 1.06E-07 .4 SC ET7 STOL 21 8 EH2 S2 CU 34 1.06E-07 .4 ILC2 ET5 TRAN 4 1 AF Nr R2 l 35 1.01E-07 .3 SE ET8 SLOC 26 6 HP OA CS 36 9.64E-08 .3 TE E16 LOSP 15 6 AF OAl CS 37 8.73E-08 .3 ALC1 ET2 MLOC 3 1 R2 R3 38 7.45E-08 .3 TEC2 ET7 STOL 27 8 EH2 OA2 ' 39 7.21E-08 .2 SLC2 ET8 SLOC 2 5 R2 40 6.38E-08 .2 TEC2 ET5 TRAN 13 2 AF NF 0A1 [ 5.85E-08 .2 2 AF NF R2 41 TLC 2 ETS TRAN 4 42 5.32E 08 .2 AEC2 ET) LLOC 7 1 LP l 43 4.45E-08 .2 SE ET8 SLOC 26 3 HP OA CS 44 4.42E-08 .2 TEC1 ET6 LOSP 13 5 AF OA1 R3 f 45 4.38E 08 .1 AEC2 E12 HLOC 13 2 HP OA I v.----+_ ~ - . _ _ _ _ _ _ . _ _ _ _ - _ - _ _ _ _ . - _ _ . _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
. RPR C3 '91 19 16 F R Ot1 L I C El45111G TO TVA SEQUOY PAGE.024 i Page 22 of 25r ;
Table 7 (Continued) Dominant Cequences -- Base Case Core Damage Percent Damage Event Tree Sequence Support State Frequency State Humber and failed Nodes . 46 3.93E 08 .1 SLC1 ET8 SLOC 3 2 R2 R3 47 3.68E-08 .1 TE ET6 LOSP 15 5 AF OA1 CS 48 3.55E 08 .1 TE ET9 A1WS 31 6 RT1 PL AM CS 49 3.29E-08 .1 SEC2 ET8 SLOC 31 1 RT ! 50 3.26E-08 .1 ALCl ETl LLOC 3 1 R1 R3 i F
$1 3.10E-08 .1 ALCl ET2 MLOC 3 2 R2 R3 [
52 2.99E 08 .1 AEC2 ET) LLOC 7 2 LP
^
53 2.68E-08 .1 SL ET3 SLOC 6 2 CS R2 54 2.40E 08 .1 SE E13 ERCW 1 4 55 2.37E 08 .1 AEC2 ET2 MLOC 21 'l RT I 56 2.36E-08 .1 TEC2 ET4 TRAN 12 5 AF OA1 57 2.24E-08 .1 TLC 2 ET4 TRAN 3 1 AF R2 58 2.20E-08 .1 SEC2 E18 SLOC 23 1 HP OA 59 2.07E 08 .1 SLWC1 E13 SLOC 4 1 DR 60 1.96E-08 .1 TEC2 ET9 ATVS 11 1 RTl AF 61 1.95E-08 .1 ALC2 ETl LLOC 2 2 R1 62 1.87E-08 .1 TE ET7 STBL 31 8 EH2 OA2 EH5
, 63 1.69E-08 .1 SLC1 ET8 SLOC 3 5 R2 R3 64 1.61E 08 .1 TEC2 E16 LOSP 12 1 AF OAl 65 1.43E-08 .0 SE E12 CCW 3 3 52 2.93E-05/ year Total Core Damage Frequency .
[ ppg, z.2 est te:53 FROM LICENSING TO TVA SEQUOY PAGE.00G Page 23 of 29-l Table 8 Core Damage Frequency Results by Initiating Event initiating Core Damage Percent Conditional Event Frequency Initiating Event of Total PrcDability Frequency '
- 3 Day A0T - =
Large LOCA 6.28E-07 2.1 % 2.09E-03 3.00E 04 Medium LOCA 1.77E-06 6.0 % Small LOCA 2.21E-03 8.00E 04 1.09E 05 37.2 % 1.56E 03 7.00E-03 Transients 3.65E 07 1.2 % 9.61E-07 3.80E-01 Transient (no NF) 2.16E 06 7.4 % 3.13E 07 6.90E+00 LOSP 2.16E-06 7.4 % 8.65E-05 2.50E-02 Station Blackout 3.40E 06 11.6 % 1.36E-04 Small LOCA
- 2.50E-02 4.78E 06 16.3 % 4.78E-06 1.00E+00 '
ATWS 2.03E 06 6.9 % 2.03E 06 1.00E+00 Loss of AC Bus 1.92E-07 .7 % 9.61E 07 Loss of DC Bus 2.00E-01 1.18E-08 .0 % 6.54E-05 1.80E 04 Loss of CCW l.96E 08 .1 % 1.51E-02 1.30E-06 Loss of ERCW 2.40E 08 .1 % 1.00E+00 V-Sequence 2.40E 08 8.60E-07 2.9 % 1.00E+00 8.60E 07
- 7 Day A0T -
Large LOCA 6.28E-07 2.1 % 2.09E 03 3.00E-04 ' Medium 1.0CA 1.78E-06 6.1 % 2.22E 03 8.00E-04
- Small LOCA # este6 1.09E 05 37.2 %
