Text

Rev 2 to Regulatory Compliance Instruction 02.3, Tech Spec Interpretation Request,Processing & Maint
	ML20090K300
Person / Time
Site:	Brunswick
Issue date:	03/06/1986
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20090J504	List: ML20090K300; ML20090K431; ML20090K435;
References
	FOIA-91-170 02.3, 2.3, NUDOCS 8909280209
	Download: ML20090K300 (75)
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CAROLINA POWER & LIGHT COMPANY BRUNSVICK STEAM ELECTRIC P1AVT PNIT 0 PROCEDURE TYPE:

RIGULATORY COMPLIANCE INSTRUCTION (RCI)

NLHBER:

02.3 PROCEDURE TITLE:

TECHNICAL SPECITICATION INTEFJRETAT*DN REOUEST.

PROCESSINC. AND MAINTENANCE REVISION 2

/

MJ4 <h

' ATE :

O [,, d APPROPJ BY:

GeneyaJ Manarer/

/

Director - Regulatory Compliance 0b'OM

\\q 0 0-RCI-02.3 Rev. 2 Pate 1 of 7

i LIST or ErrtcTIVE Pacts (trP)

L RCI-02.3 Paees Revision r

1-7 2

1 L

r b

t 0-RCI-02.3 Rev. 2 Page ? of 7

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

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1.0 INTRODUCTION

1.1 Purpose i

To provide the method by which formal interpretation of technical l

specification (s) shall be requested, processed, and maintained.

l 1.2 Scope 1.2.1 A technical specification interpretation is not intended to revise er substitute fer the existing technical specifications, but to clarify the existing technical i.

specifications. As such, a technical specification interpretation should be issued only if the existing technical specification:

i 1.2.1.1 Is not consistent with the physical or intended design of the plant.

1.2.1.2 Does not adequately demonstrate operability or is not suf ficiently conservative.

1.2.1.3 Contains typographical or administrative errors that present operability or surveillance t

concerns.

1.2.1.4 is determined necessary by the Plant Nuclear Safety Coenittee (PNSC).

Consequently, a technical specification interpretation is issued only if a technical specification change is l

required, unless determined otherwise by the Director -

Regulatory Compliance.

1.3 References 1.3.1 RCI-02.1 Request for Operating License Changes j

1.3.2 RCI-02.2, Operating License, including Technical

}

Specifications. Amendment Issuance, and Ieplementation j'

1.4 Respensibility 1.4.1 The Director - Regulatory compliance shall enture that this instruction is implemented to provide accurate.

documented, and retrievable information, 1.4.2 Personnel shall follow this instruction to obtain formal interpretation of technical specification requirements (s).

l; 0-RC1-02.3 Rev. 2 Page 3 of 7

b 2.0 INSTRUCTION 2.1 Inquiries requiring interpretation of technical specification purpese, intent, and/or meaning shall be made utilizing Technical Specification Interpretation Request Forn.

2.2 The originator shall complete Blocks 2 through 7 of the Technical Specification Interpretation Request Form and shall forward the interpretation request to his supervisor for concurrence.

2.3 The ertrinater's supervirer and/o- -resra

'dd-aa'a-

fedfa**a concurrence with the request by signing Block 8. and shall forward the request to the Regulatory Compliance Unit for resolution.

2.4 Regulatory compliance shall review the request for interpretation and shall 2.4.1 Assign a technical specification interpretation serial number into Block 1 of Attachment 1, and log the interpretation.

2.4.2 Return the request to the originator, or his supervisor /

manager / director, if a formal interpretation is not required (i.e., sufficient docusentation exists to provide a response to the reauest), with a response citing the reason for interpretation reiection.

2.5 Rerulatory compliance shall investigate valid requests, provide an interpretation in Block 9 of Attachment 1. and sign Block 11.

The need for a technical specification chanee vill be indleated in Block 10, 2.6 The Director - Regulatory Compliance, or his designee, shall reviev the interpretation, and shall indicate his concurrence on Block 12 of Attachment 1.

4 2.7 Rerulatory Compliance shall present the completed Technical Specification Interpretation Request Form to PNSC for approval on-Block 13 of Attachment 1.

