Text

Packages Consisting of marked-up Rev 0 to Safety Evaluation 2144 & Slides of Facility Safety Enhancement Mods
	ML20238E692
Person / Time
Site:	Pilgrim
Issue date:	12/30/1987
From:	; BOSTON EDISON CO.
To:	;
Shared Package
ML20238E673	List: ML20238E671; ML20238E692;
References
	FOIA-87-643 NUDOCS 8801050292
	Download: ML20238E692 (64)
	v • d • e

N w' Proposed Chang 2 Safety Evaluation No.:

2. 14 4 SAFETY EVALUATION i

PILCRIN NUCLEAR PCMER STATION Rev. No.

PDC POI System Calc.

Initiator:

Dest:

Group:

No.:

Name:

No.:

Date:kl5fd9

c. v. Mileris NED FSMC 86-51 Direct Torus Vent Provide a direct torus Description of Pr..mosed change test or exp{ertaent:

vent to the main stack b7VS) 1 SAFETY EVALUATION (IELUSIONS:

The proposed change. test or expertaent:

^

1.

6) Does Not ( ) Does increase the probability of occwrence or consequences of an accident or malfunction of equipment leportant to safety previously evaluated in the FSAR.

2.

() Does Not ( ) Doe's create the possibility for accident or malfunction of a differest type than'any evaluated previously la the FSAR.

3.

F) Does Not ( ) Does reduce the margin of safety as defined in the basis i

for any technical specification.

BASIS FOR 1AFETy tvaittaTION rTWMMf0NS:

SIE ATTACHED SHEETS Change Change h Recommended

( ) Not Recommended SE Performed by b'

Date M M

Exhibit 3.07 A

.p Sheet 1 op,-M, 1550c9 L.. \\n*

C ~* he u "'* W...,i w

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3.07-13 Rev. 4 8801050292 871230 PDR FOIA SORGIB7-643 PDR L

bafety r. valuation N3.: 244 sArtTY tvAusn0N PILCRf u NUCLEAR ptMR STATION f#E F7" l2. o # //

Rev. No.O A.

APPROVAL This proposed change involve a change in the Tec,hnical Specifications This proposed change, test or experiment does ( ) does not 9d involve an unreviewed safety question as defined in 10CFR, Part l

50.59(a)(2).

/ This proposed change involves a change to the FSAR per 10CFR 50.71(a) and is reportable ender 10CFR50.59(he). scope of this safety b

Use of the ven-is beyond t Commente:. evaluation and will be covered in the emergency operating procedures.

A separate safety evaluation will be written to cover use of the vent.

The safety evaluation basis and,c lusion is:

Q Approved Not Apppoved J

.rsei A M -

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REVIEW $PPROVAL U '\\'

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ORC REVIDI sG DTvs a5e.,

(, h This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with the Directorate of Liceastag. NRC prior to implementaties.

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This roposed change does set involve an unreviewed safety h/f'/

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Date QRC Meeting Number 89-f/

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SATE!Y E7AI.UATION NC. W REV. NC. 0 SHER 3 0F,Lf, A. DISCRIPTIC:: OF FROPOSID CHANGE This modif:.ca icn provides a direct von: path f rom the torus to the main stack hypassing the Standby Gas Treatment Syatem (SGTS) equipment c: the terus purge exhaust line.

The bypass is an 8" 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 SGTS).

The downstream and of the bypass is connected to the 20" main stack line downstream of SGTS valves AON-108 and A0N-112.

An 8"

butterfly valve (A0-5025),

which can be reactely operated from the main control room, is added downstream of 8 ' valve A0-50425.

This valve acts as the primary containment outboard isolation valve for the direct torus vent line.

The new pipe is ASME III class 2 up to and inclusive L."P!*_^E-P25... U" Sinib/ mHMJ,pfilis.i; g*Tg7 aa p.Ma nna s.'.".:.Wu.i y:s.n.

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AC solenoid The proposed modification replaces the existing (powered from for A0-S C425 with a

DC solenoid valve valve essential 125 voit JC) to ensure operability during a station

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blackout.

The new isolatie valv 0-5C25, is also provided with a DC solenos.d powered from 125 volt DC source. M Both of these valves fail closed.

One inch nitrogen lines are added to provide ba.ckup nitrogen to valves A0-50423 and A0-5025.

The present logic of A0-50423 is being modified to override containment isolation signals by keylock remote manual action.

New valve A0-5C25 closes on containment isolation signals but ts J

provided with the same isolation override contnil logic as AO-i 50425.

When 5025 and 50425 are in the containment isolation bypass clode a separate logic has haan..added to isolate both i

in the Torus vapot high radiaHEn~~je,vpl; d C adcoitplikhe valves if there is a

overriis is.Jals space.

This high radiation x yleek remote manual action.y g g g g. g p A

20" pipe will replace the

" " E,.- ;

a a-v.or cLuct between SGTS valves A0N-108, A0N-112 and the existing 20" pipe to the main stack.

Tbs existing 20" diameter duct downstream of AO-5042A is shortened to allew fitup of the new vent line branch connection.

A rupture d.isk will be included in the 8" piping downstream of valve A0-5025.

The rupture d:.sk will provide a second leakage barrier.

The npture disk is designed to open at a pressure below direct venti =g

pressure, but will remain intact during normal operating c::cditions.

This Safety Evaluation demonstrates that plant safety during normal design basis operation is not degraded by the installation of the direct tons vent, vent operation is not ad r by thaC5 Shts an <rnya se YM %. ve. Ins w:n n.+De.m %

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SAFETY EVALUATION NO. 2W REV. NO. O SEEET + OF, L safety evaluation but will be addressed by another safsty Eval.:a ion in conjuncf. ion with the Energency Operating Proced:res.

A*.though this safety evaluation does not address the vent operat* on, a description of the venting caperation and logic of the installed modification is provided below for information.

ne purpose of the vent is to relieve excess containment pressure and prevent catastrophic containment failure as directed by the Emergency operating procedures.

Use of the vent will be under management administrative control and would require that the keylock switches for operation of the A0-50423 and A0-5025 valves be placed in the Emergency Open Position to override the containment isolation signal which would be present if the contai:mont was at high pressure.

