ML20237D035

Rev 0 to Standby Liquid Control Sys Control Capacity Equivalency Rept

ML20237D035
Person / Time
Site:	05000000, Pilgrim
Issue date:	01/29/1987
From:	Earle R, Sawabe J GENERAL ELECTRIC CO.
To:
Shared Package
ML20237C946	List: ML20237D007 ML20237D035
References
FOIA-87-652 PDC-86-75, PDC-86-75-R, PDC-86-75-R00, NUDOCS 8712220325
Download: ML20237D035 (42)

Text

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I PDC 66-75, Rev. O l

Safety Evaluat on, Rev. 0 Sheetiof3 113 I GENER AL h ELECTRIC l

STANDBY LIOUID CONTROL SYSTEM 1

• 1 CONTROL CAPACITY 1

EQUIVALENCY REPORT

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. b PREPARED FOR THE BOSTON EDISON COMPANY t.

PILGRIM NUCLEAR POWER STATION JANUARY.29, 1987

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PREPAPED BY: R.T. EARLE z

E'C GAda Y21[?

VF.RITIED BY: J.K. SAWABE

).')t..Q l IA Mi i

Section L4 verification Material in ORF C41-00095/2 8712220325 071218 t

PDR FDIA FM

_DR; SORGIB7-652

1 PDC 86-75, Rev. 0 Safety Evaluati n, Rey, o Attachment G.,

Sheet 2, of c) 2.12 j i

DISCLAIMER OF RESPONSIBILITY This document was prepared by or for the General Electric Cocpany.

Neither the General Electric Company nor any of the contributors to l

1 this document:

l A. Makes any warranty or representation, express or ifeplied, with respect to the accuracy, completeness, or usefulness of the or that the use of any f

information contained in this document, J

inf ormation disclosed in this document may not infringe 1

I privately owned rights; or, Assumes any responsibility for liability or damage of any kind which may result f rom the use of any information disclosed in B.

l i

this document.

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PDC 86-75. Rev. O j

Safety Evaluation, Rev. 0 Q-2lg; l.

Sheet 3 of ej 4

AESTRACT y

This document was prepared for the Boston Edison Company'te address i

i

) at the the requirements of the Standby Liquid Control System (SLCS

]

Pilgrim Nuclear Power Station for compliance with the NRC ATJS. Rule s

10CFR50.62. The plant specific values used to demonstrate compliance with the NRC ATds Rule are the same as the stinimum values provided in 4

tha system Technical Specifications.

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PDC 86-7', Rev. 0 Safety Evaluation, Rev. 0

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Attachment i Sheetfcfc) 1 TABLE OF CONTENTS 11 Abstract 1

1.

Introduction l'

2.

Discussion 1

2.1 SL,C System Design Basis 1

2.2 NRC ATWS Rule e

1 3.

Analysis 2

3.1 Equivalent Control capacity Definition 3

3.2 Equivalent Control Capaci,ty. Calculation 4

l 4.

Summary 5

S.

References i

i 111 -

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PDC 86-75, Rev. O

{

Safety Evaluati n, Rev. O I

1. 83 8 Sheetfofp 1.0 IN RoDUCTION d

EBosten Edison has requested an evaluation of the minimum require L

concentration (weight percent) of soi.t= pentaborate f or the Pilgrim l

.Sthndby Liquid Control System to comp.y with the NRC AWS ru eThe minimum concentrat requirements in 10CTR50.62 (Ref erence 3. ).

is to be based on equivalency to the minimum 86 gpm,13-weight percent sodium pentaborate control capacity repairement stated in the NRC AWS i

l rule. Equivalency is calculated using the ratio of the specif c Pilgrirn minimum values to reference plant values that the rule is based on.

For Pilgrim, a minimum solution concentration of 8.42 percent is required. This is based on the assum-tions that the minimum

~

bgen enrichment of the sodium pentaborate decahydrate exceeds 54.

i one pump is required to operate and the actual B

atomLpercent, ccpacity of each pump exceeds the required minimum pump flow rate.

2.0 pISCUSSION h 2.1 SLC System Design Basis The generic design basis for the SLC System is to provide a specified cold boron shutdown concentration te the reactor vessel as describ The S*,C System was typically designed in NEDE-24222 (Reference 4.).

l to provide the specified cold shutd::wn concentration in'about one or two hours. During reload licensing evaluations, this shutdown concentration is verified by analysis to be adequate to render the core suberitical. The considerations in the reload evaluation are independent of AWS and injection rate is not directly considered.

The AWS rule requires the addition of a new design requirement to the generic SLC System design basis. Changes to flow rate, solution concentration or boron enrichment, to meet the ATWS Rule, must not i

invalidate the original system design basis.

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CC 56-75, Rev. 0 52fety Evaluatiqn, Rev. O

{2 g 3 Mtachment'2.

5 eet 6 of cy l

2.2 NEC A'I~a's Rule Paragraph (c)(4) of 10CTRSO.62 states, in part:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and bcron content in control capacity to 86 gallons per minute of equivalent 13-weight percent sodium pentaborate solution."

The NRC Staff has provided clarification of equivalent control 1

1 capacity (Reference 5.) as follows:

llow (1) The "equivalen't' in control capacity" wording was choosen to a flexibility in the implementation of the requiremetr. For example, the equivalence can be' ob'tained by increasing flow rate, boron concentration or boron enrichment.

sodium pentaborate

' 12')' The 86 gallons per minute and 13-weight percent i

i of ATWS, ware values used in NEDE-24222, " assessment of BWE P.it gat on

/6 plants Volumes I and II", December 1979, for BWR/4, BWR/5 and BWR ifferent values with a 251-inch vessel inside diameter. The f act that d E 24222.

"The flow rates given here are normalized from a 251-inch i.e.,

diameter vessel plant to a 218-inch diameter vessel plant, i

lent the 66 gpm control liquid injection rate in a 218 is equ va i

(p to 86 gpm in a 251. This is done to bound the analys s.... p.

2-15 [NEDE-24222))."

i l nce are vessel b'oron concentration required to achiev d the

~

i The time required to achieve that vessel boron concentrat on.

h minimally acceptable system should show an equivalence in t e 24222.

l parameters to the 251-inch diameter vessel studied in NEDE-1 3.0 ANALYSIS

. 3.1 Equivalent Control Capacity is a very The NRC equivalent, control capacity concept of the ATWS rule i

f the simple, direct criterion that does not require considerat on o cific core nuclear

. mixing efficiency or to account for plant-specharacteristics liance with

PDC 86-75. Rev. 0 f

Safety Evaluation Rev. O

-Attachment '2, Q,24 9, Sheet}lof9 if the following relationship is shown to l

the equivalency requirement be true:

(Equati,on 1)

E

>=

1 p*U251

• C

_19.8 86 M

13 where the plant-specific parameters are defined as:

Q = minimum SLCS flow rate (one or two pump operation as appropriate), gpm.

M = mass of water in the reactor vessel and recirculation system at the hot rated conditions, 1bs.

i ht C = minimum sodium pentaborate solution concentration, we g percent.

isoto'pe enrichment (19.8% for natural E = minimum expected B boron), atom percent.

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(the mass of water in the reactor vessel and l

t is

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recirculationshstematratedconditionsinthereferencepan The value of M d on rated lbs for a BWR/3/4. This value was calculated baselevel, control rods 628,300 temperature, rated void content, normal waterfully withd l

internals dimensions.

3.2 Equivalent control capacity calculation The NRC requires the use of minimum ' plant-specific values *to i

l

demonstrate compliance with the equivalency requirement. For s n ilgrim can pump operation, S4.5 atom percent boron enrichme (Equation 2.)

• 86

• 19.8 C >=

13

• M M

0 251

/

t tion (weight where c is this case is the minimum allowed concen raof llowed individual pump flow rate, M is the mars of waterreci percent) ichment

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l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -

PDC 86-75. Rev. 0 Safety EvaluatioD, Rev. 0

{2yy, Sheetgof9 I

..s 1evel. The water mass is based on the same conditions as l

t was plant _ water mass. The minimum allewed individual pump flev ra e level obtained. f rom Reference 1. The minimum allowed boron enric i

will become part of the system Technical Specification and Des gn Sp cification (Reference 1).

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j Q = 39 (minimum rated) gpm M = 507,850 lbs E = 54.5 %

2) gives Using the current Pilgrim plant-specific values (in E o

pentaborate.

(Equation 3.)