2.6 Following FNSC approval, the completed Attacheent I shall be routed to Document Control for distribution in accordance with applicable Records Management Instructions (i.e., to holders of plant controlled copies of technical specifications, lucluding the historical files).

2.9 t'pon receipt of Attachment 1, holders of plant controlled copies of technical specifications shall dnsert the completed form into the

" Interpretations" section of the technical specification binder, or a separate " Interpretations" binder, and shall annotate the affected technical' specification page to indicate existence of the interpre-tation in their binder.

0-RCI-02.3 Rev. 2 Paec 4 of 7 I

4

2.10 If a technic ti specification change is required by the interpreta-tion, Regulatory Compliance shall initiate that change in accordance with RCI-02.1.

2.11 At least annually, Regulatory Compliance shall review the

" Interpretations" section of the technical specifications, shall void any interpretations which are no longer applicable or have been renedied as a result of Technical Specification Amendment Issuance, and shall provide this information (in the form of a listing) to Document Control for issuance to holders of controlled copies of it;htical specificattens. including the historical filer.

2.12 Interpretations established by this procedure sey be canceled or deleted in concurrence with RCI-02.2 or at the direction of PNSC.

0-RCI-02.3 Rev. 2 Page 5 of 7

ATTACHMENT 1

~

TECHNICAL SPECIPICATION INTERPRETATION REQUEST FORM

1. Serial No.

2. Technical Specification

Reference:

3. Technical Specification Page No.:

4. Subjects

5. Unit (s) Affected:

B S EP--

6. Description of Request (concise detailed description of requested interpretation or problem ares):

7. Originator:

Date

8. Reviewed By:

Datet

9. Interpretation:

10) Technical Specification Change Required:

Yes No

11) Prepared By _

Date

12) Concurrences Date Director - Regulatory Compliance

13) Approved:

Date PNSC 0-RCI-02.3 Rev. 2 Page 6 of 7 i

i i

4 6

w--w.

1r.%=-.

i i

ATTACRMINT 2 I

TECHNICAL SPECIFICATION INTERPRETATION SERIAL LOG Serial No.

T/S Reference T$C Number 1 Date Initiated Date Approved ~

I s

0-RCI-02.3 Rev. 2 Page 7 of 7

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

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6. Descriptic of Requeet (cer.ctra detailed dcecriptien of requet.ted interpretatien or problem arts): b./sa v 5. ca vicE. w/A rric, t>v e q M L'5.T G tT C W 54E' E iOcc2E.'lt 7 S/mt,rv-uk t ca vm a m se, Ee z na s 'Te oie ac. S pe r me.mc eJ f

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9. !cterpretatient m%w M rc er I>?.cViccs es s C rrewiccN c rJrs t 01 S ERWcE v.'A%L PUM P5 A ND Recus art A C.TIONS.

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FNSC RECEIV.ED MAY1 0 H 3 EM NQO' CObIEOL' C 401-02. 3 Rev. 2 Fece 6 ef 7 e

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te w"at I are C tersice *6t:r t)ttons sho!! te, CFCAACLE with at leasts a.

Four OPERABLE nuclea? strvice water pumps for,the site cao4Lle of supplying service water to-the nuclear heaeers; and in adottion, b.

Two CFERABLE Unit l(ti conventional service water

pumps, each powered from a elfferent division and cacatie of sucolying both the Unit 1(2) service water nwelear heacer anc conventional header.

I Reference informations Divisico Assigr#ents of Fumes Divassen I (Et, E31 Division !! (E2,E4) la NSW 12 NSW Ea NSW BB NSW 15 C5W' 1A C5W En C5W tC C5W G C C 5 s-ES C5W A c.

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Wit- : Sly tnree CFEAAELE nuclear service water pumes fcr tne site, restore at least four nuclear service wate-osmos to CFERABLE status nitnin 7 days or ce i e;-

at le.st MCT SsuTDCWN withan tre newt 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months anc C*LO SwuTCCwN wi tni n the f ollowing E',

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With only c ne CAC ACE'.E Uni t itE) conver.tional service

water pues, operation may continue proviced tha' two nuclear t.rvice water purcs art CPCRAPLE on the unitt restore at Itast two convet.tional service water tw*ps pc erec fecm cifferent olvisions to CPERA9LE status witnin 7 cays or be in WCT SHUTDCWN within the mest 12 howrs anc COLD SHUTDCLN withnn the following 26 hewes.

