ML20236C396
| ML20236C396 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/15/1987 |
| From: | Mileris G BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20236C394 | List: |
| References | |
| FOIA-87-643 2144, 2144-R, 2144-R00, NUDOCS 8710270144 | |
| Download: ML20236C396 (30) | |
Text
{{#Wiki_filter:htE Proposed Change- Safety Evaluation No.: 2. M4 6 REV. O SHEET l OT // PILCRIM NUCLEAR P N ER STATION Rev. No. , PDC PCM Systee Calc. Initiator: Dent: Group: No.: Name: No : G. V. Mileris NED FSMC 86-51 Direct Torus Vent Date:kJ5fd9 Provide a direct torus Description of Pronesed venE tochange,in the ma stacktest or exp(eriment: DTVS) SAFETY EVALUATION CONCLUSIOlt$: The proposed change, test or esperiment:
- 1. (X) Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment taportant to
- safety previously evaluated in the FSAR.
- 2. f) Does Not ( ) Doe's create the possibility for accident or malfunction of a different type than any evaluated previously in the FSAR.
- 3. 6) Does Not ( ) Does reduce the margin of safety as defined in the basis ,
for any technical specification. , RASIS FOR SAFETY EVAlt1ATION 0110L0$1(18S: SEE ATTACHED SHEETS Q ange - Gange N Recommended ( ) Not Recommended SE Performed by Date M M 7 Exhibit 3.07-A Sheet 1 of 3 *
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S&7e Q Leaseation SAFETY EV4tt1ATION fNGF r~ A o f / / PILCRIN NutttAR PCkER STATICM q Rev. No.o A. APPROVAL This proposed change _ involve a change in the Tec,hnical
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Specifications. This proposed change, test or experiment does ( ) does not. 0d 'J involve an unreviewed safety question as defined in 10CFR, Part 50.5f(a)(3). e to the FSAR per 10CFR
/ This proposed change involves a ch 50.71(e) and is portable und R50.5 .
Use o the vent eyond e scope of this safety Comments: .. evaluation and will be covered in the emergency operating procedures. A separate safety evaluation I will be written to cover use of the vent. 1 l The safety evaluttien basis and c
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. A. _OISCRIPT;0N 07 PF.0POSID CHANGE This modihcatien p:ovides a direct vent pathTreatment from theSystem torus bypassing to the main stack on the torus purge exhaust line. The bypass is the Standby Gas (SGTS) equipment an't" line whose upstream end is connected to the gipe between primary containment isolation valves A0-5042A&B (on 8 portion of 20" torus purge exhaust line to SGT8). The downstrema and of the bypass is connected to the 20" main stack line downstream of 8078 valve (AO-5025),
i valves A0N-108 and A0N-112. An 8" butterfly which can be remotely o erated from theThis main control roca, is-valve acts as the added downstream of 8 ' va ve A0-50425. l primary containment outboard isolation valve fortothe direct and torus inclusive g vent line. The new pipe is ASME III class 2 up l l' "N* ^$$ *h : id M!!$N?#!!!EWill.is.h. ars. prov1ED opsM e4 k m s+rso m e4 no -sess' .AC solenoid 4&5
)x The proposed modification replaces solenoid the existing valve (powered from for A0-5042B with a DC valve essential 125 volt DC) to ensure valv operability during a station I blackout. The new isolatio 0-5025, is also provided with a DC solenoid powered from 125 volt DC ocurce. M 4 One inch nitrogen lines are Both of these valves' fail closed. and A0-5025.
added to provide backup nitrogen to valves A0-5042B being modified to override The present logic of A0-5042B is remote manual action. , containment isolation signalson by containment keylock isolation signals but is New valve A0-5025 closes override control logic as A0-provided with the same When 5025 isolationand 5042B are in the containment isolation 5042B. baan..added to isolate both bypass mode a separate logic has in the Torus vapot valves if there is a high radiadiin~)evelbd C adeoiiplit e4 ) high' radiation overrits is Dal space. This keylock remote manual action. 4 , , gg@
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A 20" pipe will replace the e9 " L, 2; u s ever aucti between SGTS valves A0N-108, A0N-112 and the existing 20" pipe to the main stack. The existing 20" diameter duct downstream of A0-of the new vent line branch 5042A is shortened to allow fitup connection. A rupture disk will be included in the 8" piping downstream ; of valve A0-5025. The rupture disk will provide a second3 leakage barrier. The rupture disk is designed to open at a pressure remain intact during below' direct venting pressure, but 'will 4 normal operating conditions. This safety Evaluation demonstrates that plant safety during
- normal design basis operation is not degraded by the installation *S of the direct torus vent. Vent operation is not ad r pg by tha 0,,1<els art snumy& so 1%f %< vslits wlIl ow+[reakk isoladim $pis ress1 u>s'Lat- o So- <t.%LtbothYsivas vs\ver.
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s SAFETY EVALUATION NO.2 W REV. NO. O SHEET g OF g safety evaluation but will be addressed by -another safety conjunction with the Energency operating Evaluation in Procedures . > Although this safety evaluation does not address the vent operation, a description of the venting operation and logic of l the installed modification is provided below for information. l The purpose of the vent is to relieve excess containment ~ ressure and prevent catestrephic containment failure as directed will be p:ry the Emergency Operating procedures. Use of the vent control and would require that under management adxtinastrative the keyltek switches for operation of the A0-50425 and A0-5025 valves be placed in the Emergency Open Position to override the , l containment isolation signal which would be present if the containment was at high pressure. Prior to opening the vent ! valves the 830T system would have to be shutdown and valves AON-108 and AON-112(the outlet of SBGT)placed in a closed Ao- 0Hfosition, B and M g . If there is high radiation in the torus vapor spaceThatisolation signal can be overr4dden b A0-5025 will reisolate.