13

• 507,850

• e6

• 19.8 C >=

628,300 39 54.5

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C>=

B.42 4.0

SUMMARY

hd te When the concentration of enriched sodium pentaggrate deca y ra

) is equal to, or (enrichment exceeding 54.5 atom percent boron B S rule, greater than 8.42 percent, Pilgrim meets or exceeds the NRC equivalency requirements.

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PDC B6-75. Rev. 0 Safety Evaluati n, Rev. O Attachment %

Sheet 3 of 9

,2, f g j 5.0 ~rETERENCES O, Standby Liquid control System

1. Doc. No. 257HA169, Rev.

Design Specification.

2, Standby Liquid Control System

2. Doc. No. 257HA169AV, Rev.

Design Specification Data Sheet.

3. 10CFR50.62, NRC ATWS Rule, June 1984.

b

4. NEDE-24222, Assessment of BWR Mitigation of AWS,' Decem er 1979.

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S. USNRC Generic Letter 85-03, Clarification of Equivalentcontrol Capacity fo l

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Liquid Control Systems, January

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28, 1985.

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

PDC 86-75, Rev. O Safety Evaluation No. l. l31 Rev. O Sheet of ATTACHMENT 3 Recommended Technical Specification Changes The pages of the following sections and figures of the Technical Specifications that need to be updated due to SLCS modification (PDC 86-75)

- have been marked with suggested updates and included in this attachment for your review.

Technical Specifications Sections 3.4 & 4.4 Technical Specifications Figures 3.4.1 & 3.4.2 I'C l

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l SURVEILLANCE RFAllRIMEN75 LIWITING CONDIT1pg FOR OPERATION 4.4 ST ANDBY L10VID CONTROL S'S~EM 3.4 ST ANDBY L10010 CONTROL SYSTEM Aeolicability:

Aeolicability:

Applies to the operating status of the Standby Liquid Control Applies to the survelliance System.

requirements of the Star.ccy Liquid Control System.

Obiettive:

Obiective:

To assure the availability of a system with the capability to To verify the operability of the shutdown the reactor and maintain Standby Liquid Control System.

the shutdown condition without the use of control rods.

Specification:

Specification:

~

A.

Normal System Availability A. Normal System Availatitity 1.

During periods when fuel is in the reactor and prior t he op bility of W S W W sta Liquid Control Systes shall be n t n*

h ndby Liquid verified by the perform.ance of Control System shall be the following tests:

operable, except as specified in 3.4.B below. This system 1.

At Imt m p m"m need not bE operable when the each pump loop sha'll be f

t tested by n

on a c n rol r

rods are fully inserted and demineralized water to the Specification 3.3.A is met.

test tank.

2.

At least once during each operating cycle:

a.

Check that the system relief valves trip full open at pressures less than 1800 psig, and reseat on a falling pressure greater than 1275 psig.

b.

Manually initiate the system, except explosive valves.

Pump boron solution through the recirculation path and back to the Standby Liquid Control Solution Tank. Check that each pump flow rate exceecs 39 GPM against a system head of 1275 psig.

?*: TING COC*~ IONS FOR OPERATION SURVE1LLANCE REOUTREMENTS t

3. t.

STAND ** L10UID CONTROL SYSTEM 4.4 STANDBY L100fD CONTROL SYSTEM c.

Manually initiate one of the Standby Liquid Control System loops and pump demineralized water into'the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper operation of the valves, and pump operability. The replacement charges to be installed will be selected from the same manufactured batch as the tested charge.

d.

Both systems, including both explosive valves, shall be tested in the j(

course of two operating cycles.

B. Surveillance with Inocerable B.

{#eration with Inocerable Comoonents:

[opconents:

1.

When a component is found 1.

From and after the date that a redundant to be inoperable, its redundant component shall component is made or be demonstrated to be found to be inoperable, operable immediately and Specification 3.4.A.)

daily thereafter untti the shall be considered fulfilled and continued inoperable component is repaired.

operation permitted provided that the component is returned to an operable condition within seven days.

1 96 4

l

SURVEILLANCE REQUIREMENTS

,,. LifENG CONDITIONS FOR OPEPal*&N 3.4 ST ANDSJ_LIDUID.. CONTROL SYS'EH 4.4 STANDBY LIOUID CONT E SYSTEM r

<i C.

52dium Pentaborate Sciution C. 5941u9 Pentaborate Solutian At all times when the Standby The following tests shall be Liquid Control System is performed to verify the required to be operable the availability of the Liquid following conditions shall be Control Solution:

met:

1.

Volume: Check at least 1.

The net volume -

once per day.

concentration of the Liquid Control Solution 2.

Temperature: Check at in the liquid control least once per day.

tank shall be maintained as required in Figure 3.

Concentration: Check at 3.4.1.

least once per month.

Also check concentration 2.

The temperature of the anytime water or boron is liquid control solution added to the solution, or shall be maintained above the solution temperature 48'F.

is at or below 48'F.

3.

The enrichment of the 4.

Enrichment: Check liquid control solution Boron-10 enrichment level maintained at a by test anytime boron is shallby0 added to the solution and boron B isotope enrichment exceeding 54.5 prior to restarting from D

atom percent.

each refueling outage.

Enrichment analyses shall D.

If specification 3.4.A. B, or be received within 30 days C.1 or C.2 cannot be met, the of test performance. 4dhom reactor shall be placed in a tn t =m : =

Cold Shutdown Condition with r a check shall te mace to 7 all operable control rods ensure that Boron levels fully inserted within 24 meet the original design hours. If the enrichment criteria by comparing the requirements of specification

/

enrichment, concentration 3.4.C.3arenotmet[he6eg

/ and volume to established

' criteria. IT the Sc*ob

[the Boron-10 IsotoMrenrichment top 4.5 Atom '

levels ofo hof meet 6

percent within seven days th9n Tf+bC b

' tbt or'8S peil dess$h CPNcNb i

1 levels ince.i from the _ time of enrichment fl be f ecealin 9 l

"'F l recess.fsubmit a report to

/ NE PCACSOP Sb4 the JNRand advise them of Gld Nidown (cnd. hon

',th E

gli o conf *f rods Twil/

O S*"Retable.u m 24 potas,

[planstobringthesolutionup to a demonstratable 54.5

' des ')h L*d

'< s crited 5) atom percent Boron-10 N AY4ed t b 5 7 ttn 8- [*e M o ol l bvb Isotopic Enrichment.

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L___________________

i PDC 86-75, Rev. 0 Sefety Evaluati

, Rev. 0 Attachnent 't Sheetgof9 2/3y Icvel. The water mass is based on the same conditions as t t

was plant water mass. The minimum allowed individual pump flow ra e level

' obtained f rom Ref erence 1. The minimum allowed boron en:ic d Design will become part of the system Technical Specification an Sp2cification (Reference 1).

p = 39 (minimum rated) gpm M = 507,850 lbs E = 54.5 %

2) gives Using the current Pilgrim plant-specific values (in Ec o

pentaborate.

(Epation 3. )

13

• 507,850

• B6

• 19.8 C >=

628,300 39 54.5 a

M' 8.42 C. >=

4.0 SUFF.ARY i

t When the concentration of enriched sodium pentaggrate decahyt ra e

) is equal to, or (enrichment excetding 54.5 atom percent boron B WS rule <

greater than 8.42 percent, Pilgrirn meets or exceeds the NRC A equivalency requirements.

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1 PDC B6-75 Rev. 0 l

Safetf Evaluati n, Rev. O q

Att W p.ent 9.

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Sheet,9ofy

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t, 5.0 EEFERENCES I

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O, Standby Liquid Centrol System

1. Doc. No. 257EA169, Rev.

Design Specification.

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2. Doc. No. 257HA169AV, Rev. 2, Standby Liqui.d Control System Design Specification Data Sheet.

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3. 10CFR50.62, NRC ATWS Rule, June 1984.

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4.. NEDE-24222, Assessment of BWR Mitigation of AWS/ December

/

1979.

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Letter 85-03, Clarification df Equivalent

5. USNRC Generft-Control Capacity for' Standby Liquid control Systems, JanucW ")

l' 28, 1985.