6 With only two CPE4ABLE Unit 112) conventional service water numes gewerec from the same civiston, coeratten may centinue providee that two nuclear ser<tce

.ater e.+cs are CPEccPLE ce the unitt resteer at least toe c e n er.t a c na l soe v ic e wa t er swmcs pc erec fecm cifferent civissces to CFE:AELE status a. thin 7 ca,s or te in -CT S-U**C N witnan the nost 12

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ATTACEMINT 1 TECENICAL SPECITICATION INTEAPRI!ATICS RIQUEST TCLM f) Serial No.

81-09 (Rev. 1)

1) Technical Specification

Reference:

Table 3.3.3-2 Iten Id 3/4 3-34

2) Technical Specification Page No.:_

3)

Subject:

Core Spray Time Relay

4) Unit (s) Af fected:

BSEF--

Both

!) tescriptice of Request (concise de.tailtd desgri-tipo of requested

't"*

'C CI 058 J'3*J

ncicates that interpretatice er prcble: area):

the trip setpoint fer the E21-K16A, B time delay relay is 14 $t 516 sec.

This relay is actually a 5 sec. relay.

L'h a t is the correct value?

t t

6) v:,ginat:r:

- Gick 5/15/94 Tony Date:

Beb Poulk, Jr.

5/15/84

.,) Rev* eved By:

Date:

The correct value f er E21-K16A, B is 4.5 5 t $ 5.5.

The K16

e. )..,e

,etag r.

v;rks in series with the STR LA2,132 (2. A2, 2.B2) which is a 10 second ti=e relay te provide the 15 sec. delay fer core spray.

This is in accordance with plant desir. (?Ms79-125 and 79-166).

A T/S chance to reflect this has been initiated.

10) Technical Specificatien ChanSe Required:

Yesy (84TSB30) No 11)-Prepared By:

'd~

Date r[."

-. 1 ( + L Date Thd' 5"

12) Cencurrence:

L. rect:r - Ryyulat:ry Cenp;1ance

13) A;;::ved:

.0').S b )

Date 2 Y, t

P54;

-?.;;-: 1. 2 Fev. O y..

3 ci -

t 4

ATTAC1 DENT 1 TICKh'ICAL SPECITICATION Ih7ETJRITATION AIQt'EST TOR

1. Serial No.24 - 10 ke v, I.o

~

2. Technical Specifiestion ReferencetMihaVE 3. d. b.1 -

V

3. Technical Specification Page No.t_ N 4-4 -16, 4 -lk 4-16

4.

Subject:

P%urtfTe i wb e po f fW bkb

5. Cci t (s ) Af f e c t e d :

R$r?-

6. Description of Request interpretation er probles area)t(concise detailed descr(ptio,n o{ re ut tad

__ dc

- E M w a-F, n e k.N.vf c

w A m L g m nor m ; ra % 2. u... u. 6 4. %

fedsd 4., EO,b W.Do-aw ad dab -2kis9 feS<e e3 W & F<R ' '

JJepo - M u ar [ dron 24 5-) 5 ve.h a <e.eo ard and ok

' (b w Am 4Le m.

7. Orininator

)

Da t e t_ [, - I(o - 6 9 -

s. Reviewee 3 : III d/Y 7

4/tc /f 9 Date:

5.1:terpretatieU.:'

br Rhhb.

9 RECEIVt-D JUN 3 01989' MD DOCL CONTROU

10) Technical Spe fication Change Recluired:

Ye sFJTW 4- ' Nc_

11) Prepared By:

M b."LG R /4/27[F Date

12) Cencurrence:

Da t e G!

a.-e D rec]or - gr,ulatory Co pliance

13) Approved _

/N/

I (* e Date_

4 "ki-I k PNSC

- O C-R02-02.3 Rev. 2 Page 6 of 7

.e

_,_ _. W ma n '~

e w

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n.

r-J.

-we

The existing T/S curves have not been revised to -reflect NEDO-2 4161 and NEDO-24157, and are not considered sufficiently conservative.