Prior to opening the vent valves the 8BGT system would have to be shutdown and valves AON-108 and ACN-112(the outlet of SBGT)placed in'a closedAO- 0522 and M osit t.on.

If there is high radiation,1n the torus vapor spaceThatisolation algnal can be overridden by a g

)

se4 ek. A0-5025 will raisolate.- manual keylock actuated switche I

S. P~RPOSE OF THE CHANGE

' CONSTRUCTION mis modification will provide t4-ability te add *=ea of the severe accident concerns by direct venting of the torus to prevert-primary containment over-pressurization during an extended station blackout.

If use can be justified this modi.fication opens a direct exhaust path from the torus vapor space to main stack.

C. SYSTEMS, SUB5YSTEMS, COMPONENTS AFFECTED Sie modification affacts the containment Atmospheric Cont.r:1 System in the following manner:

he torus purge exhaust line inboard isolation valve A0-5042B and the associated 8" pipe are the components of the CEC 8 affected by this proposed modification.

With incorporation of the subject modification, the CACS will depend on both essential AC (for valve A0-5042A) and essential DC (for A0-50425) toperformitsgn on. p The new 8" torus vent line will be connected to ex sting 6" "ACS piping between valves A0-5042B and A0-5042A.

"his modification affects the Standby Gas Treatment Systaa in the following manner:

-l I

SAFETY EVAtUATION NO.dlN REV. No. 0 SEEET f 0FA j

SGTS fan octle: valves (AON-108 &

AON-112), ductwork f rom these valves to -he 20" line leading to the main etack, and 1

the 20" line leading to main stack are the components of l

this system affected by the proposed modification.

Valve AON-108 is normally closed, fail-open.

Valve AON-112 will be revised to be normally closed, fail-closed, and these valves will be provided with essential Dc power and 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 isoletion valve A0-50425 logic.

The addition of containment outboard isolation valve (AO-5025) and associated controls i

will also affect the PCIS.

j D. $AFETY FUNCTION OF AFFECTED SYSTEM / COMPONENTS Containment Atmospheric Control' System This system has the safety function of obviating the possibility of an energy Iglease..Within the primary g

following a

%egr(d,gk.rgaction(pre,;stprdbyLCoolI.ng containment free a HydrogerFoxy ~en postulated LOCA combined with

'*" " F

! CONSTRUCTION 8tandby Gas Treatment System This system filters exhaust air from the reactor building and discharges the processed air to the main stack.

The system filters particulate and iodines from the air stream 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 protection against the onset and consequences of accidents involving the gross release of radioactive materials from the primary containment by initiating automatic isolation of appropriate pipelines which penetrats the primary containment whenever monitored variables exceed pre-selected operational limits.

Pc m es Cu% d hsh*m 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.

\\

EATETY EVALUATION NO.2l N REV. NO., O sHEETgOF //

E. EFFEC"$ ON SAFETY FUNCTION Ccetainment Atmospheric Control System and Standby Gas Treatment system, ano Primary Con ainment Isolation system 3

j The modification changes the solenoid A0-5042B control from AC to DC enabling it to ope = (from its normally closed I

i Position) when required even during extended station blackout.

Ductwork at the outlet pipe and the new vent line is connectedof the 5337 system is replaced with

~

to the 20" line at the outlet of the 83GT system.

Addition of a new 4" vant line with containment isolation valve A0-5025 off the torus purge and vent line results in a flow path that could vent the contain'nent directly to the stack bypassing the sBGT system during normal plant operation.

New logic is being added that allows override of the containment isolation signal on existing valve AO-50423 and provdes the same logic for valve A0-5025.

This could allow venting of the containment directly to the stack bypassing BBGT or subjecti'ii~~sbe BBO system to high G y {'T containment pressure.

3 g.

4 F. ANALYSIS OF EFFECTS ON SAFETY F1 M GdsiC T 1 tIg.m "I M I t -- oJINLl%,IIVIM An analysis of the effects on suesy

~~

~'S, SGTS, and PCIS systems vent is described as follows:for the installation of the direct torus The change from. AC to DC control on A0-5042B does not adversely affeet its ability to open AO-50423 when he containment is being purged, or iseWW mA *.cc6cLAad c]M. y The modifications to the ductwork and 20" line leading to the main stack do the safety related systems.not affect the safety function of any of __

During normal plant operation, the CACs and the use the torus 20" purge and vent line to perform their SGTs do not safety functions.

The containment isolation valves are in their normally closed fune. ion of the PCIS.

position, satisfying the safety p

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9 SAFETY EVALUATION No.1144 REV. NO. O sHErT 7 or,~24, (nw-encyv% %Wh& W During plant startup and shutdown"when the purge and vent line is in use, the logic does not allow A0-4b 6 to open p unless 5&42 g is closed.

In addition a rupture disc W downstream of valve A0-5025 will provide a second positive

means of preventing Isakage and prevent direct release up the stack in theevent of a single failure of A0-5025 during containment purge and vent at plant startup.or shutdown.

During containment high pressure conditions, the CAcs does not use the torus main exhaust line to communicate to the SGT5 for performing its safety function.

The existing CACS logic cannot override the containment isolation signal and I

open valves A0-5042A, & A single fmilure of A0-50423 logic M would still protect the 85GT 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, sGTs, and PCIs Systems, or any other safety-related systems.

This safety evaluation dose-not provide justification for use of the DTV8.

The installation of this modification does require a

Technical Specification Change.

The installation of this modification does not involve an unreviewed safety question.

ISSUiO rh 1

CONSTRUCTION l

l l

SHEETff OF //

Safety Evaluation No.:

2.i44 l

SAFETY EVAltlATION WORK SHEET Rev. No.

O A.

Systes Structure Component Failure and Consequence Analyses.

Systes Structure r e ent Failure Modes Effects of Failure Comments SEE ATTACHED SHEET 2.

3.

General Reference Ikterial Review FSAR CALCULATIONS REGULATORY SECTION PNPS TroaccAL SPECS. DESIGN SPECS Mt0CEDURES EllDES STANDARDS GDES SIE ATTACHED SHEET

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For the proposed hardware chanfle identify the failure modes that' are likely for the components cons'. stent with FSAR asseptions. For each failure unde, 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 systee) or plant safety functions FSAR Chapter 14 and Appendix G).