- get o k- sa=;ual keylock actuated switches if vamtink,1p, y.centinue.uw tx ,
- 5. PURPOSE OF THE CHANGE
! CONSTRUCTION ---
This modification will provide t b hility te sof_Issa of the severe accident concerns by direct venting of the torus an to prevent- primary containment over-pressurization during use can be justified this extended station blackout. If ' ~ modification opens a direct exhaust path from the torus vapor space to main stack. , 4 C. SYSTEMS, SUBSYSTEMS,COMPONENTi{AFFECTED Atmospheric This modification affeets> the containment control System in the following manner: inboard isolation valve A0-The torus purge exhaust linel o' pipe are the components ofWith the
$042B and the associated psipposed modification.
CAC8 affected by this subject modification, the CACS Will incorporation of the valve A0-5042A) and depend on both essential AC (for essential DC (for A0-50425) to perform itsgn gn. p i The new 8" torus vent line will be connected to ex sting 8" g CACS piping between valves AD-5042B and A0-5042A. Atandby Gas Treatment System
- This modification nffects the in the following manner:
p SAFETY EVA!.UATION No.2I4i REV. NO. 0 SMEET 6 0F,jf SGTS fan outlet valves (AON-108 & AON-112), ductdork from f these valves to the-20" line leading to the. main stack, and l the 20", line leading to main stack are the. components of { this system affected by the proposed modification. Valve A0N-108 is. norma 11y clesed, fall-open. Valve A0N-112 1 will be revised to be normally closed, fail-closed,.and l these valves will be .provided with essential DC power and i local safety related nitrogen supplies (Ref. PDC 86-70). ] This modification affects the Primary Containment Isolation system (PCIS)-in the following manner: This system is affected by the modification to containment ' isolation valve A0-5042B logic. The addition of containment 4 I outboard isolation valve (AO-5025) 'and associated controls will also affect the PCIS. l D. SAFETY FUNCTI0tt OF AFFECTED SYSTEM / COMPONENTS Containment Atmospheric Control system obviating the This system has the safety function of primary possibility of. an energy .I8 lease..within the following a containment from a Hydroger c^0xygen postulated 1,0cA combined with %egrdok.rgactionQpre Jst.andbyi. cool e Eng
' " " " * ~' ! CONSTRU.CTION 8tandby Gas Treatment System the reactor building This system filters exhaust air from the main stack. The and discharges the processed air to particulate and iodines.from the air stream I
system filters in order to reduce the level of airborne contamination released to. the environs via the main stack. The SGTS can also filter exhaust air from the drywell and the suppression pool. Primary Containment Isolation System This system .has the safety. function of providing timely and consequences of accidents
.protection involving. against the onset-release of radioactive materials from rhe._ gross i the primary containment by initiating automatic isolation.of primary appropriate pipelines which penetrate the containment whenever monitored variables exceed pre-selected operational limits.
Pr. m 1 f u % d h b W The primary containment system, in conjunction with other safeguard features, limits the release of fission products in the event of a postulated design basis accident so that offsite doses do not exceed the guideline values of 10CFR100.
1 s' SAFETY EVALUATION NO 28 % REV. NO. O SHEET g OF g E. EFFECTS ON SAFETY FUNCTION Containment Atmospheric Control System and' Standby Gas Treatment System, an0 Primary Containment Isolation Bysten " The modification changes the solenoid A0-5042B control from AC to DC enabling it to position) when open (from its normally closed blackout. required even during extended station Ductwork at the cutlet of the 83GT system is replaced with pipe and the new vent line is connected to the 20" line at the outlet of the 830T system. a Addition of a new 8" vent line with containment isolation valve A0-5025 off the torus purge and vent line results in a flow path that could vent the containment directly to the stack bypassing the eBoT system during nonnal plant operation. New logic is being added that allows override of the containment provides isolation signal on existing valve A0-50423 and the same logic for valve A0-5025. This could allow venting of the containment directly bypassing 8BGT or subjectiFg--the SBGT s to the stack to high containment pressure. lS$y{}ystem { gg l
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F. ANALYSIS OF EFFECTS ON t 8AFETY FW@ff089f$fa)
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An analysis of the effects on safety siumi.4wn. -S, SGTs, and. PCI8 systems for the installation of the direct torus vent is described as foll Ws: ' The' change from At to DC control on A0-50423 does not adversely affect its ability to o AO-50425 when be containment is be.ing purged, or iseWre_ pen vwA.u.stes M e M.y The modificatA>.as to the ductwork and 20" line leading to the main ste a do not affect the safety function of any of the safety related systems. During normal plant. operation, the CAC8 and the 8GT8 do not use the torus 20" purge and vent line to. perform their . safety functions. ~The containment isolation valves are in their normally closed position, satisfying the safety ! function of the PCIS. - - A m ,m oam .ma , m % ec.-,. ~b- t i g W. M. + . .w .GW DTvs % e kh "" Qd-wb .. n u..p A.e.- -a a - - +~ u
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[ SAFETY EVALUATION No. 2 44
. RIV. No. 0 SHEET 7 OFZ During plant (nw-excpq C<mO~.%& W startup and shutdown"when the purge and vent line is in use, the logic does not allow A0-Ib2.s' to open M ,
unless S&f2 S is closed. In addition a rupture disc W downstream of valve A0-5025 means of- preventing leakage will provide a second positive' l and prevent the stack in theevent of a single failure ofdirect release A0-5025 up during containment purge and vent at plant startup or shutdown. During containment high pressure conditions, the CAC8 does not use the torus main exhaust line to SGT8 for performing its safety function. communicate to the logic cannot override the containment isolation The existing CAC8 signal and open valves A0-5042A, ett A single failure of A0-50423 logie M would still protect the BBGT from high pressure because A0-5042A, and A0-5025 would still be closed. Valve A0-5025 and the ; rupture disk downstream would also prevent any inadvertent discharge up the stack. G.