A T/S change has been submitted and considers the latest revisions to 10CTR50, Appendices G and H.

Pending issuance of tha new amendment, this TSI provides conservative curves to be followed.

The attached curves have been calculated using the latest pertinent provisions of 10CFR50, Appendices G and H, ASTM E-185 and the trend equation curve of R2 1.!!, Rev.

2.

The attached curves have been adjusted to ensure complete conservatism with regards to the existing T/S curves.

Revision 1 to this TSI was issued with the hydrostatic and leak test curves calculated for less than or equal to 4 ef fective full power years (ETPY).

Both units 1 and 2 have surpassed 4 ErPY.

Revision 2 to this TSI provided hydrostatic test curves for less than or equal to 5 ETPY.

Additionally, revision 2 provided separate pressure / temperature curves for Unit 1 and 2 because Unit i has a smaller irradiated RTNDT than Unit 2 as calculated by the trend curve equation.

Revision 3 to this TSI was issued to revise the Unit 2 Hydrostatic and Leak Test curve.

The curve was applicable up to 4.75 ETPY vice the then existing curve which was good to 5 ETPY.

This revision allowed the hydro to be performed at lower temperatures.

Revision 4 to this TSI is being issued to implement the latest operating curves.

The Unit 1 and Unit 2 hydrostatic / leak test curves currently in revision 3 are no longer applicable in that Unit I has surpassed 5 ETPY and Unit 2 'has surpassed 4.75 EFPY.

The curves being imple=ented by this TSI have been adjusted to ensure complete conservatism with regards to the existing T/S curves.

Revision 5 to this TSI is being issued to implement the latest Unit 1 pressure / temperature curve for Normal Operation With Core critical (Figure 3.4.6.1-2).

This pressure-temperature curve has been revised using the latest NRC guidelines along with actual neutron flux / fluence data.

Revision 6 of this TSI is being issued to revise and clarify the 1 criticality curve for Unit 1 that was transmitted via revision 5.

The original criticaljty curve for Unit 1 ends at approximatsly 153 degrees on the x-axis and 460 psig on the y-axis (Ref. Figure 3.4.6.1-2 of Tech. Specs.).

In revision 5 of this TSI it was determined that the area below the above x-y coordinates was an (o

undefined area.

The lover, portion of the curve was therefore plotted using the latest NRC guidelines.

Regulatory Compliance has determined that the original Unit 1 curve should be drawn straight down from the above listed coordinates.

This revision f

revises the curve issued via revision 5 by removing the cross-hatched area and by adding the lower portion of the curve.

The

lir:it 'has been set at 170' r instead of 153' r to account for neutron embrittlen,ent.

This revision vill suffice for the-

/

interin until the-permanent Technical Specification change w

Request, which includes the cross-hatched area, is approved-.by the NRC.-

l

+

ATTACHMINT 1 i

ICLCl.EAILSATETY EVALUATION CHECKLIST Identification.and description of Item Being Evaluated

_TSI 84-10. Rev. 6 I

l circle one i

1.

Does this item represent a change to the facility.

ttdu--!**d 1-the T nR?

h Yet It: yes, describe the change and the FSAR section(s) t involved.

2.

Does this. item-represent a change to procedures as described in the FSAR?

Yes No If yes,. describe the change and the FSAR section(s) involved.

3.

Does this item represent a test or experiment nga described da. the TSAR?

Yes No If-'yes, describe the test or experi=ent.

4.-

~DoesIthis item require-a revislor. to the FSAR?-

-Yes' No-(If_yes, submit appropriate-change (s) per RCI-04.1, as applicable)~

5.

Does.the proposed-change, test or_ experiment require ~

a cha'nge to the technical specifications?

es*

No

.(If yes, submit appropriate change (s) per RCI-02.1)

Technical Specification change 88TSB04_has been submitted.

C-RCI-03.'1 Rev. 5 Page 11 of 14 '

. (Cont'd)

Circle one 6.

The change, test, or experiment shall be tested against the following criteria:

6.1 Will the probability of occurrence of any accident previously evaluated in TSAR (Chapter 15) be increased?

Yes No Uncert:in BASIS

See attached.

6.2 Will the consequences of any accident previously evaluated in the TSAR (Chapter 15) be increased? Yes No Uncertain BASIS

See attached.