Date N J Prepared by NOTE:

It is a requirement to include this work sheet with the Safety Evaluation.

Exhibit 3.07-C 3.07-18 Rev. 4 l

SAFETY EVALUATION No.El#4 REV. FO. O SHEET 4 OF /f I

SATETY EVALUATION WORK SHEET system / structure / component: Isolation valve'A0-5025.

'^ g Failure Mode s Failure to close/F *I e pu W G t-P#Y"lr d

I::ects of Failure: Bypassing of' Standby Gas Treatment System; Loss of Containment Isolation Ocments : lE Valve position indication is provided in the main control room.

Rupture disk provides protection.

Failure Mode: Excessive Leakage.

It:ects of Failure: Release of Radioactive material above allowable.

Comments: Periodic LLRT perforr.ed to ensure leak tightnese.

Rupture disk provi' des protection.

(Rupture disc is tested periodically)

System / structure / component: Direct Torus Vent system piping.

Ta11ure Mode: Structural railure.

Errects.of Yailure: Less of containment integrity and 8GT8 inoperable.

for design basis temperature a::d Ccaments: Qualify piping.

pressure to ASME III and 331.1.

System /5tructure / Component : Primary Containment Isolation System.

Failure Mode: Logic. failure.

Ettacts of Failure: No effect.

a*J ry/wa dl4 k, M Comments : Redundant trains of logicy

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Safety Evaluation No. 2 # #4/

Rev. No. O Sheetjpofy SAFETY EVALUATION WORKSHEET j

FSAR CALCULATIDES/PNPS REGULATORY DESIGN GUIDES / STANDARDS SECTION TECHNICAL SPECS.

SPECS / PROCEDURES CODES 5.2 3.7/4.7

. Spec. SM-34 10CFR50 I

Spec. M300 W REG CR-624 5.3 Spec. M600 W REG BMI-139, Vo1 1 Spec. M301 5.4 Spec. M600 W REG BMI-139, Vol 1 Spec. M-611 W REG 0700 7.3 Specs. 17322-M-SAM-16 ASME B&PV Code Sections 10.11 17322-M-SAM-12 III & XI Table 7.3-1 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 Calculation Number Teledyne Cale./Later 17322-M-640-1 N-3,r d I' d &

571-28-17322 and 571-29-17322 UC HUI'~'

CONSTRUCTION 10394-115-C3 17322-640-Cl20.0 17322-630-C200.0 l

17322-640-C100.3

'17322-640-C110.0 17322-640-C100.5

Safety Evaluation No. 214 4 Rev. No. O Sheet ll of it PILGRIM STATION l

FSAR REVIDI SHEET l

References:

PDC 86-51 Sugport a change to provide a Direct Torus Vent Line.

List FSAR test, diagrams, and indices affected by this change and corresponding i

~

FSMt revision.

Affected FSAR Revision to affected FSAR Section is shown on:

Section Prelisinarv Eigi Section 5.3.3

~

Section 5.3.4

[N-Sect a 5.4.1 f

N*.bd'_i Section 5.4.7 7.y d \\

l Section 5.4.8

[ C{ J\\ tb l R,. U C T l O N,d.

Section 7.3.4 Table 5.2-4 Table 5.2-5 Table 7.3-1 l

Figure 5.2-16, 5.4-1 M 227, Sht.1 Figure 10.11-1 M 220, Sht.1

(

In adfition, the PNPS Technical Specification. Table 3.7.1, Notes for Table 3.7.1, Bases 3.7 D/4.7.D.

PRElinIRARY FSAR REVISION (To be completed at time of Safety Evaluation prepration.)

l Prepared by: M A /Date: 64; d 7 Reviewed by:

ate:

'Mf/

Appewed by:

/Date:

/f/87 FIItAt FSAR REVISION (Prepared following operational turnover of related systans structures of components for use at PNPS.)

(1)

Prep red by:

/Date:

Reviewed by:

/Date:

(1) Attach completed FSAR Change Request Form (Refer to NOP).

Exhibit 3.07-A Rev. 3 (Sheet 2of3)

Safety Evaluation No. 2.144 Rev. No. O Sheet 1 of Je/

ATTADMENT 1 RECOMMENDED FSAR CHANGES The pages of the following sections, tables, & figures of the FSAR that need to be updated due to this modification (PDC 86 51) have been marked with suggested updates and included in this attachment for your review.

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l.

l FSAR Sections: 5.2, 5.3, 5.4, 7.3, 10.11 FSAR Tables:

5.2-4, 5.2-5, 7.3-1 The following drawings will be revised as part of the Plant Design Change package (PDC 86-51) but are not included herein.

Dwg. 1. D.

FSAR FIGURE TITLE i

I M 227, Sht. I 10.11-1 P&ID Containeint Atmospheric,

I Control Systes l

M 220 Sht. 1 5.2-16, 5.4-1 P&lD Compressed Air Systos 1

l M 294 5.2-17, 7.18-2 Heating Ventilation and l

Air-Conditioning Standby Gas i

Treatment System Control Diagram i

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G

r a :-: n Rav. No. O l9 Attachmsnt 1, Sheet 5 of s'olenoids or u'pon loss of instrument air to the air operators on the C'.

dampers.

(.

The demister is designed to remove entrained water droplets and mist l

~

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 eshaust fan prevents the heating toil from operating when the fan 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 filters are iodide-impregnated activated carbon filters capable of removing in excess of 99 percent of the lodine in the air stream with 10 percent of the lodine in the form of methyl todide (CH I) under g

j entering conditions of 80 percent relative humidity.

i The accident evaluations ustmg the standard NRC approach are l

cescribed in Section 14.9.

In these antlyses the SGTS charcoal filters were credited with renoval of 95 percent of the influent iodine.

l The system will start automatically upon a high radiation signal frem j

the operation (refueling) floor ventilation exhaust duct monitor, or l

upon receipt of high drywell pressure or low reactor water level l

J signals.

The system can also t>e manually started from the control room.

Upon receipt of any cf the initiation signals, both fans j

start, all SGTS isolation dampers open and each fan draws air from the isolated Reactor Building at a flow rate of approximately 4,000 ft'/ min.