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7 PDC 86-75. Rev. 0 l

t.i. f i

e' Saiety Evaluation No. 2 is i Rev. 0

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Sheet of a

t ATTACHMENT 3 1

j Rtcornended Technical Specification Changes o

4 1

figures of the Technical The pages.of thE 'ej.ltdnyq4ctions s),det,thSLCS modification (PDC 86-75) l Spet'ifications trat need to,be updatM have been markey with'/ sum 45ted updates and included in this attachment I-L

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if your. review.

l TecNnical SpecNicatjens Sections' 3.4 & 4.4

'[ < Te-firical Specifications figures 3.4.1 & 3.4.2 l

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LMITING COND1710NS FOR OPERATION SURVEILLANCE REQUIREMENTS 3..c STANDBY L10Q1D CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTE*

Acolicability:

Acolicability:

Applies to the operating status of the Standby Liquid Control Applies to the surveillance 1-System.

requirements of the Standby Liquid Control System.

Obiective:

Obiective:

To assure the availability of a system with the capability to To verify the operability of-the shutdown the reactor and maintain Standby Liquid Co,ntrol Systen.

l the shutdown condition without the use of control rods.

Specification:

Specification:

~

A.

Normal System Availability A. Normal System Availability 1.

During periods when fuel is in the reactor and prior t The operability of the Standby Liquid Control System shall be cond on, he S ndby Liquid verified by the performar<e of Control System shall be the following tests:

operahle, except as specified in 3.4.B below. This system 1.

Atleastonceperedth need not be operable when the each pump loop shall be g

tested ky rol n

o a r

t g rods are fully inserted and demineralized water to the Specification 3.3.A is met.

test tank.

2.

At least once during each operating cycle:

a.

Check that the system relief valves trip full open at pressures less than 1800 psig, and reseat on a falling pressure greater than 1275 psig.

1 b.

Manually initiate the system, except explosive valves.

i Pump boron solution through the f

recirculation path and back to the Standby Liquid Control Solution Tank. Check that each pump flow rate exceeds 39 GPM against a systen head of 1275 psig.

I

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PITING CONDIT 3 5 FOR OPERATION SURVE1LLANCE REOUTREMENTS 3.4 STANDBY L1'UID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM c.

Manually initiate one of the Standby Liquid Control System loops and pump demineralized water into*the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper. operation of the valves, and pump operability. The replacement charges to be installed will be selected from the same manufactured batch as the tested charge.

d.

Both systems, including both explosive valves, shall be tested in the course of two operating cycles.

B.

Ooeration with Inocerable B. Surveillance with Inocerable Comoonents:

Cz.anonen t s :

1.

From and after the date 1.

When a component is found that a redundant to be inoperable, its component is made or redundant component shall found to be inoperable, be demonstrated to be l

Specification 3.4.A.1 operable issnediately and i'

shall be considered daily thereafter until the fulfilled and continued inoperable component is operation permitted repaired.

provided that the component is returned to an operable condition within seven days.

l 96

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. LIu1T1NG CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.4

$TANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM C.

Sodium Pentaborate Solution C. Sodium Pentaborate Solutien At all times when the Standby The following tests shall be Liquid Control System is performed to verify the required to be operable the availability of the Liquid following conditions shall be

. Control Solution:

met:

1.

Volume: Check at least 1.

The net volume -

once per day.

concentration of the Liquid Control Solution 2.

Temperature: Check at in the liquid control least once per day.

tank shall be maintained as required in Figure 3,

Concentration: Check at 3.4.1.

least once per month.

Also check concentration 2.

The temperature of the anytime water or boron is liquid control solution added to the solution, or shall be maintained above the solution temperattrre 48'F.

is at or below 48'F.

3.

The enrichment of the 4.

Enrichment: Check liquid control solution Boron-10 enrichment level shall bg maintained at a by test anytime boron is boron B'O isotope added to the solution and enrichment exceeding 54.5 prior to restarting from h-atom percent.

each refueling outage.

Enrichment analyses sh.all D.

If specification 3.4.A, 8. or be received within 30 days C.) or C.2 cannot be met, the of test performance. 4dhen reactor shall be placed in a

^::t nd h :=

Cold Shutdown Condition with Fa _ check shall be mace to 7 all operable control rods ensure that Boron levels i

fully inserted within 24 meet the original design hours. If the enrichment criteria by comparing the i

requirements of specification

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enrichment, concentration 3.4.C.3arenotmet[he6ag

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and volume to established

[the Boron-10 Isotonicenrichment to/54.5 Atos ^ N8"

' criteria. IT the Scwb

( g gglo hof meet l g,p gggyg%

fevels N OC b*'

percent within seven days

+h4n 1

levels %ec.i from the__ time of enrichment recess.4submitareportto

/ NE Petetor shyll H t ecealin 9 l

h* D"l the NR and advise them of Gld Gigidown bd.h'on Mth gli o FwlI7 /

D5""pera6le*co h1NJ tods!" 24 eio ns.

j d35hh p ans o bring the solution l*d ?'

cvhcW4s up to a demonstratable 54.5 hv%

atom percent Boron-10 ff Af4cp ths Dtna p e h o ol Isotopic Enrichment.

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Safety Evaivation me.: 2. IS I SAFETY tvALUATION PILERIM WUCLE AR P0htIR STATION Rev. No. 0 PDC PCW System Calt.

Initiator:

Dept:

Grou;;

mo.:

Name:

No.:

Date: 5/4/g7 Tra. W atke neb F S S M c.

8 6 - 7 5 j.+*5 j67 S@DS 9 4,,)

t S 7-SO4-Description of Proposed change, test or experiment: -

hmekeel Bet on L J r T u*c s 4 k u 4e GLc h L wie. I Spec M e. + ten and Cha m e 5.

assoc u n 4 e ot

$AF[TY [YALUATION ComCLU510W5:

The proposed change, test or experiment:

(M Does Not 1.

consequences o(f an accident or malfunction of equ

) Does increase the probability of occurrence or safety previously evaluated in the FSAR.

o 2.

(X) Does Not ( ) Does create the possibility for accide t of a dif ferent type than any evaluated previously in the F5AR n or malfunction 3.

for any technical specification.DQ Does Not ( ) Does reduce t s

l 3A$f5 FORA $AFETY EVALUAff0W CONCLUSIONS:

Ser A +4a c k e ed

% e+s

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Change bd Recossended Change

( ) Not Recommended N ) SE Performed by M

AL Date _

6!P7 r

Exhibit 3.07-A Sheet 1 of 3 Rev.3 1

1 s is.

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a Safety Evaluation No. _20 s k.e N5 IS7 l

l A.

Dest-intion of Prongsed Chance. Test or Eroeriment:

This rnodification replaces the Standby liquid Control Systems's (SLCS) existing sodium pentaborate solution (natural boron with 19.8 atom percent B10) of 9.4 to 16 weight percent concentration with an enriched sodium pentatorate solution (bofi enriched greater than 54.5 atom percent BIO) of 8.42 to 9.22 weight percent concentration.

In addition, this modification recalibrates and revises the setpoints of the level and temperature sensing instruments, relocates the SLC pump 207B test button near the test button for pump 207A and locates a new pressure gauge near the test buttons.

instrumentation is required, due to the solution concentration chang The SLC's Technical Specification change includes the solution concentration requirements, surveillance requirements, and bases for the new enriched sodium pentaborate solution.

B.

Puroose of Chance The enriched sodium pentaborate solution is being added to the SLCS to comply with the NRC's Anticipated Transient Without Scram (ATHS) Rule (10CFR50.62).

The reduction in maximum allowable solution concentration from 16 to 9.22 weight percent reduces the maximum solution saturation temperature from 70*F to 38'F.

This reduces the possibility of Technical h

Specifications requiring reactor shutdown as a result of solution temperature requirements.

The additional system changes are being performed to simplify testing and minimize enriched sodium pentaborate loss.

C.

Systems Subsystems. Comoonents Affected:

Standby Liquid Control System - This modification affects the SLCS in the following manner:

The performance of the system is improved by this modification.

The modified system performs at increased reactivity control capacity to meet the NRC ATHS Rules equivalence requirements of 86 GPH/13 weight percent of normal sodium pentaborate solutic,n.

If the enrichment option was not used, two pumps would be required to meet the NRC's ATHS Rule.

With the enrichment option, the reliability of the system is maintained, since only one pump is required to satisfy the NRC's ATWS Rule.

The system retains one redundant pump.

The low crystallization temperature (38'T corresponding to a 9.22 weight percent concentration) of the enriched sodium pentaborate solution will further improve the system reliability.

reactor shutdown because of solution temperature requirements.This reduces the possibil

_1 t

- _ - - - - - - - - ~ - - ' ' ~ ~ ~

Sakty Evaluation No.

• k P.s c - s ta % ~r Tne storage tank high and lowlevel alarms are being maintained at their original volume setpoints.

The high level alarm alerts the operator to a solution volume near the storage tank overflow.