63 Will the probability of occurreene of malfunction equipment important to safety previously evaluated in the TSAR be increased?

Yes

'o Uncertain BASIS

See attached.

6.4 Will the consequences of malfunction of epip=ent important to safety previously evaluated in the TSAR be increased?

Yes e Uncertain BASIS

See attached.

6.5 Will the probability of an accident or possibility for calfunction of equipment important to safety of a different type than already evaluated in TSAR l

be created?

Yes No Uncertain BASIS

See attached.

l

BASIS RIQUIRED - UTILIZE ADDITIONAL SHEETS AS NICESSARY C-RCI-03.1 Rev.5 Page 12 of 14

ATTACHMINT 1-(Cont?d)

Clrele one Willthetarfealspecificationsbereduced?in of safety as defined in the basis 6.6 to any techn YeshUncertain BASIS

See attached.

7.

If any response to Question 6 is yes, it is to be Yes assu=ed that the proposed change or test or experiment constitutes an unreviewed safety quistion within the ceaning of 10CTR50.59.. Based on this determination, does the subject change or test or experiment constitute an unreviewed safety question?

s 8.

Will the change, test, or experiment have a significant adverse effect on the environment?

Yes No 9.

-Will the change, test or experiment. raise a potential safety. concern on the unit to which it applies, gr...to the__pj;h e r unit ?

Yes No 10.

Dees the change, test, er experiment affect or bring abcut a-change to other plant procedures?

Yes No NOTE:

A ccpy of this safety Evaluation shall accompany the package through

review,

11. -Indicate the sections of the TSAR researched to confirm the deter =inatiens =ade in Ita=s 1, 2,

3, 4,

6, and 8 above.

5.3.2, Charter 15.

12.

Indicate the sections of the technical specifications researched to confir: the deter =inations, as applicable, made in Items 5 and 6 above.

3/4.4.6. Tierures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-3, and table 4.4.6.1.3-1 Prepared by:

dNL Date:

bMM

Title:

SR. Me c.k. $bbkTd PNSC Approval:

3 Dates (if-required).

BASIS REQUIRID - UTILIZE ADDITIONAL SHEETS AS NECESSARY 0-RCI-03.1 Rev. 5 Page 13 of 14 4

e

SATETY EVALUATION 6.1 The pressure-temperature curve being implemented b,v this Technical Specification Interpretation is based on the latest NRC guidelines relative to the RPV pressure-temperature curves (Revision 2 of Reg.

Guide 1.99, 10CTR Appendix G, and Appendix G of ASKE Section.III) along with actual neutron flux / fluence data.

Reactor coolant system temperature and pressure are currently utilized to comply with the requirements of TS Section 3/4.4.6 and have been evaluated to confirm that they are representative of the vessel shell temperature and vecer' --

r"~e The reacter ert: ant system amperatu:c, mecc.re c.

the recirculation pump suction, is actually 7 rer than that of the vessel shell during various phases of operat n (i.e., reactor startup, operation, and immediately following reactor shutdown) because of the effects of gamma heating of the reactor vessel.

Therefore, use of recirculation pump suction temperature is more conservative during these operational phases.

Since the coolant system data is representative of the vessel shell temperature, the probability of a pressure boundary failure vill remain the same and vill provide the sane limitations of the consequences of a pressure boundary failure.

It is therefore concluded that the probability of occurrence of any accident previously evaluated in chapter 15 of the FSAR will not be increased.

6.2 The accidents analyzed in Chapter 15 of the Updated FSAR are not af fected by the revised pressure-terperature curve.

The curve is designed to provide fracture protection for the reactor vessel coolant beu dary.

The consequences of a pressure boundary failure' are not impacted by this change.

Since the curve is based on the most current regulatory guidance and fluence data, it can be concluded that there is nct a significant increase in the consequences of an accident previously evaluated.

.f.3 The pressure-temperature curve being implemented by this TSI dces nct change any of the postulated accident scenarios or the accident initiators.

Additionally, this change does not adversely affect the operability of any safety related equipment.

Therefore, the probability of occurrence of malfunction of equipment important to safety will not be increased.