After a preset time delay, one fan is stopped.

Cross-connections between the filter trains are provided to maintain J

A W 4 10w on the charcoal the required decay heat removal cool ny filters in the inactive treatment trLin.

t'im yistparge} to ftie SGt Wn6are N

a the main stack through a 20 in undergro nd powered from the emergency service por lo @ g g g *TT p g g j

i distribution system.

L.. _.

~-

Drywell and torus purge exhaust can also be directed to the SGTS for j

processing before release up the main stack.

See Section S.2.

The High Pressure Coolant Injection System (HPCIS) gland seal steam condenser exhauster discharge is also routed to the SGTS during accident conditions.

The Reactor Buildina Heatina and Ventilating _

System it d i s e n u e d i n' S e c ti on 9 0. 9'.i During a severe accident, the

( torus can be directly vented to the main stack bypassing the SCTS.

5.3.3.5 IEIn' Stick see section 5 4.z.

I The location of the main stack is shown on Figure 1.6-1.

The top of the stack is at elevation 400 f t msl.

The structural design of the 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 secondary containment act together to provide the principal mechanisms fc,r the mitigation of the consequences of an accident in 5.3-4

. 19 Actechnent 1,' Shsst 6 of the drywe'll.

If the leakage rate of the building is low, and the leakage air is filtered and discharged to the elevated release point (utilizing the SGTS and the main stack) the offsite ractation doses that result from postulated accidents are reduced significantly.

The Reactor Building is a Class I structure (with the exception of the secondary containment access locks which are -Class II structures) designed in accordance with all applicabic codes.

Design of the Reactor Building for a maximum inleakage rate of 4.0C0 f t'/ min at a building subatmospheric pressure of 0.25 in of water a t neutral wind conditions, 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.

J With the Reactor 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 drawing air from the building at a flow rate of 4,000 f t * / min.

Exhaust fan outlet damper controls cm 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 vithin 3 sec.

The.RBICS also autd=lly.. trips the,, Reactor stargs,p.

SGTi.

The normal Building supply and exhaust fans, anc design flow rate in the Reactor Build l,ng opte@gh pyfueftq)J oor exhaust duct is 40,000 ft*/ min.

During c

. if t r

increased to approximately 50,000 ft, ak

%)ie,

~ t W t'd more than 3 set for fission products 9a*"

"a 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 environmer.t (bypassing the SGTS filtration processes, and _ the elevatec relene

_poin_t,p, rov,1,ded_,by _the main stack) is possitiefexcepc when direct corus. }

t

( vent path is used durine a severe accident, r--

The SGT5 filters exhaust air from the Reactor Building and discharges the processed, air to the main stack.

The system filters particulate and lodines from the. air stream in order to reduce the level of airborne contamination released to the environs via the main stack.

i When the system is exhausting from the Reactor Building. the building is held at a minimum subatmospheric pressure of 0.25 in of water.

Appendix G

identifies requirements for establishing secondary containment (Safety Action 27), following an assumed pipe break l

Inside the. primary containment (Event 39), and following an assumed spent fuel handling accident (Event 40).

Secondary containment is not established following assumed pipe failures which result in the release of steam into the Reactor Butiding (Event 41).

The following piping which is located within the Reactor Buildi-ng' 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

FNFS-TSAR Shset 7 of 19 CorCROL OT COMBUSTIBLE GAS C0!

CKA!IOic w CCNTAImen 5.4 W

5.4.1 Introduction This A system for control is provided as required by 10CFR50.44(g). c system is provided for (LOCA) combined with following a postulated loss of coc2 ant accider.tof Core Standby Cooling Systems degradation, but not total failure, (CSCS).

Degradation, but not total failure of the core standby of the CSCS is cooling function means that the performance l

postulated, for the purpose of design of the Combustible Gas Contro not to meet the acceptence criteria in 10CTR50.4 System (CGCS),

to the extent postulated in 10CTR50.44(d)(1).

The degree of!!

that there could be performance degradation of the CSCS is not postulated to bei.,*

f*

sufficient to cause core meltdown.

E The Containment Atmospheric Control System (CACS) is provided to,",

4 1

release within the Piimary,g,

obviate the possibility of an energy reaction following a postulated,-*,5 i

Containment form a Hydrogen-Orygen LOCA combined. with degraded CSCS functioning.

This is to be,g than 4t. h, m,

(

3 an at.1csphere containing less accomplished by maintainingthe Drywell and Pressu:re Suppression Chamber (T Thei oxygen in

,eg!

system will:

{

Perform initial purging of the Primary Containment i

u 1.

~

nitrogenmakeupgasduringnormal.(([

2.

Provide for a supply of D

>,e operation or amergency geee kNas tg. the Standby 3IN 3.

Provide for normal and p rge

.O Gas Treatment System (SG S' !for

kEtt%nglt:pnJ!gions ' 3f[ \\

e n

rU N S T h; h h @'3v*i3@

r

.:

Y I

Provide for emergency exhqs 4.

t ie release of contaminated Dryven.L'-*- -- 9 m + a

Ie Provide pneumatic supply to instruments inside the drywell I Isj s

i 5.

!I Source of Hydgrogen Accumulation in Containment

a. ' "

5.4.2 Following the postulated desig= basis LCCA combined with degrade CSCS functioning, hydrogen and crygen may be evolved within the U

f rom postulated metal water reactions and from fiO

!'I~~ 1 g for post primary containmentIn the Pilgrim Nuclear Power Station design,

$*t l radiolysis.

gas control, the oxygen concentration is chosen g, y "U accident combustible as the parameter to limit.

I*O

..g g 5.4.3

System Description

- -: e.,

L E>8 requires a method for control of hydrogen gas that may

,e*

<w be generated in a E'aA primary containment following a postulated LOCA 10CTR50.44 (a) by metal-water reaction of fuel cladding and coolant, radiolytic decomposition of coolant, and metallic corrosion.

10CTR50.44 (b) requires a

system to measuce primary containment hydroge.

5. 4 '.