Transients 3.65E 07 1.2 % 1.56E 03 9.61E 07 7.00E-03 l. 3.80E-01 Transients (No NF) 2.16E-06 7.4 % 3.13E-07 6.90E+00 LOSP - 2.16E-06 7.4 % 8.65E-05 c.50E 02 Station Blackout 3.40E-06 11.6 % 1.36E-04 Small LOCA 2.50E 02 4.78E-06 16.3 % 4.78E-06 1.00E+00 ATWS 2.03E-06 6.9 % 2.03E-06 1.00E+00 Loss of AC Bus 1.92E-07 .7 % 9.61E-07 Loss of DC Bus 2.00E 01 1.18E-08 .0 % 6.54E-05 1.80E 04 Loss of CCW 2.74E-08 .1 % 2.10E-02 Loss of ERCW 1.30E 05 2.40E-08 .1 % 1.00E+00 2.40E 08 V-Sequence 8.60E-07 2.9 % 1.00E+00 8.60E-07 f
FPR t ,23 '91 19:10 FR0f1 LICEl4Sil46 TO 1VA SEQUOY PAGE.026 Page 24 of 29 Table 8 (Continued) Core Damage Frequency Results by Initiating Event initiating Core Damage Percent Conditional Initiating Event Event Frequency of Total Probability Frequency Base Case large LOCA 6.28E-07 2.1 % 2.09E-03 3.00E 04 Medium LOCA 1.76E-06 6.0 % 2.20E 03 8.00E 04 Small LOCA 1.09E-05 37.2 % 1.56E 03 7.00E 03 Transients 3.65E-07 1.2 % 9.61E 07 3.80E 01 Transients (NoNF) 2.16E-06 7.4 % 3.13E-07 6.90E+00 LOSP 2.16E.06 7.4 % 8.65E-05 2.50E-02 Station Blackout 3.40E-06 11.6 % 1.36E 04 2.50E 02 Small LOCA* 4.78E 06 10.3 % 4.78E-06 1.00E400 ATWS 2.03E-06 6.9 % 2.03E 06 1.00E+00 Loss of AC Bus 1.92E 07 .7 % 9.61E 07 2.00E 01 Loss of DC Bus 1.18E-08 .0 % 6.54E 05 1.80E 04 Loss of CCW l.44E 08 .0 % 1.llE-02 1.30E 06 Loss of ERCW 2.40E-08 .1 % 1.00E+00 2.40E-08 V-Sequenc 8.60E-07 2.9 % 1.00E+00 8.60E 07 Total Core Damage Frequency for all three cases: 2.93E 05/ year Note: All frequencies are frequency / year Note: This is a consequential small LOCA. Note: The core damage frequency reported for each of the cases differs from that in Reference 1. An error was found while requantifying these Cases.
Page 25 of 29 I i i
5.0 REFERENCES
1- Saiquoyah Nuclear Plant Individual Plant Evaluation Analysis Notebooks, I Volume IV, Tab 2 Event free Models, Revision 0, April 27, 1988. l 2- Sequoyah Nuclear Plant Individual Plant Evaluation Analysis Notebooks, , Volume 111. Tab 2, High Pressure Injection and Racirculation System ! Notebook, Revision 0, April 29, 1988. j i i i i i [
,, l l
l I i i [ l l L k w - - - , , , , - .,,n,. _ - . _ . -, . . . - - - - . , _ _,,_ , .. . ._.., - - ---- -
. . AP,R os '91 19:10 F ROli L I C E f 4$ 114G TO TVA SEGUOY PAGE.020 +
5 Page 26 of 29 . 41AcilVITY CONTA0L 5Y$ffMS CHARGING PUMP 5 OP[RAf!NG (!HITING (ON0lfl0N FOR OPERAt!0N
. i 3.1.2.4 At least two charging Pumps shall be OPIAAllt.
APPL!CAllLITY1 MODES 1, 2, 3 and 4 ' AJ.712: 7 alay.s Vith CP(AABLE only one ntatus Charging within!'"pump " OPERA 84,(, re6 tore at least tvs charging pum
.$5hour Hut 00WN MARCIN equivalent to at letst it delta k/k_a", 200' r;; 4eee : t ? e*44- GDLU
[. ::t-?[nyl-ee-beJIn M c 'e ;' *gWWIDOWM ^ = : 10 ^*f**u'.5 ;;;67; within H " " the ^.h.9 next 30 ~ and
$URVI!Lt.ANCf RE0VIR[M(Nf5 4.1. 2. 4 At.least two charging 99eos shall be demonstrated 0* ERA 5LE cy verifying that on recirculation flow, each pump covelops a discharge pressure of0.5.
a greater than or equal to 2400 psig when tested pursuant to $pecification 4 SEQUOYAH UNIT 1 3/4 1-10 Amene. tent No. 1 A r. ..