6.4 As stated above, this change does not alter any of the postulated accident scenarios or accident initiators.

Nor does it adversely af f ect the operability of any safety related equipment.

Therefore, the consequences of malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.

6.5 The accidents-analyzed in chapter 15 of the Updated FSAR are not affected by the revised pressure-temperature curve.

This curve is designed to provide fracture protection for the reactor coolant pressure boundary and does not create any new accident mode.

Accident modes for the reactor coolant pressure boundary, due to nonductile failure are well understood within the industry.

The pressure-temperature curve merely provides the protection mechanisms to preclude such a failure.

Therefore the probability of an accident er

possibility for talfunction of equipment important to safety of a different type than already evaluated in the TSAR vill not be created.

6.6 pressure-temperature curves are designed to provide a specific cargin of safety.

This margin is required to be at least as great as that specified in the ASHI Boiler and Pressure vessel Code,Section III, Appendix G, and Appendix G to 10CTR50.

The revised curve is based on the latest FRC guidelines (Regulatory Guide 1.99, Rev. 2). along with actual neutron flux / fluence data for the Brunswick Units.

Thus the curves provide a greater confidence level than the present curves.

It car thereferr be cer.02vded thet the cargin cf safety at defined in the basis to any technical specification vill not be reduced.

PRESSUPE-TEVPERATURE LIMITS REACTOR VE5SEL BEEP UNIT NO. t t

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ATTAC1 DENT 1 TECHNICAL SPECIFICATION INTERPRITATION RZQUEST FORN

8) Serial No.

T4-l9

-1) Technical Specification

Reference:

Table 3.3.1-1

2) Techef es? - F,"rf ff retie-Pt;* Fr.:

34 3*3

3) Subjects _ RPS-Turbine Control Valve Fast closure Control Oil Pressure Low Channels.

4) Unit (s) Affected:

BS EP-Unit 2 only

'S

5) Description of Request (coccise detailed description of requested interpretation or problem area):

T/S 3.3.1. Table 3.3.1-1 Item 10 requires 4 operable channels per trip system, however, the BSEP design has only 2 channels per trip system.

Bov many channels per trip systen*are actually required?-

6) Originator ). b 4 Dates _ /0/21/f/

7) Reviewed By:

)

s ec Dater /D

/

g f

a -

9) Interptr eation: -The correct number of required channels per trip systen is 2.

E. ASIS : The number in the T/S represents a typographical error introduced when the page was retyped for T/S change-request-involving digital to analog modifiestions. When the NRC issued the amendment (#97), the error was also incorporatac. The applicable submittal. dated 1-26-83 did not discuss a change to the. turbine control. valve switches nor was this-portion of the page " barred" (Cont Q w.h q,)

10) Technical Specification Change Required:

Tet_ V No 11)-Prepared By: O6 M4 Date-fo/.27/e<f

12) Concurrence:

/ W Date_

/0 /

f/

f Director - R,ptulatory Compliance

(

.+

13) A;preved:

i-.:St-Date se /2 9 [ W PNSt

/

C-R0!-C;.3 Rev. O Page 5 e 6

9)

Interpretation (Cont.)

- indica ting a change.

(The Safety Evaluation issued with the amendment did not-address. changing the number of channels por was this portion of the new amende.ent page " barred" indicating'a change). Thus the number of required channels involving the turbine control valve pressure switches is an error and-should be 2 instead of 4.

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

,c' i

n A77AC10ENT 1 TECHNICA1. SPECITICATION INTER.PRETATION REQUEST FORM

6) Serial'No.-

Y 4-20

1) Technical Specification

Reference:

4.1.4.2b.

3/' I'10

2) Technical Specification Page No.:

3)

Subject:

RSCS Operability

4) Unit (s) Affected:

BSEP 1 & 2

5) Description of Request (concise detailed description of' requested interpretation or problem area): T/S 4.1.4.2b requires demonstrating RSCS OPERABLI by attempting to select and move an out-of-sequence control rod in each of the other three rod groups as soon as.RSCS is autcoatically initiated when reducing THEFMAL PC'='IR.