Kevision 5 - July 1985

.p.

se

Safety Evaluation No. 2. Mtl R2v. NO O

~

PNPS-FSAR Sheet 8 of 10 The valves in redundant paths are powered from independent class IE distribution systems each of which is powered from an eneroency y

diesel generator after a loss of offsite powerdhe control switches for redundant valves are located in separate Class IE control panels in the main control room.

conduit and permanently installed t

equipment required for purging and repressurization functions are g

located in seismically designed, missile protected buildings, except Eh'e 5ii.l.. connectives w'hicE,,,are locat,ed outside of. secondary

b

_ containment but _ separated. Redundant conduit systems are separated 5&k commensurate with identified hazards. All conduit and permanently ie g I

installed equipment required for purging and repressurization ug o

functions are supported to meet seismic Category I requirements f

supply equipment described previously.

except for N2 The solenoid valves are ASME III Class 2 and are qualified environmentally and seismically to the requirements of IEEE 323-1974, IEEE 362-1973, and IIEE 344-1975 for the expected conditions. The valves are rated at 120V ac and are designed to operate between 80 and 110 percent of rated voltage.

This range is compatible with expected bus voltages at Pilgrim Nuclear Power Station.=

l e The control switches have been qualified to the requirements of

=q o

IEEE 323-1974 for operation in a control room envirocraent.

The

p. O l

Class IE panels (see Table 7.8-3) and the l

,g switches are mounted on combination has been qualified to IEEE 344-1975 for the operating gg Basis Earthquake.

The switch's electrical ratings ex=ced leading g*

A g l

re quireinents.

m EU h

cable and wire used for this modification have been qualified to

i. e The IEEE 383-1974 for fire and ambient conditions exceeding those g,*f, required for this installation. The 600 V No.12 AWG control cable o

nas voltege and current espabilities well above that required.

o,,

is 3 3

{

'da'i[@theregpitf(been l E{

~

control of the sc hnoid valves ii gp automatie iselstion capability.

Iso:

.c provided because:

ygeged 45Tgt)-CTION n.

l e

m 2-4 i :n, g 1.

The valves are always ki operation e, *

> eu The valves are required to be operated during a high drhell eNo 2.

pressure condition and must be available independent of u

reactor water level.

High drywell pressure and low-low

[ "E $

' ~~

reactor water level are the normai containment isolation

/

signals N

makeup and ventilation valves are also provided for use under anont.ecident conditiers. These will automatically close upon receipt of an accident signal. However, these valves may be used after an accident provided the required power supplies are available and a low-low water level signal is not present.

Refer to Section 5.2.3.5

~

and Tables 5.2-4 and 7.3-1.

O 5.4-3 Revision 2 - July 1983

Safety Evalustien No. 2. i tig Rev. N3.D Attcchm2nt 1 l

PNPS-PSAR She2t 9 of 19 i

5.4.5 Com=ustible Gas Monitoring The e xis:.r.g containment combustible Gas Honitoring System (CCG,MS) consists ef two redundant, remotely operable, seismically qualified analyzers. The hydrogen analyzers can continuously monitor hydrogen l

main Drywell h%-ogen concentra' tion and have a remote readout in the control reen.

5.4.6 Radiological Consequences of Containment Venting An evaluation of offsite doses which would be ir. curred as a result of containme2t venting to limit containment pressure has been performed in a manner consistent with Regulatory Guides 1.3,1.7, and 1.45.-

The resu'.ts of this analysis indicate that the doses to receptors at l

the LFZ wculd be well within the' limits of 10CFE100.

This analysis l

assumed that venting at the rate of 50 standard its/ min through the SGTS would be initiated at 80 hr after the reactor was made i

ical and venting would_ continue for 30 days, t,27

_,, taper tv6 pages) j l

5.4 {B3,, Reference s i

~

1.

GE Letter No. SSX:79-64.

2.

July 13, 1979 Letter, W. J. Neal (GE) to S. A. Giusti (Bechtel).

t 3.

BLE-459 dated September 25, 1975.

4.

Supplesent No.1 to Dresden Station $pecial Report No. 39 and QAD Cities Special Report No.14.

I3 5 U ~c ~> e... R rU f

! CONSTRUCTION u

l I

4 O

l 5.4-5

Safety Evaluation No. L84e/

Rev. O Sheet 10 of 19 5.4.7 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 during these accidents is estimated to exceed its design capacity. Thus, the primary containment fails, releasing reactor fission products to the secondary containment and potentially to the environment as well. The direct torus vent i

line (DTYL) provides an emergency primary containment vent path to prevent, or at least slow down, the buildup of potentially4n!asjpg,.p_ressure within e

YY

. h Y $ ' ?a Y h 5.4.7.2

System Description

pg, The DTYL is an 8" carbon steel pipeline connecting the 20" torus main adaust line to the underground 20" sain stack exhaust line. The 8" DTYL stares at a new brant between the existing 8" containment isolation valves in the r section of the 20" torus main exhaust line. The DTYL terminates in the 20" sain stack exhaust line, several feet downstream of the SGTS outlet valves.

The line includes an 8" air-operated, normally-closed butterfly valve which serves as the outboard containment isolation valve for the DTYL. Both electrical power and valve operator active gas (air or nitrogen) supply are taken from " essential" or reliable sources, or are backed-up to ensureJet_

the system is available during a station bla-c.,. i.;; gFiniW6iinnt71F"-~

i

!SSUED FOR u

ith' Winter '86 ' Adde'ndaK 3e"ctio'n k The DTYL mects ASME B&PV Code (1980 Edit. ion III, Subsection NC for Nuclear Class 2 requirements up-to andrincludin'g the." h isolation valve. The new piping downstream of thadsolttRii~iaVe ' meets ANSI B31.1 (1977 Edition through Winter 1979 Addenda) requirements.

During normal or general transient conditions, the DTYL outboard isolation valve would remain closed. In response to a severe accident, plant annagement could direct the control room operators to enploy the DTYL to relieve excessive pressure within the containment. In this case, the operator would follow a written procedure to perform the following basic actions:

Close, or confirm closed, the outboard isolation valve for the torus o

main exhaust line Close, or confiria closed, the SGTS outlet valves to prevent the high o

containment pressures from back-pressurizing the SGTS filters Open the two DTYL isolation valves o

Optimally, turn off the SGTS which likely came on automatically in o

response to a high drywell pressure signal I

ai=v...............