The original vclume concentration requirements were such that, should evaporation occur, a low level alarm would annunciate before the temperature-concentration requirements were exceeded.

For the original solution, the maximum possible attainable concentration at the low level alarm was 14 weight percent.

This corresponded to a saturation temperature of 60*F which is less than the original 65*F setpoint of the heat tracing. This ensured the operator was given an alarm before crystallization could occur from high solution concentration. The requirement for a low level alarm to annunciate before temperature-concentration requirements are exceeded is not needed because of the new lower solution concentration requirements (8.42 to 9.22 weight percent).

Since the maximum concentration of the new solution is 9.22 weight percent, the maximum possible solution concentration obtainable from evaporation without a high or low level alarm is approximately 10.3 weight percent *,

This corresponds to approximately a 44*F solution temperature (low solution temperature alarm setpoints is 48'F).

Due to the 53*F setpoint of the tank heater and heat tracing and the design room temperature of 60*F to 100*F, solution concentration changes due to evaporation would be slow.

The operator would be alerted to a solution concentration change from evaporation by either the low level alarm or the technical specification monthly surveillance requirements before the crystallization point is reached.

• Hiah level Alarm (9.22) - (4430) (9.22) - 10.3 weight percent Low Level Alarm

('595D)

G.

Summary The SLCS by itself cannot cause an accident and it does not interact with any other system whose malfunction could cause an accident.

Hence, this modification on the system does not increase the probability of occurrence of an accident.

This modification increases the system's control capacity to satisfy NRC ATHS rule requirements. The modified system is more effective than the existing system in bringing the reactor to the cold shutdown condition from rated power. Hence, the modification does not increase the consequences of an accident.

This modification does not call for the safety equipment of the system to work at higher pressures, temperatures and more severe conditions than the existing levels. The modification makes the SLCS pumps redundant and it does not change the logic of the system. Hence, the modification does not increase the probability of the malfunction of the equipment important to safety.

This modification increases the margin of safety for system availability by reducing the possibility of system unavailability from solution temperature requirements.

This modification increases the margin of safety for flow rate requirements (required 39 GPH; available 78 GPM) and for minimum volume of solution requirements (required 2068 gallons at a mid-range concentration of 8.82 percent; available 3960 gallons).

Safety Evaluation No. m b*

%.c-5/-Q ' L *'

f The upper limit, 9.22 weight percent, concentration of enriched sodium

~

pentaborate has a saturation temperature of 38'F.

To preclude precipitation, the minimum solution temperature will be maintained above 48'F, which is 10*F above the saturation temperature of the maximum concentration.

In order to ensure a solution temperature greater than 48'F, the technical specifications will require determination of the solution temperature daily.

This frequency is considered adequate because the room minimum design temperature is 60*F, and any temperature change would be gradual.

In addition, the daily monitoring will be barcked up by the tank heater, heat tracing, and low temperature alarms.

If the solution temperature in either the tant or pump suction lines reaches 53*F, the tank heater or heat tracing will commence operation.

If the solution temperature in either the tank or suction lines continues to dro; to 48'F, the operator will receive an alarm in the control room.

Technical specifications will then require that the reactor be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a solution temperature less than 48'F.

In order to comply with the ATHS Rule (10CFR50.62), the boron in the

~

percent of Bgrate solution must be enriched to greater than 54.5 atom sodium penta 10 The technical. specifications will require that the B enrichment be greater than 54.5 atos percent.

If the B10 enrichment is found to be less than or equal to 54.5 atom percent, the Technical Specifications will require the operator to determine if the original shutdown criteria (equivalent of 703 ppm of natural boron) can be met.

If the original shutdown criteria can not be met, Technical Specifications will require that the reactor be placed into cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the original shutdown criteria can be met, Technical Specifications will require that the B10 enrichment be returned to greater than 54.5 atom percent within seven days.

If at the end of this seven day period, B10 enrichment is still less than or equal to 54.5 atom percent, Technical Specifications will require that the NRC be notified within This ensures that if the B g to bring the enrichment into compliance.

seven days with BECo's planI enrichment is less than or equal to 54.5 aton percent, the operator will shutdown if the original shutdown criteria cannot be met, or bring the enrichment into compliance if the original In order to ensure a B10 enrichment grea the Technical Specifications will require that the Bgr shutdown @ teria can be met.

than 54.5 percent enrichment be determined prior to restart from a refueling outage or any time boron is addef0to the storage tank. This frequency is considered is a stable isotope and enrichment changes can only

. adequate because B occur when additional boron is added.

In addition to ensure that the 10 boron added is enriched properly, station procedures will require that B enrichment be determined as part of the receipt inspection before release the matrial for use. Technical Specifications will also require that o{0enrichmenttestresultsbeknownwithin30daysofsamplingthe Bmaterial in the Standby Liquid Control Storage Tank. The 30 day time period allows sufficient time to perform the enrichment test and receive the test results.

It is considered adequate from a safety point, due to the station procedure requirement to determine enrichment as part of the receipt inspection before release of the material for use.

The requirement to determine enrichment after the addition of boron to the storage tank functions as a backup check to the station procedures.

--__.m__________

Safety Evaluation 1

No.

c2. U (W. c - Gl1 's,7 of the system is improved by this modification due to The response t']B 0 injection into the reactor.

higher rate of The relocation of the test button for SLC pump 207B and the addition of the pressure gauge will facilitate the system testing and does not affect the safety performance of the system.

1 D.

Sa f et y Function c' Af f ected Systems /Comoonents The safety function of the SLC system is to provide a backup method, which is independent of the control rods, to maintain the reactor subtritical as the nuclear systen cools, in the event that not enough of the control rods can be inserted te counteract the positive reactivity af fects of a colder moderator (Ref PN?S-FSAR, Rev. 6, Section 3.8.1).

This modification has an impact on the safety analysis (Ref. PNPS-FSAR Section 3.8.4) and the Technical Specification Section 3.4 which need to be updated to include the NRC ATHS Rule (10CFR50.62) requirements.

E.

Effects on Safety function The enriched sodium pentaborate modification to the SLC will upgrade the system to the rea:tivity control capacity requirements of the NRC's ATHS Rule (10CFR50.62) and still provide the equivalent of 700 ppm of natural boron to maintain the original system shutdown requirement.

The low crystallization temperature (38'F corresponding to a a 9.22 weight percent concentration) of the enriched sodium pentaborate solution allows the reduction of the tznk heater and heat tracing setpoint to 53*F.

This temperature is 5*F above the low temperature alarm setpoint of 48'F.

The low solution crystallization temperature and the new tank heater and heat tracing setpoint will reduce the possibility of reactor shutdown because of solution temperature requirements.

The adation of the pressure gauge and the relocation of the safety relateu test butbn for SLC pump 207B will not have any adverse effects on the safety functions of the SLC system. Materials for these changes will be procured, installed and tested in accordance with safety related requirements.

F.

Analysis of Effects on Safety Functions As per GE analysis (see Attachment 2 to this safety evaluation), use of an 8.42 or greater percent concentration of enti hed sodium pentaborate (enriched to greater than 54.5 atom percent bio) will meet or exceed the NRC ATHS Rule 10CFR50.62 requirements of the SLCS at Pilgrim Wuclear Power Station. This analysis is based on an injection rate of 39 gallons per minute (Ref. 1 IL GE Calc. No. DRF C41-00095/2 L4, Sht. 13A, SLCS Volume &

Concentration Chart). As each pump of the system has a minimum discharge capacity of 39 gallons per minute, the design is adequate to satisfy the NRC ATHS Rule requirements.

The minimum concentration of 8.42 percent and aB10 enrichment greater than 54.5 percent provides a total margin of 136 percent beyond tme amount needed to shutdown the reactor. _ - - - _ _ _ _ - - _ _ _.

l Safete Evaluat':c No,di38 G, c, c. s /.s :-

l

'ne technical specification changes will provide adequate operational and surveillance requirements for the SLCS modification and will not reduce tre targin of safety.

This modification does not involve an unreviewed safety question.

l

eferences 4

1.

General Electric Company Letter, Pilgrim ATHS SLC System Modification.

R. G. Ferguson to R. N. Swanson, dated 2/2/87.

e 1 i

l

/

I

Safety Evaluation <2 G (

SAFETY EVALUATION P!LGRIM NUCLEAR POWER STATION Rev. No.

(0 )

A.

APPROYAt

( )

This proposed change does not involve a change in the Technical Specifications.

@ This proposed change, test or experiment does ( ) does not h involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2).