SUMMARY
The installation of the Direct Torus Vent (DTV8) does not affect the safety functions of the CACs, 50Ts, or any other safety-related systems. This safetyand PCIs systems, evaluation does not provide justification for use of the DTV8. The installation of this ' Technical specification Change. modification does require a The installation of this modification does not involve an unreviewed safety question. ; ISSUE 3r$4 i CONSTRUCTION
SHEETjf Or // Safety Evaluation No.: 2.i44
- SAFETY EVALUATION WMtK SHEET Rev. No. O A. Systee Structure Component Failurs and Consequence Analyses.
Systee Structure h onent Failure Modes Effects of Failurt Commen'ts SEE ATTACHED SHEET . 1.
- 2. ,
- 3. j
-. l l
i 1 General Reference Material Review CALCULATIONS REQlLATORY ; FSAR l SECTION PNPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GLilDES STANDARDS , SEE ATTACHED SEEET I l l, s t~ t* ) p~ . s i v v ;., s. ; .h e n s., e c r. 3. . '. .....,mm, gy T w 1 ius u, a IVW B. For the proposed hardware chanfle, identify the failure modes that' are likely for the components. cons' stent with FSAR assumptions. For each - failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each systes) or plant safety functions FSAR Chapter 14 and Appendtx G). Prepared by . Date N NOTE: It is a requirement to include this work sheet with the Safety - Evaluation. Exhibit 3.07-C l 3.07-18 Rev. 4 l l
.t SAFETY EVALtJATION No. Il##
REV. No. 0 SHEET 9 OF /f SAFETY EVALUATION WORY~3 REFT l I' . i Syst em/6tructure/(.omponent : Isolation Valve A0- 5025. #'^ l Failure Modes Faihure to close/F41
- Pu W G *- P"* Y "'k di '
Errects of Failures Bypassing of Standby Gas Treatment System; Loss of containment Isolation comments: lE Valve position indication is provided in the main control room. Rupture disk provides protection. l Failure Mode: Excessive Leakage Egrects of Failure: Release of Radioactive material above
. allowable.
l Comments: Periodic LLRT performed to ensure leak tightness. Rupture disk provides protection. (Rupture disc is tested periodically) system /8tructure/ Component: Direct Torus vent System piping. Failure Mode: Structural tailure. and Effects.of Tailures Loss of containment integrity SGT8 inoperable. Comments Qualify piping' for design basis temperature and pressure to ASME III and 331.:.. System / structure / Component: ' Primary Containment Isolation System. Failure Moder Logic failure. Efracts or Failure: No effect. Comments: Recuncant trains of logic y am)(vg/w<. dl4 k, M
. . ISSU D n CO N STi'...a iU-[
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Safety Evaluation No. 214'y , Rev. No. 0 ' Sheetf oofft SAFETY EVALUATION WORKSHEET i FSAR CALCULATIONS /PNPS REGULATORY DESIGN GUIDES / STANDARDS SECTION TECHNICAL SPECS. SPECS / PROCEDURES CODES i i
- 5.2 3.7/4.7 . Spec. SM-34 10CFR50 Spec. M300 NUREG CR-624 l 5.3 Spec. M600 NUREG BMI-139, Vol 1
- Spec. M301 i 5.4 Spec. M600 NUREG 8MI-139, Vol 1 Spec. M-611 NUREG 0700 i 7.3 Specs. 17322-M SAM-16 ASME B&PV Code Sections 10.11 17322-M-SAM 12 III & XI !
Table 7.31 Specs M-603, M303 Appendix G Spec. E-347 R. G. 1.26 R. G. 1.97 Appendix 0 Spec. E-347 A IEEE-79 IEEE-23 - IEEE-44 ANSI B31.1 l Calculation Number i Teledyne Calc./Later , - 17322-M-640-1 - 571-28-17322 and 571-29-17322 N- gUC! L', I' M ' M'r- CONSTRUCTION l 10394-115-C3 17322-640-C120.0 17322-630-C200.0 , 17322-640-C100.3 4 : - 17322-640 C110.0 17322-640-C100.5 4
Safety Evaluation No. 214 W Rev. No. O Sheet ll of II . PILGRIM STATION l FSAR REVIEW SHEET
References:
PDC 86-51 Support a change to provide a Direct Torus Vent Line. List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision. ' Affected FSAR Revision to affected FSAR Section is shown on: Section Preliminary f_igal Section 5.3.3 Attachment 1
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Section 5.3.4 Attachment 1 i T-Section 5.4.1 Attachment 1 j r iJ;b.bt~~i g.W Section 5.4.7 Attachment 1 m.
,1 Section 5.4.8 Attachment 1 i Attachment I u Section 7.3.4 Table 5.2 4 Attachment 1
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STRUCTloIN Attachment 1 4 Table 5.2-5 Table 7.3-1 Attachment 1 . Figure 5.2-16, 5.4-1 M 227, Oht.1 Figure 10.11-1 M 220, y,t.1 In addition, the PHPS Technical Specification. Table 3.7.1, Notes for Table 3.7.1, Bases 3.7D/4.7.D. PRELIMINARY FSAR REVISION (To be complet$d at time of Safety Evaluation preparation.) Prepared by: M A /Date: 66; d 2 Reviewed by: X ate: AIf/ Approved by: /Date:If/87 FINAL FSAR REVISION (Prepared following operational turnover of related systems structures of components,for use at PNPS.) (1) ,
/Date: Reviewed by: /Date:
Prepared by: (1) Attach completed FSAR Change Request Form (Refer to NOP). Exhibit 3.07-A Rev. 3 (Sheet 2 of 3)
w i Safety Evaluation No. 21 ## Rev. No. O Attachment 1 . Sheet 1 of M /9 ATTACK 1ENT 1 , RECOMMENDED FSAR CHANGES The pages of the following sections, tables, & figures of the FSAR that need to ' be updattd due to this modification (PDC 86-51) have been marked with suggested updates and included in this attachment for your review. FSAR Sections: 5.2, 5.3, 5.4, 7.3, 10.11 FSAR Tables: 5.2-4, 5.2-5, 7.3-1
'The tallowing drawings will be revised as part of the Plant Design Change package (PDC 86-51) but are not included herein.