Rev. 0 Sheet 11 of 19 5.4.7.3 Radiological Consequences of DTVL Use The exhaust gases released by the DTYL during a severe accident would have initially beert %shed" by the, suppression pool water which would redule tne These exhaust gases are vented to the highest vent particulate released.

point (main stack), avoiding the ground level release of radioactive material in case of cont.aintnent failure due to over-pressurization.

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Sh:et 12 of 19 i

Traversing incere probe

)

i RHR reactor shutdown cooling supply I

I RHR reactor head spray Reactor water sample lines Reactor water cleanup og

"I Drywell purge inlet and makeup gas *, **

8 u

Drywell main exhaust

,7 -

em suppression chamber exhaust valve bypass *, **

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Suppression chamber purge inlet and makeup gas *, **

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suppression chamber main exhaust I

Drywell exhaust valvir:99 pass *:, --

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.. _....__ _ _-...---~

containment atmosphere sampling lines g,2 E t o.S 3 y~'!

containment makeup and ventilation valves are also

Cl g

provided for use following an accident cor.dition.

.7,{

These are remote manual operated.

Refer to eE$

Section 5.4.3.

g" 2 ',

e 5 E.

The reactor water low level isolation signal can be by-passed. These valves may be opened anytime. provided y b ';"

1 g%E the low-low water level signal is not present.

i *

l lj i

The second and lower of the reactor vessel low water level:

Eh, O t "k isolation settings was selected low enough -to allow the.

removal of heat from the reactor for a predetermined tisme; yg following the scram, and high enough to complete isolation; y 5 5-in time for the operation of C5CS in the event of a large

,b,3 g.

break in the nuclear system process barrier.

This second ora low water level setting is low enough that partial losses of '

j feedwater supply would not unnecessarily initiate full

/

isolation of the reactor, thereby disrupting normal shutdown

/

or recovery procedures.

Isolation of the following s... /

pipelines is initiated when the reactor vessel water level falls into this second setting.

All four main steam lines HPCI, RCIC, and main steam line drains 7.3-12 Revision 2 - July 1983

de ety tvaluatluri nw. y4 Rev. No.O l

PNPS-TSAR Attachment i

~

Sheet 13 of 19 included here to make the listing of isolation functions complete.

6.

Primary Containment (drywell) High Pressure High pressure in the drywell could indicate a breach of the nuclear system process barrier inside the drywell.

The automatic closure of various class B valves prevents the release of significant amounts of radioactive material from the primary containment. Upon detection of a high drywell pressure. the following pipelines are isolated.

See Table 7.3-1. Signal F.

Traversing incere probe RHRS shutdown cooling supply RHRS reactor head spray l

Reactor water sample lines Drywell equipment drain sump discharge l

Drywell floor drain sump discharge Traveling incere probe tubes I

Drywell purge inlet and makeu$ gas *,!SSUED O 1 a

i ig

..' [,,

, '.j, ( }.

J,.e

.-,...m..

-3 i

Drywell main exhaust i. ',

}

i 3

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suppression chamber exhaust valve bypass *, **

Suppression chamber purge inlet and makeup gas *, **

q sion chamb,er main exhaust O_j rect torus vent *** )

Drywell exhaust valve bypass *, **

I RER to Radwaste Ceetsinment atmosphere sampling lines The primary containment high pressure isolation setting is selected to be as low as possible without in& acing spurious isolation trips..See Table 7.3-1, $5.gnal F.

Containment makeup and ventilation valves are also provide d for use following an accident condition.

These are remote manual operated. Refer to section 5.4.3.

lv i

7.3-15 Revision 2 - July 1983 8

f y

Q -- -

&,e PNPS-FSAR Sheet 14 of 19

~

The reactor water low level isolation signal can be by-

)

passed. These valves may be opened anytime provided the low-low water level signal is not present.

7.

RCIC System Equipment Space High Temperature l

i High ten:perature in the vicinity of the RCIC System equipment could indicate a break in the RCIC steam line.

I The automatic closure of certain Class A valves prevents the I

excessive loss of reactor coolant and the release of l

significant amounts of radioactive material from the nuclear system process barrier. When high temperature occurs near the RCIC System equipment, the RCIC turbine steam line is isolated.

The high temperature isolation setting is selected far enough above anticipated normal RCIC system a

{

operational levels to avoid spurious operation but low 9

enough to provide timely detection o# 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 Direct torus vent valves are provided for use following ))

U) an accident condition. The primary containment (drywell high pressure isolation signal can be bypassed. These l

are rrmote manual operated valves.

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7. 3-15e Revision 2 - July 1983

KOV. h0. v Attachment i PNPS4SAR Sheet 15 of 19 The logic arra gement used for this function, shown on Figure 7.3-7 is an exception to the usual logic requirement. he:ause high steam flow is the second method of detecting an H K: turbine steam line break.

12.

HPCI Turbine Steam Line Low Pressure HPCI turbine steam line low pressure is used to automatically close the two isolation valves in the HPCI turbine steam line, so that steam and radioactive gases will not escape from the. HPCI turbine shaft secls into the Reactor Buildtug after steam pressure has decreased to such a low value that the turbine cannot bt operated.

The isolation setpcInt is chosen at a pressure below that where the HPCI turbine can operate efficiently. See Table 7.3.1,

~

Signal L.

13.

Reactor Water Cleanup System Space High Temperature High temperature in the vicinity of the reacter water cleanup (RWCU) equipment and piping could indicate a break in a RWCU line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier.

When high temperature occurs near the RWCU equipment, the RWCU system is isolated.

The high temperature isolation setting is selected far enough above anticipated nonnal system

_9 de%oua==-isoistforg::TeT" Tov operational levels to avoid s

eetion gof g $ipeJreak',;"See-h enough to provide timely l

eJ JV LtJ b*

Table 7.3-1, Signal J.

D *d '-

O h ]I Reactor Water Cleanup System High Uk-)M c[' I r. ![]'-

i r, V.,

14.

.. = = = = -.

RWCU high ficu could indicate a break in a RWCU la The automatic closure of certain Class A valves prer unts the excessive loss of reactor coolant, and the release of significant ameunts of radioactive materials from the nuclear system process barrier. Upon detection of EWCU high flow, the RWCU line is isolated. The high flow trip setting was selected high enough to avoid spurious isolation, yet low enough to provide timely detection of line break.