Q This proposed change involves a change to the FSAR per 10CFR 50.11(e) and is reportable under 10CFt50.59(b).

(4 Comments:

IN ve/wr D.

(f d o p -t' The safety evaluation basis and conclusion is:

% Approved

() Not Approved sb PTA c)V/y7 All Shf87 iscipline Tdoup Leader'Dite

/

Supporting Discipline W Leader /Ba3 3.

REVIEW APPROYAL O

la cra Pu4 2i a.pphes 41bt ud;4dw

(

Comments:

% W Drc cturd e f boren s e

W.'

h enc N h Y

~

$f9/y7 5&SA$roup Leader /Date

~

Qkf7 C.

ORC REVIEW

() This proposed change involves an unreviewed safety geestien and a request for authorization of this change must he filed with the Directorate of Licensing, IRC prior ta tuplementaties.

() This proposed change does not involve as unreviewed' safety question.

Ott Chairman late ORC Meeting Number cc:

Exhibit 3.07-A Rev. 3 1

Sheet 2 of 3 i

Safety Evalution No.:

Lb.

SAFETY EvatuATION WORK SHEET Rev. No. O A.

System Structure Component Failure and Consequence Analyses.

System Structure Component Failure modes Effects of Failure Conments 1.

SLtS SKS 5. lion Sec. A fkc. bed 'Sbeef

%P 4 3t?F 2.

SLcs L.Gl;+y k See Amek,ol Sked SLtolesn i

Mc_ Rea c 4&

by SLC

~

General Reference Material Review FSAR CALCULATIONS REGULATORY SECTION PNPS TECHNICAL SPECS.

DESIGM SPECS PROCEDURES SUIDES STANDARDS CODES l.2

3. 4 / 4. 4-6E SPEC. 2 5' 7 N A 1(A locFe 50. 6 Z T

2,3 3,%

h

5. L 14, L A PPEkm G l

B.

For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each l

failure mode, show the consequences to the systes, structures or related l

components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix 6).

A 4!GE7 Frepared by Date l

i NOTE:

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

Exhibit 3.07-C aev. 2 I

l' i

I

No. D *%*

kr. c - W4 H )

SAFETY EVALUATION HORK SHEET A.

System / Structure / Component Failure and Consequence Analyses 1.

System / Structure / Component:

Standby Liquid Control System Failure mode:

SLCS solution temperature less than 38'F.

Effects of Failure:

Enriched sodium pentaborate solution crystallizes in the pump suction pipe rendering the system inoperative.

Coment s :

In order to ensure a solution temperature gteater than 38*F;

1. Solution temperature will be determined daily.

2. Tank heater and heat tracing commence operation when the solution temperature reaches 53*F.

3. Solution temperature of 48'F will alarm in control room, reactor must be placed into Cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of alarm.

Ih 2.

System / Structure /Comoonent: Standby Liquid Control System Failure Mode:

Inability to shutdown the reactor.

l Effects of Failure: Core damage, release of radioactive materials.

i Comments: Analysis performed by General Electric to assure the modified Standby Liquid Control System will provide the equivalent of 700 ppm of natural boron to maintain the original shutdown requirement.

1

PILSWin 51 A110N F5AR aEvitti SHEET

'.Referen:es:

I/4b7 Safety Evaluation:

EbC B6 75 nev. me.:

O cate:

Support a change L.ist FSAR ' test. diagrams, and indices af f ected by this change and corresponding FSAL revision.

Affected FSAR Revision to af fected FSAR Section is shown on:

Section Preliminary Final SECTeoA5 3.8.1 3.B.4,3.s,5,&3.8.6Attachment 1 3

1 FsG 3.%.i M 24cy, geg, g 7 Fa c. 2. ir. E M IF-2.3 FIG. 3.%. 3 & J. 9. G, Attachment i I

Attachment i

Attachment i

PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation

~

prepa ration).

9 I7 N kuuA /Date: 4f(!F7 Reviewed by:

ate:

Prepared by:,

v Approyed by: 6A

/Date: M/U l

~

FTWAL FSAR REY!Sf0N (Prepared following operational turnover of related systees structures of components for use at PNPS).

(1) i Prepared by:

/Date:

Reviewed by:

/Date:

i' (1) Attach completed FSAR Change Request for1m (Refer to NOP).

Exhibit 3.07-A Rev. t Sheet 3 of 3 l

l PDC 66-75, Rev.

g 242) I Safety Evaluation, Rev. O l

SheetIof 16 ATIACHMDC 1 RECOMMENDED FSAR CHMGES The pages of the following sections, ables & figures of the FSAR that need to be updated due to SLCS modification (P 86-75) have been marked with i

suggested updates and included in this attachment for your review.

FSAR Sections:

3.8.3, 3.8.4, 3.8.5, 3.8.6 G ste Tar +c:

?. 5 - h-3. 0-2 W FSAR Figures:

3.8-3, 3.8-6 The following drawings will be revised as part of the Plant Design Change package (PDC 86-75) but are not included herein.

Dwg. I. D.

FSAR FIGURE TITLE M249 3.8-1 P&ID S14 System M-lF-213 3.8-2 SLC System Process Diagram G

PDC Bi-75, Rev. 0 Safeta Evaluat(on. Rev. 0 Attactment 1 N

A !3 Sheet 2 :.IG t

FHPS-FSAR 3.8 STANDBY LIQUID CONTRot SYSTEM 3.8.1 safety objective The safety objective of the Standby Liquid Control System (Stes) is to provide a backup method, which is independent of the coctrol rods, to maintain the reactor suberitical as the nuclear rystem cools in the event that not enough of the control rods can be inser.te d to counteract the positive reactivity effects of a colder moderator.

3.8.2 safety Design Basis 1.

Backup capability for reactivity control shall be' provided, independent of normal reactivity control provisions in the nuclear reactor, to be able to shut down the reactor if the normal control ever becomes inoperative.

2.

The backup system shall have the capacity for controlling the state rated operating

,,7,*'~

re, activity difference between the steady condition of the reactor with voids and the cold shutdown f.-

condition, including shutdown margin, to assure complete shutdown from the most reactive condition, at any time in tie core life.

time required for actuation and effectiveness of the backup h

The 3.

rate of control shall be consistent with the nuclear reactivity change predicted between rated operating and, cold shutdown A fast scram of the reactor or ope raticcial control conditions, of fast reactivity transients is not specified to be accomplished by this system.

4.

Me ans shall be provided by which the ' functional performance capability of the backup control system components can be verified periodically under conditions approaching actual use re quirements. A substitute solution, rather than the actual neutron absorber solution, may be injected into the reactor to of the Re dmdant. Control test.the operation of all components system.

absorber shall be dispersed within the reactor core 5.

The neutron in sufficient quantity to provide a reasonable margin for imperfect mixing or leakage.

The system shall be reliable to a degree consistest with its role 6.

intentional or as a special safety system; the possibility of

~

accidental, shutdown of the reactor by this system shall be minimize d. '

3.8.3 Description The piping and instrumentation for the SLCS is shovn on Figure 3.8-1.

Figure 3.6-2 is a process diagram for the system.

The SLCS is manually initiated from the main control room to pu p a boren neutron reactor if the ope rator believes the absorber solution into the cannot be shut down or kept shut down with the control rods.

reactor

~"--~

- " - ~ ~ ^ -

PDC 86-75, Rev. O Safety Evaluatipn, Rev. O Attachment I L

Sheet J of16 N3 /

PNPS-TSAR t.,

rods is expected to always assure However, insertion of control should it be required. The boron shutdown of the reactor nuclear fission promptcbsorbs thermal neutrons and thereby terminates the d

['

chain reaction in the uranium fuel.

The SLCS is needed only in the improbable event that not enough control rods can be inserted in the reactor core to accomplish s g shutdown and cooldovn in the normal manner. The SLCS therefore is *J y

a at a steady rate within the Y G j

sited only to shut the reactor downthe shutdown Cooling systems, and keep the reactor 4

b, o capacity of e

4 going critical again as it cools, boren solution tank, the test water t'ank, the two positive 4

and associated local 1 f

the displacement pumps, the two explosive valves, controls are mounted in the Rea gf 5

a' The liquid is piped into the reactor vessel and Mr valves and g g primary containment.the bottom of the core shroud so that it mises with j

x discharged near See section 3.3, Reactor g

the cooling water rising through the core.and Seepon 4.2, Reactor vessel

~

Internals Mechanical Designs g

gwh Ql f f vesself and Appurtenances Mechanical Design.

l aisodiumpentaborate The specified neutron absorber solution is h

t, 1tispreparedbyp

-y etn P -

n v:st :u

. nr ::: M ri; -.d in demineralized water. An air sparger is.