Dwg. I. D. FSAR' FIGURE TITLE M 227 Sht. 1 10.11-1 P&lD Containment Atmospheric Control System M 220, Sht. 1 5.2-16, 5.4-1 P&ID Compressed Air System M 294 5.2-17, 7.18-2 Heating Ventilation and Air-Conditioning Standby Gas Treatment System Control Diagram ISSUEO eUn CONSTilUCTION N j
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1 Rev. No. O I Attachm nt !, Sheet 5 of 19 l f' ,
scienolds c' ucon loss cf instrument air to the air operators on the {
ca cers. The demister is designed to remove entraincd water droplets and mist from the entering air stream. The electric heating coil is designed An to reduce the relative humidity of the air stream to 80 percent. interlock with its associated exhaust fan prevents the heating toll { from operating when the f an is shut down. Each HEPA filter is designed to be capable of removing at least 99.97 percent of the 0.30 micron particles which impinge on the filter. The charcoal l l filters are lodide-impregnated activated carbon filters capable of removing in excess of 99 percent of the lodine in the air stream with 10 percent of the iodine in the. form of methyl iodide (CH 1) under 3 l entering conditions of 80 percent relative humidity. using the standard NRC approach are ; The accident evaluations described in Section 14.9. 'In these analyses the SGTS charcoal l filter! were credited with removal of 95 percent of the influent lodine. The system will start automatically upon a high radiation signal frem the operation (refueling) floor ventilation exhaust duct monitor, or upon receipt of high drywell pressure or low reactor water level signals. The system can also be manually started from the control room. Upon receipt of any of the initiation signals, both fans ! start, all SGTS isolation dampers open and each fan draws air from the isolated Reactor Building at a flow rate of approximately ,c 4,000 ft'/ min. After a preset time delay, one fan is stopped. I Cross-connections between the filter trains are provided to maintain the required decay heat removal cool ny eW rflow on the charcoal 1 filters in the inactive treatment trLin. Q tDiyistharge; to the main stack through a 20 in underyotnd rftf. ftie- SGt Vernt\are powered from the emergency service porGGNSTRt!"* TION distribution system. L._- - i Drywell and torus purge exhaust can also be directed to the SGTS The for processing before release up the main stack. Seegland Section 5.2. l High Pressure Coolant Injection System (HPCIS) seal steam condenser exhauster discharge is also routed to the SGTS during , accident conditions. The Reactor Buildina Neatino and Ventilating _$
. System it ffitrutsed in Caetion 10.9'i During a severe accident , the 5 l
{ torus can be directly vented to the main stack bypassing the SCTS. p ' 5** Section 3 4 7-5.3.3.5 Main Stack I The location of the main stack is shown on Figure 1.6-1. The top of The structural design of the the stack is at elevation 400 f t msl. stack is discussed in Section 12. , 5.3.4 . Safety Evaluation The SCS provides the principal mechanisms for the mitigation of the consequences of an accident in the Reactor Building. The primary and principal secondary containment act together to provide the mechanisms for the mitigation of the consequences of an accident in 5.3 4
sev. .w . v , l Attechment 1. Sheet 6 of 19
- the crywell. If the leakage rate of the building is low, and the
, leatage air is filtered and discharged to the elevated release peint (utilizing the SGTS and the main stack) the offsite radiation ecses that result from postulated accidents are reduced significantly. The Rea: tor Building is a Class I structure (with the exception of the secondary containment access locks which are Class IlDesign of the structures) designed in accordance with all acclicablerate codes.
of 4,000 f t'/ min at Reactor Building for a maxirum inleakage a building subatmospheric pressure cf 0.25 in of water at neutral wind conditicns, results in a low exfiltration rate even during high wind conditions. In the event of a pipe break inside the primary containment or a fuel handling accident, Reactor Building isolation will be effected and the SGTS will be initiated. Both SGTS exhaust fans will start. Af ter a preset time delay, one fan is stopped. With the Reartor Building isolated, each fan in the SGTS has the capability to hold the building at a subatmospheric pressure of 0.25 in of water when dr& wing air from the building at a flow rate of 4.000 f t'/ min. Exhaust fan outlet damper controls on each fan are provided to maintain the required flow rate. The RBICS performs the required isolation actions of the SCS following receipt of the appropriate initiation signals. Following ;
-initiation, the Reactor Building ventilation isolation dampers close The .RBICS also autr;jipily.. trips the Reactor within 3 sec. ~
Building supply and exhaust fans, anc L.ng star s- p. The normal design flow rate in the Reactor Build op @QhSGT.S..CrefuehqqQ1oor exhaust duct is 40.000 f t'/ min. Duringt, - do hp.1'l e increased to approximately 50.000 f a( J% more than 3 sec for fission products 4W4-44m' rettu!ated ce handling accident to travel from. the operating (refueling) floor ventilation exhaust radiation monitors to the isolation dampers. Thus, no direct release of fission products to the environment (bypassing the SGTS filtration processes, and_ the eleva_ted release e is possible.~eicept when direct torus } goin_tgo_vided_by th_e _ main stack) ( vent path is usef durine a severe accident. r-- The SGTS filters exhaust air from the Reactor Building anti discharges , the processed. air to the main stack. The system filters particulate m , and lodines from the air stream in order to reduce the level of l airborne contamination released to the environs via the main stack. J When the system is exhausting from the Reactor Building, the building ! is held at a minimum subatmospheric pressure of 0.25 in of water. l