See Table 7.3-1, Signal J.

15. High Reactor vessel Pressure High reactor wssel pressure is used to automatically close the two isolat:.:n valves in the RHR pumps' shutdown cooling suction pipims so that the RHR low pressure piping vill not be threatened by overpressurization.

The isolation setpoint is chosen at a pressure below where the RER piping could,be

_ overpressurized. _See Table 7.3-1. Signal U.

,M.

See Insert C oe page 16.

7,3,l7

1 i

1 Safety Evaluation do. 2 844 Rev. O OfRT O Sheet }g of 19 Mg

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r h

Radiation in Reactor Building Refueling Floor Vent Ex t, Duct I.

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High radiat in refueling floor vent exhaust duct uld indicate a

/

gross release o ission products from the fuel igh radiation in i

[

refseling floor exh duct initiates isol on of the following pipe

\\

lines.

(See Table 7.3 ignal V.)

o Suppression chamber main e st o

Direct torus vent The high radiation setting is selected enough above background rad son levels to avoid spurious iso on, yet low enough I

to promptl ect a gross release of fission products m the fuel.

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h. I Suppression Pool High Radiation High radiation in torus could indicate a gross release of fission i

N ets from the fuel. High radiation in torus initiates isolation of the direct torus vent pipe line to prevent release of significant t

amounts of radioactive inaterials. The high radiation trip setting is l

1 selected high enough above background radiation levels to avoid f

1 spu-ious isolation, yet low enough to promptly detect a gross release of fission products fro:n the fuel to the main stack, within the allowable limit. See Taole 7.3-1 Signal Z.

L

Ph75-TSAR Att C nt 1 Sh2et 17 of 19 r

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13. High temperature in the spaces occupied by the RHRS e-

.Te (shutdo r.:

cooling) and piping outside the primary sqo containment is sensed by temperature switches that activate alarms only, indicating possible pipe breaks.

gggf

,I sUb A typical arrangement is' shown on Figure 7.3-8.

Au*.oc.a t i c s-m m

EOfN isolation oc high temperature is net required since the C x o en }

reactor vessel low water level asolation function is

$5 !

adequate in preventing the release of significant amounts of efIIo radioactive material in the event that either of these two T hMU i systems suffers a breach.

225*'

14. High temperature in the vicinity of the RWCU system is E' *k e,%E sensed by four sets of two binetallic temperature switches.

c5$xL A set of two temperature switches is installed in each of 7uS3 the four areas to be monitored: each set is a one out of two 5 W i",

trip system and capable of initiating isolation.

.m v, cc i

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15. High flow in the Reactor

dater Cleanup Syster supply line is

' 5 % $.

sensed by two differential pressure switches which monitor oStco the pressure difference across an elbow installed in the 8E

E a -

Reactor Water Cleanup System supply line.

The arrangement of the differential pressure switches is similar to that

o shown on Figure 7.3-12.

The. tripping of either switch

\\

jnitiates isolation of the RWCU mystem.

Channel nd logic relays are high reliability relays equal to type HTA relays made by the General Electric Company.

The relays are i

selected so that the continuous load will not exceed 50 percent of the cortinuous duty rating.

7.3.4.8.1 High Temperature Sensors e es f 5 P. 3 C 7 D i ! f"T[ r*j ::. !

S*T I

The

location, spatial independence.. Yn Yst'ahet'- U kp *fc W T'

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tripping of the high temperature senses ~2n7hiNa$"Ife# air.HntIMPfi"

[E8 EL j

turbine steamline and RCIC turbine steamline are detailed in this gyf section.

e e

x5 *w i Table 7.3-3 lists the areas outside the primary containment where

[#o*

main steam, F.PCI, and RCIC steam lines are routed. This table also lists the leak detection sensors, summarizes the physical separation

} TS E hTI '

of sensors, and specifies the set points at which asolation of the i d MS i respecting steam line would be initiated.

"Ee Potential leak sources and rates which would initiate isolation are

{I j,e as follows:

e m.3 ogg Main Steam Tunnel

%iE qg This area contains feedvater.

cleanup, and main steam piping.

gg, Isolation of the main steam lines will be initiated at ventilation exhaust temperatures from the main steam tunnel of 160*T to 170*T.

,,ygk 27g which would result from steam leaks equivalent to 10 gal / min and lf ygg )

greater. The feedvater and cleanup systems normally operate at 0:

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hunater haan unter laumi - seelste emia stems lies (essept

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end saase esaar lamp maves.

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use ba kt ta nun ekssdeum and head eseums (hip spese spa towersten alare anars as ames alasen).

tan sais stema line pressere as salat to smaa terwas (as

re ende anar).

las dryeall heare. slame aussaisemst opng galges.

S Imr remeter pressere peretasius to epse een syrer and M T

unisos.

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. heat essanager.

thenAt lign&d es. trol syehen assunted.

I 21*

Summe sanani evitah fMa emmesel ress.

hamster hash unter 1swel. Samhane unia egens 11as (easept

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PDC 86-52B FIRE WATER INTER-TIE TO RHR SCOPE TO PROVIDE A PIPING CROSS-TIE WITH A REMOVABLE SPOOL PIECE BETWEEN THE FIRE WATER MAIN LINE AND THE RHR LINE TO BYPASS THE 8" MANUAL ISOLATION VALVE WITH A 3" LINE CONTAINING A FLOW METER AND A GLOBE VALVE FOR THROTTLING' PURPOSES g

TO RECONNECT THE RHR LINE TO THE 4" RPV HEAD SPRAY LINE PGE 11

PDC 86-525 INTENT o TO MAKE FIRE WATER AVAILABLE FOR RPV INJECTION AND CONTAINMENT SPRAY e TO BE ABLE TO REGULATE THE FIRE WATER FLOW SO THAT FLOW RATE CAN BE OPTIMlZED WITHIN EOP REQUIREMENTS 10 CFR 50.59 CONSIDERATION THE FIRE WATER TIE-IN DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION i