3 4 7. $ solution.4 To ' prevent system plugging, the

' f' ~

provided in the tank for mixing. tank outlet is raised above the bottom of th k

ter.8 a strainer.

finiti:1 core At all times when it is possible to make the reactor' at least dp gal of du, critical, the SLCS shall be able to deliver equivalent into the solution or 0 01.

( 2b perce3 3 sodium pensaborate _

- m. m,,ovu a w

.c 7ccerpn ecr ey ste in stan liqui control tank and lling with reactor.

east th low le 1 alarm olume.

he lg sol ion is design oncentra on -at th low level larm po

and, I

demin alized v ar to a a

be di ed up o the erflow vel vol to au v_ Jo#

{ *b va ai 1 ss s r to low the si_t _A_t(en tamperatur # -

g, M.

-v y,,

g

== w = ion temperature of the specified solution is urr so the equipment containing the solution is installed in a room in which the I

.j saturat The u

' ' ' " ^

^^^^n h

to be controlled '

8 air temperature is tank, and a An electric innersion heater in the n' ' h; d : :+* MI f) - 2

~

' ^-- 3.

int:f r 'l:

contro11ere" f;.,g:f i:

nr = y is esso used to elevate the y

temperature w

- - +

• : - n..,......r s. assure that the boron dissolves when first added to p

and or a

/ temperature High or low temperature, high or low liquid level, gTEI the water.

shorted heater causes an alarm in the control room,k a % aasa h

's sited to inject the solution into Each positive displacement pur

^-

C; solution level in the the reactor in 50 to 125 min, The pump and system design tank, at all reactor operating pressures.The two relief valves are set to exceed the pressure is 1,500 psig.

suf ficient margin to avoid valve operating pressure by a flooded The relief valves are installed with the discharge reactor le aka ge.

3.5-2

PDC 86-75, Rev. U Safety Evaluation. Rev.0 Attachinent 1 GENERAL ELECTRIC CO.

Sheet 4 of /6 d/3/

Nuclear Energy Business Operations ENGINEERING CALCUL4.lON SHEET DATE NJVSE A _

MEET or By SUBJECL d<>lg d,$ h sh 6 -R.1 9%Jg i

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~ - -- --- _-____ __._______ _______

PDC 86-75, Rev. 0 Safety Evaluation. Rev. 0 Attachrnent 1 Q SheetSoff6 M3 I PNP 5-FSAR to prevent evaporation and precipitation within the valve.

To II prevent bypass flow fr'om one pump in case of reitef valve failure tn' I

the Itne from the other pump, a check valve is installed downstream of each relief valve line in each pump discharge 1tne.

i The two explosive actuated injectiori valves provide high assurance of l

l opening when needed and ensure that boron will not leak into the reactor even when the pumps are t>etng tested.

The valves have a firing reliability in excess of 99.99 percent.

Each explosive va.1ve is closed by a plug in the inlet chamber.

The plug is circumscribed with a deep groove so the end will readily shear off when pushed. by the valve plunger.

This action opens the inlet hole through the plug.

The sheared end is pushed out of the way in the chamber, and is shaped so it will not block the ports after release.

The shearing, plunger is actuated by an explosive charge with dual ignition primers, inserted in the side chamber of the valve.

Ignition circuit continuity is monitored by a trickle current, and an Indicator alare occurs in the control. room if either ci,rcutt opens.

lights show' which channel primer circuit opened.

To service a valve.

firing, a 6 in length pipe - (spool piece) must be removed after tamediately upstream of the valve to gain access to the shear plug.

L M

The SLCS is actuated by a three pos'ttton keylock switch on the C'

control room console.

This assures that switching from the "off" position is a deliberate act.

Switching to either side starts one

.~

injection pump. opens an explostwe valve, and closes the Reactor Cleanup System isolation valves to prevent loss or dilution of the

,d f,.

.tr boron.

A green light in the control room indicates that power is available to the pump motor contactor, but that the contactor is open (pump not J i

running).

A red light indicates that the contactor is closed (pvap g

8 running).

Ltquid flow is confirmed by a decrease in reactt"vl'ty, storage tank i hP

^

C drawdown and pump running indication. A red light beside the Leylock e$

j switch turns on when valve 1101-1 downstream of the explosive valves If the pump lights or explosive valve light Indicates that l

ts open.

the Itquid may not be flowing, the operator can tamediately turn the J

keylock switch to the other side; this switch actuates the alternate

-r Crossptping and check valves assure a flow path through equipment.

~

either pump and either esplostve valve.

The chosen pump will start f

.3 even though its local switch at the pump 15 in the "Stop* position j f' J

.for test or maintenance.

Pump discharge pressure indication is, also y

provided in the control room.

{

Equipment drains and tank overflows are piped not to the Waste Systea J

g but to separate containers (such as 55 gal drumshe-t -w b :

=

' thi Q to prevent any trace of boron froa

^

inadvertently reaching the reactor.

t%cu erws (e

44. m.w 4 wL A*o t.u

.4 Wanh, ee-esa -

Ins trumenta tion is provided locally at the standby liquid control tank and consists of solution temperature indication and control.

PDC 86-75, Rev. 0 y

Safety Evaluatipn, Rev. 0 yd Sheet 6 of /6 (_,7 / 3 Attacht.1ent 1 PhPS.FSAR 7

5 f'

tank I' vel, and heater status.

Instrumentation and control logic is presented on Figure 3.8-4.

c 3.8.4 Safety Evaluation The SLCS, although not necessary for plant operation, is required to e

f be operable when the reattor is in other than cold condition.

3 f Despite this precaution, the system is expected never to be needed for plant safety because of the, large number of Independent control Ig rods available to shut down the reactor.

To further assure this 88 availability, two sets of the components required to actuate the

{ pumps and explosive valves are provided in parallel redundancy.

%q $

w,

r The system is designed to bring the reactor from rated power to ak 4Y e

The reactivity compensation g - g l d' d cold shutdown at any time in core life.

reduce reactor power from rated to zero and allow b cooling the Nuclear System to a cold shutdown condition with the f *, y g

provided will l

~~f E

control rods remaining withdrawn in the rated pcwer pattern.

It v t

.g includes the reactivity gains due to complete decay of the menonl i

'f inventory.. !t also includes the ' positive reactivity efff ets from 4

M eltetnating steam volds, changing water density from hot to cold,b 4

l reduced Doppler effect in. urantum, reduction of neutron leakage from 2..

')d I

g *8 }

bolling to cold, and decreasing control rod worth as the nederator '

Tne. spectfled minimum final concentration of boron in theU d

-4 2

reactor core provides a reactivity worth of approutsately -0.12 A kh C,

cools.

.e#[4 g plus a margin of -0.05 A k for calculational uncertainties anP g

8 l

e.

h +,~, j assures a substantial shutdown margin.

As

'7ao 7p 3j U

The spectfled minimum ave ge concentrate f natural boron in theT 4 reactor, to p tovide the s ecified shutdown,on omargin af ter operation of 3, J JC ij

-- 9..

c :. The nintsum quantity w -f g$

the SLCS, Is EP ppa,L odlus pentaborate e injected into the reactor is calculated 'I J

ase on the required ppm average concentration in the reactor f

6

-I coolant, and the quant y of reactor coolant in the reactor vesset

'M

-4

in 1-and rectreulation loops @

7 -"for taperfect utning,

~^

is increased by 25 percent to allow D

result leakage, and volume in other small piping connected to the reactor.

L--+.Cooldown of the Nuclear System will take several hours as a minimum, to remove the thermal energy stored in the reactor, cooling water.

and associated equipment and to remove sost of the radioactive decey heat.

The controlled limit for the reactor vessel cooldown is 100*F/hr, and normal operating temperature is about 550*F. Usually, shutting down the plant with the main condenser and various shutdown

cooling systems will take 10 to 24 hr before the reactor vessel is opened, and such longer to reach room temperature 00*F) which is the condition of mastnum reattivity and therefore, the condition which requires the maximum boron concthtration,

t. h The tem injection rate is.itmited to the range of 39 to 79 gal / min.

The lower rate assures that the boron gets into the reactor in about 1 1/2 hr, considerably quicker than the cooldown rate. The upper a*e % " d1gt a h. pud L. en&3A be.w

, 54S g, ')

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PDC 86-75, Rei. D.