/
Appendix G identifies . requirements for establishing secondary containment (Safety Action 27). following an assumed pipe break inside the , primary containment (Event 39), and following an assumed spent fuel handling accident (Event 40). Secondary containment is ) l not established following assumed pipe failures which result in the release of steam into the Reactor Building (Event 41). The following piping which is located within the Reactor Building and normally contains hot fluids at reactor pressure was considered: High Pressure Coolant Injection (HPCI) turbine steam supply line; 5.3-5 Revision 6 - July 1986
y ;;S-TSA.; Attochment i Sheet 7 of 19 b 5,4 CCf.iRCL OF CCMBUSTIE;.E CAS Cot;;C; TX10t:5 IN COf.iAIMC.i 5.4.1 Introduction This A system for control is provided as reguared by 20CTR$0.44(g). l contr i l system is provided for (LOCA) combined with ' following a postulated loss of coolant accider.:of Core Standby Cooling Systems degradat2on, but not total failure, Degradation, but not total failure of the core standby (CSCS). cooling function means that the pe rf ormance of the CSCS is postulated, for the purpose of design of the Combustible Gas Co System (CGCS), not localized clad melting and metal-waterdegree reaction 'S , that there extent could be The of!$ < to the postulated in 10CTR50.44(d)(1). performance degradation of the CSCS is not postulated to be jey i, & sufficient to cause core meltdown. fo E The Containment Atmospheric Control System (CACS) is provided to , WI ,, obviate the possibility of an energy release within the Primary,j , j -l Containment form a Hydrogen-Oxygen reaction following isatopostulated be ,, g
- 3 j LOCA combined with degraded CSCS functioning. . This atmosphere containing less than 41; *43 , m, l
accomplished by maintaining an l The> oxygen in the Drywell and Pressure Suppression Chamber (Torus). ,eh ' system willt - M;i Iti
- 1. Perform initial purging of the Primary Containment
- \
a nitrogenmakeupgasduringnormal.E{,E
- 2. Provide for a supply of , >,s operation or emergency _
)eoe for nortr.a1 and p. rge hMnes tg. the Standby EeI
- 3. Provide Gas Treatment System (SGTS. ! for D6dng,Npglgions ' S3# ,
Provide for emergency exh4u,s n S hf 6 UET 1i I
- 4. -pe==*a t 1e vT release of contaminated DryweAl . J ~r- '"Ee I &.
- 5. Provide pneumatic supply to instruments inside the drywell " e, 8 ,
O$j 5.4.2 Source of Hydgrogen Accumulation in Containment mI* i Following the postulated design basis LOCA combined with degraded the!yES S 3 CSCS functioning, hydrogen and oxygen may be evolved within reactions and from {f g O r from postulated metal water for post ! *jg primary radiolysis. containmentIn the Pilgrim Nuclear Power Station design, 0$$ l: gas control, the oxygen concentration is chosen ,g5 accident combustible as the parameter to limit. E*O! ggg 5.4.3 System Description , "E $ 3
,eg 10CTR$0.44 (a) reguires a method for control of hydrogen gas that may 4**
be generated in a BW primary containment following a poetulated LOCA by metal-water reaction of fuel cladding and coolant, radiolytic 10CER$0.44 (b) decomposition of coolant, and metallic corrosion. primary containment hydrogen system to measure requires a 5.4 *- Revision 5 - July 1985 ee e
I Saf ety b a' uation No. I Mq Rev. Nc.C
, PNFS-TSAR At t a c hment - 1 s
Sheet 8 ef 19 The valves in redundant paths are powered from independent Class IE g y distribution systems each of which is powered from en emeraency diesel generator after a loss of_offsite power ffhe control switches l for redundant valves are located in separate Class IE control panels ;; in the main control room. Conduit and permanently installed ; equipment required for purging and repressurization functions are g located in se_isdeally designed,,mjssile protected buildings, except ;. I l gM$e",_fil.1.c onn,e c tions which ,,,are ,locat,ed outside of, secondary _co_n_t_a inme nt but separa,ted._, Redundant conduit systems are separated *yk gg commensurate with identified hazards. All conduit and permanently ie g l installed equipment required for purging and repressurization functions are supported to meet seismic Category I requirements uN o l except for Na supply equipment described previously. The solenoid valves are ASME III class 2 and are qualified environmentally and seismically to the requirements of IEEE 323-1974 IEEE 382-1973. and IEEE 344-1975 for the expected conditions. The valves are rated at 120V ac and are designed to operate between 80 l and 110 percent of rated voltage. This range is compatible with expected bus voltages at Pilgrim Nuclear Power Station.= i e The control switches have been qualified to the requirements of g; IEEE 323-1974 for operation in a control room environment . The g;
- switches are mounted on Class IE panels (see Table 7.8-3) and thel y g combination has been qualified to IEEE 344-1975 for the Operating gg Basis Earthquake. The switch's electrical ratings exceed loading 3o Ae lm requirements.
I The cable and wire used for this modification have been qualified to IEEE 383-1974 for fire and ambient conditions exceeding those ee required for this installation. The 600 V No.12 AWG control cable has voltage and current capabilities well above that required. 1
$(
o, { f o i control of the solenoid valves io ( I a Osi. ther'e Asio IU automatic isolation capability. Iso: g Mv('gk (been i i EE
.c provided because:
- 1. The valves are always kt vggtS*T[tt)CTION n.
e *
>,g operation ee,
>eu The valves are required to be operated during a high drywell tao 2.