'I lt' )
- l
. . - . . . . . ~ _ . _ _ _ _ -,m.-- . _ _ _ _ _ _ _ _ _ _ m __.-m_. _ _ . - _ _ _ _ _ - - . _ _ - - -
i
. ..AGR 83 '91 19:20 FROM LICENSING TO TVA SEQUDY PAGE.029 t
Page 27 of 29 EMFRGENCY CORE COOLING SYSTEMS (ECCS) l 1 3/4.5.2 ECCS SUBSYSTEMS - T, Greater Than or Equa] to 350*F_ i LIMITING CON 0! TION FOR OPERATION ! l 3 3.5.2 Two indepencont ECCS subsystems shall be OPERABLE with each subsystem j comprised oft a, One OPERABLE centrifugal charging pump, i t
- b. One OPERABLE safety injection pump, l
- c. One OPERABLE residual heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and [
- e. An OPEHABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically !
transferring suction to the containment sump during the recirculation ; phase of operation. APPLICABILITY: HODES 1, 2 and 3. l _due to safety injection pump, residual heat removal heat ! ACTION: exchanger, residual heat removal pump, or flow path inoperability [
- a. WithoneECCSsubsysteminoperable{restoretheinoperablesubsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY f within the next 6 hours and in HOT SHUTOOWN within the following 6 !
'N hours.
i
@ In the event-the ECCS is actuated and injects water into the Reactor Coolant System, a REPORTABLE EVENT shall be prepared and submitted i
i to the Commission pursuant to Specification 6.6.1. This report shall ; include a description of the circumstances of the actuation and the MO ; total accumulated actuation cycles to date. The current value of - the usage factor for each affected safety injection nozzle shall De [ provided in this report whenever its value exceeds 0.70. f SURVEILLANCE REQUIREMENTS ,
. F 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: ~
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the ' valve operators ;
removed: "
- b. With one ECCS subsystem inoperable solely due to centrifugal .
charging pump inoperability, restore the inoperable subsystem to ! OPERABLE status within 7 days or be in at least HOT STAN06Y within E the next 6 hours and in HOT $ HUT 00m wtthin the following 6 hours. l . SEQUOYAH UNIT 1 3/4 5-5 Amendment No. 36 ! I
SEQUOY PAGE.030 Page 28 of 29 RfACTIVITY CONTROL SYSTEMS CHARGING PtmPs - OPERATING LIMITING CONDITION FOR OPERATION
. ~
h3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: H00E5 1, 2, 3 and 4. M: 7cjgp With only one charging _oumo OPERA 8LE, restore at least two charging pumps to OPERA 6LE status withinu;' mr; or be in at least HOT STAN08Y and borated to a SHUT 00VM MARGIN equivalent to at least 1% delta k/k at 200'F within the next 6 ho u r sit-r4644 rea t ' a t s t tN cha r;' ^ t ; '?"4-t
- QP E m an ( g 3 g 3 g,3 m i g w 1.
g g _,p4, } %
>[7-daye-e+lin cGLD 5 HUT 00W within the next 30 hours.
onA SURVE!LLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERAaLE by \ verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 2400 psig when tested pursuant to Specification 4.0.5. NO E: With centrif chargir>g p inopera he one cooling s en (ECCS core rossin ble for an tional hours R' c res a n Ju , 9 . August 23, 1984 SEQUOYAH - UNIT 2 3/4 1-10 Amendment No. 27
APR'e3 *si isTa2 i k si R it Eli3 Iiis ~ ~ ~
~
es *
~10 Ti#i 5E'006 Pidi.031 Page 29 of 29 EMERGFNCY CORE COOLING SYSTEMS j L
3/4.5.2 ECCS SUBSYSTEMS - T, g_ Greater Thar or Equal to 350'F_ i t!MITING CON 0! TION FOR OPERATION i
)
\
- h. 5. 2 Two independent emergency core cooling system (ECCS) subsystems shall h be OPERABLE with each subsystem comprised of: l l
- a. One OPERABLE centrifugal charging pump,
~
- o. One OPERABLE safety injection pump,
- c. One OPERABLE residual heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2 and_ 3. I due to Safety injection pump, residual heat removal heat AC T ION'. exchanger, residual heat removal pump, or flow path inoperability (
- a. WithoneECCSsubsysteminoperabl/,restoretheinoperablesubsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within
~_ the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
In the event the ECCS is actuated and injects water into the Reactor
# @ Coolant System, a REPORTABLE EVENT shall be prepared and su to the Commission pursuant to Specification 6.6.1. This report shall b include a description of the circumstances of the actuation and the R28 total accumulated actuation cycles to date, .The current value of the usage factor for each affacted safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
SURVE!LLANCE REQUIREMENTS L 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: J
- a. At least once per 12 i s s by verifying that the following valves are in the~ indicated posit u s with power to the valve operators removed:
I
- b. With one ECCS subsystem inoperable solely due to centrifugal -
charg_ing pump inoperability, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STAND 8Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. kvember 23, 1984 t SEQUOYAH - UNIT 2 3/4 5-5 Amendment No. 28 l
+4 T A T 01 p o r, r A71 44}}