Safety Evaluati Rev. o l

Atu ckwnt 1 -

f Sheet 7 of g D2 /

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PMS-F5AR injection rate assures that there is sufficient string so the toron does not recirculate through the core in uneven concentrattoes

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limit i

uhich could possibly cause the nuclear power to rise and fall

(

cyc11cally.

The systes piptni The SLCS:is designed as a Class I seismic system.and ecutp USAS 831.1.0 Section I and Appendix A.

1 the test tank is designed as Class !!.

i The SLCS is required to be operable in the event of a station po j

failure, The.puuss J*

standby ac power supply in the absence of normal power.

i and valves are powered and controlled from separate buses afd circuits so that a single failure will not prevent.. system operatter.

The essential instruments and lights are powered from tee E

I 120 V ac instrument power supply.

tog 5 C

l 9 g The SLCS and pumps have sufficient pressure margin, up to the systeminjection lato reitef valve setting of 1,400 psig, to assure solutto pstg in the the reactor above the normal -pressure of aboutThe nuclear system relief and sa

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begin to retteve pressure above about 1,100 pstg; therefore, the' SLCS~

bottom of the reactor.

E Mglsplacement pumps cannot overpressurize the Nuc at

- f positive a.

• C i d u *, u Ato provide 1 concentration l of amhumf boron in 4

s J

reactorkof 700 ppm The shutdown marcin from this contentrationPilg

'$ 7' The ' system 1ts the

.I can be found ' in Appendix Q.

The analysts and models for the reload core are described in the GE Standard Application for Reactor Fuel."'

" 4 fe.r i

) f 3.5.5. Inspection and TestingOperational tes j

avoid inadvertently injecting boron into'the reactorpe ng ine [et re$

i tts /

f f'E 7 te ir ate y

n the valves to and from the solution tank closed and theop 5 u 4

ch.

j three valves (two locked closed)the test tank can be rectreviated by e

in d.$

the demineralized water Me Y[

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e unctional testing of the injection portion of the system 15' ec w accomplished bf closing the locked open valve from'the solution tad,

" opening the locked glosed valve from the test tank, and actuattag the room to either the A or S circult.

keylott switch in the control in that This starts the pump and blows open the injection valveThe ctreutt.

system is operating.

l

4 PDC 86-75, Rev. 0 Safety Evaluat; n, Rev. 0 pgpg,7 g Sheet 6 of 16 2/3/

local locked open valve to the reactor in the leatage through the injection valves can be detected at gy closing a a test connection in the line between the containment isolation ch containment.

8 (Position indicator lights in the control room Indicate that

~

valve is closed for tests, or open and ready for A

valves.

Leakage from the reactor through the first check valve the local same test connection whenever the

.-j operation,)

can be detected by opening the v4 reactor is pressurized.

functional tests, the injection valves and explosive j

to their normal After the be replaced and all valves returned Charges must positions, as indicated on Figure 3.81.

4 s

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contains demineralized water for about three sin of

.D. mineralized water from the makeup or condensate The test tank pump operation. storage system is available at 30 gal / min for ref t111ng or flu e4

_4m the system.

~

Should the boron solution ever be injected into the reactor, either <

intentionally or inadvertently,l keep the reactor subcritical, the +

p reactivity controls wil system by flushing for j l.s normal is removed from the Reactor Coolant

1. D gross dilution followed by operation. of the Reactor Cleanup System.

~1 4 J' -

boron There is practically no effect on reactor operaf tons when the r

concentration has been reduced below approximately 50 ppa..

The concentration of the sodium pentaboratein the solution' tank is )

determined by chemical analysis periodically. % e.sickwd d 4k.k The gas pressure tr. the two accumulators is measured periodically to A pressure gage and portable nitrogen supply are detect leakage.

required to te;t and recharge the accumulators.

D-3.8.6 Asts45W C *Q'm U,A tate.

se' cpt. '5b.t.t, h hm 6hr ed b hate. hht the %14.

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Revtston 6 - July 1986 3.8-6

PDC 86 '5. Rev. 0 Safety Ivaluatjon, Rev. O Attachnen: 1 PNPS-fSAR Sheet Cf of f 6 L '213

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Current Operational kuclear Safety Requirements 3.8.7 The current limiting condition for operation, surveillance requirements, and their bases are contained in the Technical Specifications referenced in Appendix 8.

t 3.8.8 References 1.

MEDE-24011-P-A.

General Electrical Standard App 11 cat 5on for Reactor Fuel, applicable revision.

P S tand by 1. k uic] G o h ol S p e Gabl Ca pudy j

l E gu iva la nc.c Raro t ) Gene % l E le.ch ic, PAcJ l /z9 in v

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Revision E. A ly 1986 3.87 l

PDC 86-75 Rev. 0 Safety Evaluatjon. Rev. C Sheet /foil(s ( ~ A I3 l 1

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I FIGURE 3.8-3 S ATUR ATION TEMPERATURE OF SODIUM PENTABORATE SOLUTION PILGRIM NUCLEAR POWER STATION FIN AL S AFETY ANALY. SIS REPORT

4 FX 86-75. Rev. 0 Safety Evaluation, Rev. 0

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SODIUM PENTABORATE SOLUTION VOLUME 1

CONCENTRATION REQUIREMENTS i

PILGRIM NUCLEAf1 POWER STATION FIN AL S AFETY ANALYSIS REPORT f

b FDC 86-75, Rev. 0 GENERAL ELECTRIC C0'.

Safety EvaluaQon. Rev. O Nuclear Energy Business Operstoons L

ENGINEERING C ALCULATION SHEET Sheet [6 of 16 2 Of

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..7 PROPOSAL ANTICIPATED TRAE!ENTS WITmUT SCRAM-RECINTION w ngp (ATWS.RPT) 1 PILGRIM 416-4208-HK1 March 1979

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GENERAL ELECTRIC C0WANY 7

o NUCLEAR ENERGY PROJECTS O!VI5;0N BWR SERVICES DEPARTMDri h

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PROPRIETARY INFORMATI'ON NOTICE l

This proposs$' document contains proprietary i '>reation of the General E1xtric Company and is furnis in confidence solely for use in considering the meri of the pmposal and

' J i

3 for nc other direct or indirect By accepting this docu-ardt feca General Electric, t recipient agrees

(1) to use this document and the in tion it contains exclusively for I

the above stated pu e.

(2) to avoid publication or other unrestricted d osure of this document or the infomation it contains. "d to make no copies of any part hereof without the prf,o r itten permission of General Electric, and (4) to return co th

.1, ocumer.t when it is no longer needed for the purpose for ich furnished or upon ttf request of General Electric.

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The MRC has issued the requirement for installation of an Anticipated Transients Without Scram-Recirculation Pump Trip (AT&tFT) modification for all operating plants within two years. Further, they have indicated licensing acceptance of the General Electric Company *Monticello ATWS-RPT" design for all operating plants. Finally, they have requested the utility's taplementation Schedule by April 9.1979.

The General Electric Company proposes to furnish the plantimprovement program described in Section 4.5 of Topical Report IE00-25016. " Evaluation of Anticipated Transients Without Scram for the Mmticello Nucleer Generating Plant", to meet this requirement.

2.

Design Objectives l

The intent of the ATWS-RPT design is to meet the following objectives:

l Shut down all recirculation pump motor generator sets with redundant N

1.

logic from the following inputs:

tn O

h%.iettht4va

/4 b) Reactor Vessel High Pressure c) heactor low Water Level The systeci is to be diverse from the Reactor Protection System (RPS).

2.

4 3.

The system is to be testable in service.

4 The system is to be designed so that as much as possible no c.

single component fativre can prevent the tripping of both O

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recirculation pumps.

The hartiware should be high quality and environmentally qualifted.

S.

6.

The system's performance characteristics are as follows:

a) Logic 6elay for trip. including dynamic response of the sensors logic, action of

< 0 53 seconds

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the breakers and collapse of the generator field b) Lw level delay timer (to be confirwd by plant unique analysis!

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3.

Jus ti fication Economic justification for the utility to accept the proposal results from successful realization of the objectives discussed in Section 2.

Further, the ATWS-RPT system improvement package is a standard, generic design, with appropriate licensing report.

4.

Hardware Description 1.

Transmitters The reactor pressure and reactor unter level will be sensed by analog transmitters. The pressure and level transmitters used are designed and manufactured to Geners) Electric specifications.