pressure condition and must be available independent of u" ' reactor water level. High drywell pressure and low-low [ *E ~N reactor water level are the normal containment isolation / signals 4 l N, makeup and ventilation valves are also provided for use under nonaceident conditions. These will automatically close gom receipt j of an accident signal. However, these valves may be used after an accident provided the required power supplies are available and a l low-low water level signal is not present. Refer to Section 5.2.3.5 l and Tables 5.2-4 and 7.3-1. i S.4-3 Revision 2 - July 1983 I
Safety Evaluation No. giaq l Rev. No.D { Attachm2nt I f
,s FNPS-FSAR Shaet 9 of 19 5.4.5 Combustible Gas Monitoring The exising Containment Combustible Gas Honitoring System (CCGMS) censists of two redundant, remotely operable, seismically qualified ,
{ hydrogen analyzers. The hydrogen analyzers can continuously monitor j Crywell hydrogen concentration and have a remote readout in the main control room. ,
]
5.4.6 Radiological Consequences of Containment Venting An evaluation of offsite doses which would be incurred as a result of containment venting to limit containment pressure has been performed in a manner consistent with Regulatory Guides 1.3,1.7, and 1.45.. The results of this analysis indicate that the doses to receptors at This analysis the LPZ would be well within the' limits of 10CFR100. 3 assumed that venting at the rate of 50 standard ft / min through the SCTS would be initiated at 80 hr after the reactor was made ical and venting would_ continue for 30 days. 4 7 - ,J,aghaftwQages) 5.4 y) ,.References 8.
- 1. GE Letter No. SSX:79-64.
- 2. July 13, 1979 Letter, W. J. Heal (GE) to S. A. Giusti (Bechtel).
- 3. BLE-459 dated September 25, 1975.
- 4. Supplement No.1 to Dresden Station Special Report No. 39 and QAD Cities Special Report No.14. ,
~
la,.S U e.; F v... ~K L. CONSTRUCTION J O 5.4-5
Safety Evaluation No. LI44 1 Rev. 0 Attachment 1 ' '
- Sheet 10 of 19 S.4.7 Direct Torus Vent Line
- 5.4.7.1 Introduction The consequences of several accident scenarios are more severe than the accidents previously considered herein. The primary containment pressure Thus, the during these accidents is estimated to exceed its design capacity.
primary containment fails, releasing reactor fission products to the secondary containment and potentially to the environment as well. The direct torus vent line (DTYL) provides an emergency primary containment vent path to prevent, or at least slow down, the buildup of potentially ma essure within t e .
$$ b _W
'EMN $N"Y M#
5.4.7.2 System Description M. The DTYL is an 8" carbon steel pipeline connecting the 20" torus main exhaust line to the underground 20" sain stack exhaust line. The 8" DTYL starts at a new branch between the existing 8" containment isolation valves in the 8" section of the 20" torus main exhaust line. The D1YL terminates in the 20" main stack exhaust line, several feet downstream of the SGTS outlet valves. The line includes an 8" air-operated, normally-closed butterfly valve which Both serves as the outboard containment isolation valve for the DTYL. electrical power and' valve operator active gas (air or nitrogen) supply are taken from " essential" or reliable sources, or are backed-up to ensur_elhet___ , the system is available during a station bla Ami-e tMt-et-Mtr%ent71n s av'"*- i !SSUED FOR o The DTYL meets ASME B&PV Code (1980 Edition ith'Vinter WAdde'ndiW5e'dtfo'n h -
!!!, Subsection NC for Nuclear Class 2 requirements up te andrinc10dirig tht . L isolation valve. The new piping downstream of the1sointibh WTiis meets ANSI 831.1 (1977 Edition through Winter 1979 Addenda) requirements.
During normal or general transient conditions, the DTYL outboard isolation In response to a severe accident, plant management valve would remain closed. could direct the control room operators to esploy the DTYL to relieveIn this case, the operat excessive pressure within the containment. follow a written procedure to perform the following basic actions: o Close, or confirm closed, the outboard isolation valve for the torus main exhaust line o Close, or e nfirm closed, the SGTS outlet valves to prevent the high containment pressures from back-pressurizing the SGTS filters o Open the two DTyL, isolation valves o Optimally, turn off the SGTS which likely came on automatically in response to a high drywell pressure signal
baIEty tvaivanon n6.! i i g Rev.-3
- Attach,9ent 1 i Sheet 11 of 19 5.4.7.3 Radiological ~ Consequences of DTVL Use The exhaust gases released by the DTVL during a severe accident would have initially been " washed" by the , suppression pool water which would reduce the
! particulate released. These exhaust gases are vented to the highest vent point (main . stack), avoiding the ground level release of radioactive material in case .of containment failure due to over-pressurization. 4 i
-. m. ___
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'o' P.ev . No . D
.t
' pgps.rsAR Attachment 1 r Sheet 12 of 19 e
Traversing 2ncore probe i RHR reactor shutdown cooling supply RHR reactor head spray .
! Reactor water sample lines Reactor water cleanup o g.
a so jif Drywell purge inlet and makeup gas *, ** e Drywell main exhaust. je suppression chamber exhaust valve bypass *, ** 4E ) Ta E f M *; Suppression chamber purge inlet and makeup gas *, ** cE
- j e
suppression chamber main exhaust gs . I g-I- Drywell exhaust valvis:Uypsiss*,----- ' RHR-LPCI supply
- l-
- p~* t ? ;'~ "~ c ..<*- <
RHR to Radweste (
' ' ' VVI'h)?ry6Y;1,,1J
- \ * +- 'a 7.,- h}' }J 8 i e
Containment atmosphere sampling lines j ,? E ,.o 8 3 y* containment makeup and ventilation valves are also
*g;
- g%vi provided for use following an accident condition. Refer to ,E$'
remote manual operated. These are - g~*, Section 5.4.3. % $oI. '
** The reactor water low level isolation signal can be by ;,y > b q {!
passed. These valves may be opened anytime provided j the low-low water level signal is not present. , y%i j
*R 1 and lower of the reactor vessel low water level i *E g ,, I The second isolation settings was selected low enough to allow the,"ot*,"
- from the reactor for a predetermined time,i ,E removal of heat g following the scram, and high enough to complete isolation. ,5,,, y g -
in time for the operation of CSCS in the eventThis of asecond large 2 H
- break in the nuclear system process barrier.
low water level setting is low enough that partial losses of ' , g ) l feedwater supply would not unnecessarily initiate full / ssolation of the reactor, thereby disrupting normal shutdown - -. / or recovery procedures. Isolation of the following pipelines is init .ated when the reactor vessel water level d falls into this second setting. I All P.,ur main steam lines . 1 HPCI, RCIC, and main steam line drains
~
J 7.3-22 Revision 2 - July 1983 l (
p .,o . s v . . . . .