Bovever, between 22% full power and 50% rod density, the RSCS has only two out-of-sequence rod groups. How can.this requirecent be 6)'hkN(tr:

b.

h%//=_

Date:

/ **

/ V --

/

~

Dates

/o 29 f/

7) Reviewed By:

~

O

(

'l

9) !sterpretation:.The surveillan e reeutrecent is satisfied by verifying the red block is functioning on the two-out-of-sequence groups as soon as RSCS is initiated. - The re_aining red block shall be verified as soon as-50' rod.

density is achieved and the RSCS is blocking 3 out-of-sequence groups. _ Additionally, 1

the, group.netch control inhibit function shall be verified as soon _as' the RSCS is automatically. initiated (as specified in GE-STS 4.1.4.2b2).

(Cont.)

10)-Technical-Specification Change Required:

Tes No

/o/e27//y

-11) Prepared By: d. C.

6tMd Date

12) Concurrence:

M_

Date

/0/2/ ft[

DirectorT-ReglateryCompliance

/

13) Approved:

p

.Date r e [Es [ P9 PNSf

/

e 0-RCI-02.3 Kev. O Page 5 of 6 m

9) Interpretation cont.

EASIS: The BSEP T/S does not agJe,e with the RSCS design or the CE-STS guidance.

The CE-STS does not specify the nue.ber of block rod groups. Tu o

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.,. w h-itc di v

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4 ATTACEMIh"T 1

.TECENICAL SPEC!T: CATION INTER.?RI!ATION REOLTST_ FOR.M L1, ferial-No.

85-01 (Rev 1)

. Technical ~ Specification

Reference:

3,6,3 Tebt, 3 A a.1

3. Technical Specification Pate No.:

'3/4 6-14 4.-

Subject:

Primary Containment Isolation Valves

'5. 7 tit (s)' Af f ec t ed:

3SEP Both

6. Description of Recuest /conetse detailed-description of requested-'

tr.terpretatien or problem area):

ne abeve ref erence 11st s cri arv e~nta!-,-. 4. M., < r.,. _..,....

Are theme the eniv valves te vk<ek ek<.....<re r.,4--

4

.--34.,q,,

7. Cririnator:

y, t*,

cS c.

Date:

6-4-83 J.

k*.

Chase

!. Reviewed 3 :7_

Date:

6-4-83

9. Interpretati:n:

The valves identified in SD Tables 2.1.0 ar.d -0.4,3 are censidered pri=ary containment isolation valves per T/S 3.6.3.

10) Technical Specificatic M nae Required::

Yes xte4?cios) No

-11) Prepared Ey

_ a/ --

Date

//d,/87 e

xn Date

//[r, /I~7

12) Concurrence:

Director (f Regulatory Coc:pliance

/

' 13): A;;reved:;

d, _ -i-Date

,i /17 f P l PNSC)

L RECE!VED i!OV 1 S GS7 0-RCI-02.3 Rev. 2 Page 6.-of 7 Et,P DGC. C.

'.2

,_.. m: _, _..

hkC C' E'!'/ E [)