Calibration and shutoff valves, 3-way balancing annifolds and fittings are provided for field assembly. See Figures (1) and (2) to for typical sketches of a Pressure Transmitter and Differential Pressure Transmitter Valve subassembly.

e t1 2.

Trip Units The trip ur.it used to generate the trip signal is designed and j

a manufactured to General Electric specifications. Features of this system are:

1.

Trip units are functionally tested or calibrated in piece f

'by~means of a portable readout assembly.

~

2.

Trip unit calibration current is a controlled ramp of 1 s/sec.

4 J

c 3.

The output of the trip unit is a voltage to a Class 1E relay.

j c-4.

The master trip unit has a display meter which monitors the transmitter cwrent for gross failures as well as failures o

in the current to voltage and filtering sections of the master trip unit.

)

5.

Trip unit "out-of-card file annunciation" is provided.

6.

The trip unit provides long-ters trip point stability.

7.

The calibrator is capable of introducing a step current for transient testing of the logic.

8.

The readout anesbly is a portable unit which any be used with all calibration units. This provides for portability of the secondary standard from card file to card file to measure the value of the trip point in various cabinets.

9.

Card extenders, a bench test unit, and calibration units are supplied to facilitate maintename as required.

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10. The trip relay is a general purpose relay with four fom I

C contacts which is rated for the seient environment at its mounting location.

l 4

i 3.

Cabinets General Electric designed cabinets which house the trip unit card file. DC power supplies, and trip relays are illustrated in Figwes 3 and 4.

The large cabinet (Fipwre 3) has space to house up to three trip unit card files thereby pemitting use of the cabinet for future installation of the Analog Trip System or other plant improvements. The small cabinet (Figure 4) has space for only one card file. Each card file has space to acce te twelve trip units. Additional card files can be factory supplied for large c:abinets per the price tabulation in the quotation letter, e

or added later by field installation.

c0 The 25 volt DC power supply used is a ferro resonant type which to is recognized for its high reliability. Two complete power o

supplies are mounted in each trip unit cabinet and connected in parallel. With this arrangement a single power supply can fail without affecting the trip function. The power supplies in the large cabinet have sufficient capacity to supply power for the

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maximum n' umber of trip units (36) that may eventually be housed 4

in a large cabinet. The power supplies for the small cabinet o

have sufftetent power to supply the 12 trip unit capacity of that e

cabinet.

o The cabinets any be lectted in the control room, auxiliary equipment room (location of existing logic cabinets), or in the vicinity of the local instrument rock in the reactor building. The cabinets must be located in an area where sexima adient temperature will be between 40-145'F. The electrical components within the cabinet are qualified to operate up to 150'F at 995 relative hweidity.

The cabinets are seismically qualified per IEEb 344, 1975, to the Safe Shutdown Earthquake ($5E) acceleration response spectra shown on Figure 5.

The large cabinets are floor mounted; the small cabinets are wall mounted.

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Fressare Trar.statter l

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r mtimmw CONDUlf 0 0CK0'JTs FOR CAltE ENTRY TOP & BOTT0ft 0

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4., Inverter _

An additional DC to AC invertar is required in each ECCS division to sinable the ATWS-RPT trips to have their backup power source The inverter is to be located and from the station batteries.

installed by the purchaser external to the trip cabinets.

Systee Description and Application _

5.

Since normal ceram is assumed to be unavailable for reducing the reactor is power, and since the transient event is one in which power reduction necessary, the ATWS-RPT systes provides another method of reducing pow The trip of both recirculation for the first 15 seconds of the event.

This pumps causes a quick reduction in core flow thus reducing the power.

C fuel quick power reduction brings the reactor pressure, neutron flux and l

c surface heat flux down in time to acceptably limit the peak pressure, m

clad oxidation and peak fuel enthalpy.

a i

The initiating variables, rtactor pressure and reactor water level, are The transsiitters are mounted on reactor sensed by analog transmitters.

d pressure instrumentation racks and connected through calibration an isolation manifolds to existing instrument sensor piping to the reactor Manifold and calibration valves are supplied to pressurf vessel.

facilitste installation of the transmitter to existing process piping.

Figure 6 is a schematic diagram showing the arrangement of the tr C

The trip units are mounted in cabinets o

and the divisional separation.

The trip units provide the which conform to ECCS separation criteria.

inputs to the logic which trips both Recirculation Pump Motor-Generato Sets' field breakers, thus shutting down the recirculation pumps.

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3. AtCW.; PUMP emtaKtA TRIP CC2L

v. A R.t sc.Totot A.- ika, Glas Figure 6

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-,. 6.

General Electric P, responsibilities 1.

Prepare cabinet outline drawing

• 2.

Design. build and wirt check the trip unit cabinets.

3.

Provide the hardware listed in Table 1.

4.

Prepare Elementary Diagrams showing devices added with location.

]

termination numbers and interconnection per the options in the J

quotation letter.

k 5.

Provide instruction manuals.

6.

Provide an ATWS-RPT system specification.

3 7.

Provide transmitter and related valve assen61y drawings.

1 8.

Provide an installation specification with installation, calibration,

{

and startup instructions.

3 9.

Provide an ATWS-RPT system description suitable for licensing submittal.

1

• Cabinet design will vary with the cabinet option purchased.

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7.

Purchaser Responsibilities 1.

Physically locate and fasten the cabinets securely to the floor or tell. as required (if cabinets are supplied).

Run raceways from the trip units to the transmitter racks, as 2.

required.

Mount, asseable and pipe the transmitter / valve subasssably to 3.

the process piping. Leak test the new transmitter installation.

Make electrical connections between the transmitter to the j

4.

trip unit and the trip unit relay to trip coils and vent solenoids.

l i

Run power cables from the existing logic cabinets to each trip l

r 5.

unit cabinet. Existing spare cables that have been made available l

l from other plant modifications may be used.

Run annunciator loop cables from the control room to each trip I

6.

p unit cabinet.

Revise the plant unique P&lD's. FCD's. Elementaries, IF s. etc.

7.

C 8.

Provide labor f or installation.

Provide all licensing for the application of the haeare into 9.

the power plant.

10. Checkout and placing in service.

j Verify the seismic spectrum the cabinets were tested to (Figure 5) f 11.

o is greater than the floor / wall response for the mounted location of the cabinets.

Perform seismic analysis as may be requireo to verify the seismic cm 12.

conditions at the locations where the transmitters will be installed O

do not exceed the allowable equipment design limits.

13. Verify the environmental conditions at the locations where the equipment will be installed do not exceed the allowable equipment desiipn limits.

Incorporate new surveillance testing requi ements for the analog 14 trip.mits and transmitters into the technt 41 specifications.

15. Provide detail test and calibration procedures based upon GE supplied instructions.

Locate and install the DC to AC inverters external to the trip 16.

cabinets.

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. Major Co.penents included in this Proposal 8.

Table 1 is a list of anjor hardmare items which will be supplied by the General Electric Company.

TABLE 1 Major Hardware Items Supplied Description p s,

item Quantitt p

p>

+I Lew) Transmitter. Manifold Valve & Fittings 1

Pressure Transmitter. Valve and Fittings 2

4 Trip Unit With Analog Output 3

8 p

4

,AVp2-Trip and Alars Relays g

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Manual Initiation Pushbutiom 6

4 O

7 As Required Light & Holder-8 As Required Fuse 1

Trip Coil for MG Set Field Breakers 9

5 AC/DC Power Supply 10*

2 Readout Assembly 11 3

Card Extender 12 3

Sench Test Unit j

13 1

C 14 2

Card Flie C.

15 2

Calibrator 16*

2 Cabinet 17 2

SC/AC Inverter Isolation valve 18 4

lcm f'n Y 0* Y'N gp L

TJs LI

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• Capacity will depend on cabinet option purchased.

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Quality Assurance 1.

Equipment and Services shall be provided in accordance with the General Electric 8WR quality assurance provam as described in Topical Report NE00-13209-04A.

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2.

The previsions of 10 CFR Part 21 apply.

3.

A Product Quality Certification (PQC) shall be provided by General Electric as the primary quality asswrance record for Equipment classified as "important to safety".

4.

SWR owner access requirements for audits and/or witness of inspection points for supplied Equipment and Services shall be arranged upon request at autually agreeable terms.

5.

All General Electric Nuclear Energy Divisions' work at the BWR M

owner's plant site shall be under the cognizance of the BWR owner o

quality assurance program.

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10. Licensing g

A description of the ATWS-RPT plant isiprovement retrofit package suitable for licensing submittal will be supplied. Additional detailed assistance can be provided on a consulting basis at our commercial rates.

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