...m PhT5i-FS AR At{$c ent 1
- Sheet 13 of 19 .
L^_ I-included here to make the listing of isolation functions compiete. )
)
- 6. Primary Containment (dryvell) High Pressure High pressure in ti.e drywell could indicate a breach of the nuclear' system process barrier inside the .drywell. The ;
automatic closure of. various Class 5 valves prevents the j l release of significant amounts of radioactive material .from J l the primary containment. Upon detection of a high drywell j l pressure, the following pipelines are isolated, see Table j l- 7.3-1, signal F. ] { Traversing incore probe RHRS shutdown cooling supply , RHR$ reactor head spray Reactor water sample lines Drywell equipment drain sump discharge Drywell floor drain sump scharge Traveling incore probe tubes ___ Drywell purge inlet and makeu gas *, % d { jt F0 ~7' ; i ,. 4 -s. Drywell main exhaust , {,,.-,....,.,.,
, } , q; i; ~~
suppression chamber exh'aust v'aTve bypas"sT ,7*~~~~ ~ ^ ~ Suppression chamber purge inlet and makeup gas *, ** I r gp.p_ression chamber ma_in exhaust recttorusven't***J I Dryw' ell' exhaud~ Fvalve bypass *, ** RHR to Radweste s containment atmosphere sampling lines The primary containment high pressure isolation setting 4s , l selected to be as lov as possible without inducing spurious I isolation trips. See Table 7.3-1, Signal F.
- Containment makeup and ventilation valves are also ' )
provided for' use following an accident condition. ] These are reacte manual operated. Refer to Section I 5.4.3. ; i l
~ , \
l 7.3-15 Revision 2 - July 1983 l
Rev. ho.D PNPS-FSAR Attachment'l Sheet 14 of 19
' 'a
** The reactor water lov level isolation signal can be by-passed. These valves may be opened anytime provide d .'
l 8 the low-low water level signal as not present. a i 7. RCIC System Equipment Space High Temperature High ternpe rature in the vicinity of the RCIC System equipment could indicate a break in the RCIC steam line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radicective material from the nuclear system process barrier. Whern high temperature occurs near the RCIC System equipr.ent, the RCIC turbine steam line is isolated. The high temperature isolation setting is a selected far enough above anticipated normal RCIC system i operational levels to avoid spurious operation but low '- I enough to provide timely detection of an RCIC turbine steam line break. See Table 7.3-1. Signal K.
- 8. RCIC Turbine High steam Flow RCIC turbine high steam flow could indicate a break in the RCIC turbine steam line. The automatic closure of certain
'g Direct torus vent valves are provided for use followiThe 'C primary c an accident condition.
high pressure isolation signal can be bypassed. These are remote manual operated valves. -
--_--=.=.-- , ,;_
- =---f f } ; -{I Q CONST!S.!CTiON 2=. .~.=:; -* * * - --
j 4 t j l I i
,~., j O
'I l
* ' 7.3-15a Revision 2 - July 1983 ;
i
' 0 f* /
~~ key, ho , v
/ Attachment 1 34':"
'* <[i 'PHFS-TSAR Sheet 15 cf 19
, , .s
' ,1' The .legac arrangement) used. for this function,.shown on the. usual s logic Tagure 7.3-7.. 1s ': an exception to .[,;
requireraent, because high steam flow is the s'econd method of'(' g detecting an HFCI turbine stee.m.line break.
, g.(
3-- 3
" j 12.'HPCITurbineSteamLine}ovPrepriure 3 +
s
, 3
'HPCI turbine steam line low pressure. f5 used to ,
.Q, e . automatically close the two isolation valves in the HPCI
. g~ O l turbine steam line, so that steam and radioactive gases will
'% not escape frop .the HPCI turbine sheft : seals into the
~
~
Reactor Building af ter steam pressure has decreased to such 1 a low value that the turbine cannot be operated. -The
-isolation c setpoint is chosen at a pessure below that where the HPCI turbine can operate efficiently. See Table 7.3.1, *1 Signal L.
- 13. Reactor Water Cleavaup Ept.ei 3 pac'e ,High Testperature High temperature in the vicinity # of the reactoryoter cleanup (RWCU) equipment and piping could indicate ia. ' break-j- in a RWCU .11M. Th+ automatic closure of certain Class &
'L val,ves prevents the ticensive losh*of reaccer coolant?amd thO release of significant amounts of radioactive material
( 'J k1'from t*4e nuclear system process barrier. When <jhigh ( f L . temperature . occurs near the RWCU equipment, th'e RWCU 16ystes
<< - t,t 'Is isolated. The high temperature isolatied ' setting is selected far enough above anticipated? 'nerphl' system
- om-itoistio@3H= Tov - ._ i ,
i .! operational enough tolevels provide to avoid timelys 'e: 6 tion gof g $ipe>Dreak',(sel h Tgble7.3-1,signalJ.
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c l.J t/ W b * ' ' y i s ,, a e,3n c'rba !!*"{ { N 5 ]
- 14. React $r' Water Cleanup System Hid. __MoWJ ik ....-{h, y b,.