TECH SPEC CHANGES NONE FSAR CHANGES SECTION 4.8 SECTION 10.8 FIGURE 10.7-1 FIGURE 10.8-1 FIGURE 4.8-1 TABLE 7.3-1 PGE 12

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PDC 86 52C DIESEL DRIVEN FIRE PUMP SCOPE TO ADD A NEW DIESEL FIRE PUMP TAKING SUCTION FROM THE CITY WATER LINE CAPABLE OF DELIVERING 750 GPM AT 125 PSI INTENT TO PROVIDE AN ADDITIONAL WATER SUPPLY FOR CONTAINMENT SPRAY AND RPV INJECTION DURING A GTATION BLACKOUT 10 CFR 50.59 CONSIDERATION THE INSTALLATION OF THE NEW DIESEL FIRE PUMP 1

DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION TECH SPEC CHANGES NONE FSAR CHANGES FIGURE 10.8-1 P&lD FIRE PROTECTION SYSTEM PGiE 13 6

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10 CFR 50.59 CONSIDERATION THE INSTALLATION OF THE FUEL OIL TRANSFER SYSTEM DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION TECH SPEC CHANGES NONE FSAR CHANGES SESCTION 10.8

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PROVIDE N2 CYLINDERS (20) AS AN AUTO BACKUP SUPPLY TO EXISTING N2 TANK O

PROVIDE AN ON SITE N2 TRAILER TRUCK (6000 GAL)

O PROVIDE CAPABILITY TO SUPPLY DRYWELL VALVES AND INSTRUMENTS FROM THE N2 TRAILER O

PROVIDE N2 SUPPLY PRESSURE IM0lCATION IN MAIN CONTROL ROOM O

ISOLATE INSTRUMENT AIR SUPPLY WITH A LOCKED CLOSED VALVE O

MODIFY DRYWELL N2 SUPPLY CONTROL VALVE TO Fall OPEN (VS FAILED CLOSED)

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SUPPLY NITROGEN FOR DRYWELL AND TORUS MAKE-UP PGE 19

PDC 86 53 10 CFR 50.59 CONSIDERATION THE BACKUP NITROGEN SUPPLY MODIFICATION DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION TECH SPEC CHANGE.

l NONE FSAR CHANGES SECTION 5.4-2 " CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT" FIGURE 5.4-1 AND 5.2-16

" CONTAINMENT ATMOSPHERIC l

CONTROL SYSTEMS" SECTION 10.11.3-1

" INSTRUMENT AND SERVICE AIR SYSTEMS" FIGURE 10.11-1

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i THIRD DIESEL GENERATOR SCOPE O

INSTALL A 2000KW SELF CONTAINED DIESEL GENERATOR WITH A FUEL STORAGE SUPPLY FOR SEVEN DAYS O

INSTALL A TWO BREAKER BUS TO FEED SAFETY BUSES A5 OR A6 FROM THE SHUTDOWN TRANSFORMER OR THE THIRD DIESEL O

INSTALL A 480V FEED FROM STATION TO MAINTAIN THE THIRD DIESEL GENERATOR AUXILIARIES READY FOR

" BLACK START" O

INSTALL CONTROLS TO ALLOW THE DIESEL AND ASSOCIATED BREAKERS TO BE STARTED / STOPPED / OPENED / CLOSED FROM THE CONTROL ROOM O

INSTALL CONTROLS TO ALLOW THE DIESEL GENERATOR TO BE SYNCHRONIZED TO THE SHUTDOWN TRANSFORMER FOR LOAD TESTING DURING NORMAL OPERATION PGE 17

PDC 86 56 A,8 INTENT TO PROVIDE AN ALTERNATE SOURCE OF POWER AFTER LOSS OF OFF SITE POWER AND FAILURE OF THE EMERGENCY DIESEL GENERATORS 10 CFR 50.59 CONSIDERATION THE INSTALLATION OF THE THIRD DIESEL GENERATOR DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION TECH SPEC CHANGES NONE 4

FSAR CHANGES SECTION 1.6 NEW POWER SOURCE 1

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TO ADD AN INHIBIT SWITCH TO PREVENT AUTOMATIC q

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TO ELIMINATE DEPENDENCE ON MANUAL OPERATOR ACTION TO INITIATE ADS BLOWDOWN CAPABILITY FOR EVENTS WHICH PRODUCE LOW VESSEL WATER LEVEL WITHOUT HIGH DRYWELL PRESSURE e

TO SIMPLIFY OPERATOR ACTION WHEN EOPS REQUIRE MANUAL OPERATOR ACTION TO BLOCK ADS ACTUATION PGE 21 6

PDC 86-73 10 CFR 50.59 CONSIDERATION e THE ADS LOGIC CHANGE DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION

COMPLIES WITH NRC REQUIRED CHANGE (NUREG 0737)

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" CORE SPRAY SYSTEM" SECTION 6.4.4 "RHR (LPCI MODE)"

SECTION 4.4

" NUCLEAR PRESSURE RELIEF SYSTEM" SECTION 7.0

" INSTRUMENTATION" PGE 22

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TO MAINTAIN THE SBLC SYSTEM PUMP REDUNDANCY 8

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P D C 6 6-102 FEEDWATER PUMP TRIP SCOPE PROVIDE REACTOR HIGH PRESSURE TRIP SIGNAL (1400PSIG) TO FEEDWATER PUMP BREAKERS INTENT TO AUTOMATICALLY REDUCE REACTOR POWER LEVEL AFTER AN MSIV INITIATED ATWS i

10 CFR 50.59 CONSIDERATION THE FEEDWATER PUMP TRIP MODIFICATION DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION TECH SPEC CHANGES NONE FSAR CHANGES SECTION 3.9.3.3 ATWS SYSTEM TRIP LOGIC SECTION 3.9.3.4 ADDED FEEDWTR PUMP TRIP SYS SECTION 3.9.1 ATWS DESIGN OBJECTIVES SECTION 7.10.3.6 ADDED REACTOR FDWTR PMP BRKERS FIGURE 3.91 (M144)

FIGURE 3.9-2 (M146)

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TECH SPEC CHANGES NONE FSAR CHANGES TABLE 4.7-2

" INSTRUMENT SPECIFICATIONS" PGE 25

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ML20238E692

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ML20238E692

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