- 3
, m -
RWCU high flow could indicate a break in a RWCU li'se, The automatic closure of certain Class R valves pr' events the excessive loss _ of reactor coolant, and. the release of significant amounts of radioactive materials' from the nuclear system drocess barrier. Upon detection of RWCU high flow, the RWCU line is isolated. The high flew trip setting was selected high enough to avoid spurious isolation, yet 0 (" lov enough to provide timely detection of line break. see " Table 7.3-1, Signal J. y {j s, w 3
- 15. High Reactor Vessel tressure j ;u J t High reactor vests.1 piessure is used'to automatically close
' 'the two isoletica Oves in the RHR pmps' shutdown 1 cooling 4%dh suction piping z.o that the RHk low p:4sture piping wilt not be threatened by, swprpressurizatson. Tne isolation setpoint 4,
is chosen at a pressure below wnerd thn RHR piping could be ' f t overpressurized. See Table 7.3-1, fignal U. a j[. - g 4 s ! ggf mes %eto EmC4*.M j
.N. See Insert C on page 16, ,/ 7 o
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" ,,,,3-17 7 P y ,
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- Safety Evaluation do. 21 4 Rev. 0
"' Attachment 1 I esd7 O - ... ____.-
Sheet 16' Of 19
. _ _ , _ _ As '
Radiation in Reactor Building Refueling Floor Vent Ex t, Duct N2 k High radia in refue' ling floor vent exhaust duct uld indicate a igh radiation in-
,p gross release o ssion products from the fuel ;'
refueling floor exh duct initiates isol on of the followirl pipe s
'[- lines. (See Table 7.3 , ignal V.)
o Suppression chamber main e st . i Direct torus vent o
'The high radiation setting is selected enough above background rad on levels to avoid spurious iso on, yet low enough -
ect a gross release of fission products m the fuel, ( to prompt 1 f Ir g mseu c -; 4 iSSUilD
... .vs cp, n.
g ,7. >..L. 3amg 95 W 3.., ,s. .;e g
~~==*~~~'~ '
g, K I Suppression Pool High Radiation High radiation in torus could indicate a gross release of fission products from the fuel. H1 h radiation in torus initiates isolation of the direct torus vent pipe ine to prevent release of significant l amounts of radioactive materials. The high radiation trip setting is selected high enough above background radiation levels to avoid , I spurious isolation, yet low enough to promptly detect a gross release ; of fission products from the fue , to the main stack, within the I r , allowable limit. See Table 7.3-1. Signal Z. I i L
)
i l
PNTS TSAR - {)
' Sheet 17 of 19
- 13. 'Hagh- ter;erature .in
~
the spaces cccupsed by the RHRS (E[ - (shutdown cooling) .and piping outsade the pramary
'-> ;y - E ' conta ireent is sensed by temperature switches that activate-c "o L - alarms only. Andscatang possible pipe breaks.
Autor:.at ic I [ U '6t
- s. -
A typical arrangement is shown on ' Figure 7.3-8. isolation on hagh temperature is not required since the
$ 5N C' #A o en i reactor vessel low water level isolation function as adequate in preventing the release of sign 2ficant amounts of radsoa:tive material in the event that either of these two
.hoMU {3I5 1 systems suffers a breach.
' "$ $ 5 h
- 14. High temperature in the vicinity of the' ItWCU system is Egge 5 sensed bY four sets of two binetallic temperature switcher.
$$,L A set of two temperature switches is installed in each of 3 u M .* the four areas to be monitored: each set is a one out of two
"* trip system and capable of initiating isolation.
E kj '=', ,,
,hph$ 15. High flow in the Reactor
- dater Cleanup Syster supply line is sensed by two differential pressure switches which monitor t]*
- gc: $ E the pressure difference across an elbow installed in the ac w E o- Reactor Mater Cleanup System supply lina. The arrangement of the differential pressure switches is similst to that .
shown on figure 7.3-12. .The . tripping of either switch g Jnitiates isolation of the RWCU system. i f Channel a$d logic relays are high reliability relays equal to type HTA relays made by the General Electric Company. The relays are. , [ selected so that the continuous load will not ex:eed 50 percent of i ( the continuous duty rating. I
~
F gd CI J": p Nq 3
- 8' \
7.3.4.8.1 High Temperature Sensors e f
'T i spatial independen:e.. Th] Q CT'D $ i'/"T[r*j.11J/i'st'ahee'-j e
*' \ The location, ~
EE*' tripping of the high temperature sens'o' rs in-'EheW2N"ife#ainlinei~HPfi" l
*L i turbane steamline and RCIC turbine steamline are detailed in this ;
section. eQ ~f i
.. w S C, lt Table 7.3-3 lists the areas outside the primary containment where !
main steam. HPCI, and RCIC steam lines are routed. This table also
.o } Eto lists the leak detection sensors, summarises the physical separation II
- of sensors, and specifies the set points at which asolation of the I M*' respecting steam line would be initiated.
c,g E f Potential leak sourcer and rates which would initsatt isolation are I i as follows: Ooe f MainSteamTunnel
%1E This area contains feeduater, cleanup, and usin steam piping.
q ,g
** Isolation of the main stearr lines will be initiated at ventilation I I' o E exhaust temperatures from the main steam tunnel of 160*T to 170'T.
which would result from steam leaks equivalent to 10 gal / min and j
/( o)$dh g g. , greater. The feedwater and cleanup systems normally operate at ;
.S
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s las errumu pensum . saaee eestalammet spyw valies. j T las remeter pmaann peresselse to spee een opew ese M unlese. U 35$..reestar a we emevessel pressure.. n n, = 1. a . glane RS stadeven _eemuas - y 8 A l ~7Il "Q "w ~ q# Ui M PIM ^7 4= L severesum et auslet er caensup erstem nearepaarstTaXshn 'j
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3C1C or RFC1 stone supply valve (as appaleable) met fin 1% I ~ mt.,e. ei .. ee .,,en .e. o .e4 g .clanee.- e. ,ee, ,, ,re., sue tie,e av - -
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