Y AY

't 1035 ATTACHMIST 1

2.t.CO.CONTCL Ji

~~~ ~ T,ECESICAE SPECITICATION INTERMETATION KIQt'EST TORM

8) Serial No.

65-04 (35) __

1) Technical Specification

Reference:

3.1.5/3.3.2

2) Technical Specification Page No.: 3/41-18,3/43-12 y

3)

Subject:

SLC Operability bu in s

4) Unit (s) Affected:

BSEP-.

5) Descriptien of Request (cencise detailed description of requested interpretation er problen ares): Specification 3.1.5 requires that SLC be eterable in tedes 1, 2, 6 5 while Specification 3.3.2 requires the K'. C ; is:1stien due to SLC initiation be operable in modes 1, 2, 6 3.

k' hen r;st the is 1stien instrurentatien for Ek'CL' be operable.

6) Originater: Geerpe M111drer Date: 1/24/83

7) Reviewed ly: N/A Date:

9) Interpretation: ETC; ise:atien is required in codes 1, 2. 3, 6 5.

125:s:

As ELC is recuired : redes 1, 2, 6 5, instrure-tatien required te

.zi tain it operable rust be in service; therefore. Instrumentat4:n required te isolate RWCU et an SLC initiatien rust be eperable in codes I, 2, 3 L 5.

10) Te:hnical Spe:ification Change Required:

YesX (54TS507) yo i

11) Frepared Sy:

/ w/

l Date

/

4 o

12) C:ncurrence:

Late is

re

: ::

Eegu;at:ry C::; liar.:e U) A;;reved:

3,'

Late f!E'

.n s e

9 5

ATTACEVJ2iT 1 4

.ZOMICA1. 5?EC17! CATION INTUJRITRION RIQUEST TORM (1 )

B) Serial No.

P5-10 0

1) Technical Specification

Reference:

Table 3.3.5.7-1

2) Technical Specift:ation Page No.:

3/4 3-61

3)

Subject:

Tire Detection Instrueents

4) Unit (s) Af f ected:

BS EP Both

5) Descripti:n of Request (ceccise detailed descriptice of recuested interpretatics or procles area): Tne fire detection instrunents specified by Tatie 3.3.5.7. for Zone 4 of the ACC building are not consistent with instrurent s actually installed.

F 6} Origi:at:r: Eill Le nard page:

11/12!!2

7) Reviewed lyt N/A Date:

9) Interpretati:n: The tini:u instruzents operable fer Zene I. of the A00 Euilding are 2 flare detectors, 5 heat detectors and 0 smoke detectors.

RAS:S:

The s e mini:r;: instru:ents vere spe:ified in a request for a !SC submitted on Sept. 7, 1982 to the h*RC.

This TSC request was inappropriately superseded by later T3C requests to the NRC dated Dec. 13. 1962 and Oct. 17. 1983.

These later requests were based on out-dated infornation, contained a (Cont.)

10) Technical Specification CS a e Requirew:

TesX (85TSB16) No

11) Prepared By:

[ef

./

'k Date

/)

J'

12) Cet:urrence:_

Date 2C[?J "

Lirector g~ugulat:ry Cc:pliance

13) Ap;reved:

b7 14ffT' Date PMC

[

g 0-?.:!- ::,3 pe.. O

?:p ! cf 6

9) Interpretation (Cent.)

typographical error, and vere issued by the NRC as Amendments 66 and 92.

A TSC has been initiated to correct the A0G fire detection instruments identiffed by Table 3.3.5.7-1.

N

A77ACIMD'T 1 TECESICA1. SPECITICATION DiTEUKE!ATION FlQUEST TOP.M

8) Serial No.

8$-12

1) Technical $pecificatico

Reference:

~ ',

2) Technical Specification Page No.:

I'

~

3) Subjects _

RVCU Response Time I#*h

4) Unit (s) Affected:

R$EP--

}) Description of Request (concise detailed description of requested interpretatsan er prebles area) _

The response time for raCU differential flov specified by T/5 Table 3.3.2-3, Itera 3.a is (,13 seconds.

However, there are 45 second timers in the logic train.

A. '. Curtner

6) Originator:_

Datet 6/22/85

7) Reviewed ty:

P. D. Nsser p.g.:

6/19/85 5

terpretatient The restense time fer the RVCU differential !!cv is J

seconds.

H:vever, this time does not include the 45 second tirers.

N E A 5 '. 5 :

The KV;U dif ferential flow instrutentatien is part of the steamline break protection provided for the RVCU system.

The instracentation coepares the inlet and outlet flows to ensure the leakage from the system is below a specified ninicu= ( 153 gal /cin.). The system incorporates a timer to (Cont.)

10'TechnicalSpecif*attedChangeRequired:

YesX (TSC 64TSBg) 7!/e;l

11) Prepared By:

_c Date

12) Ccncurrence:

- _ -w,W Date 7

4

[

L1rectcr - Regul fery Cc:pliance

/

fv

13) Ap;reved:

.C3; p Date 7 /2 7 / M-PSSC J

(

/

0-R;1-92.3 Kev. O Page 3 ef 6 l

i

9)

Interpretation prevent spurious isolation during system evoluttens.

This tiner is set at a value high enough to allow the systen evolutions to occur, and below the tine used in the GE analysis for a line break on this line, i

i 4

L b

I i

?

-i

+

i 6

l ATTAC)MTNT 1 TECHNICA1. 57ECITICA7 % IN7171RITAT10N RECUI57 TOF.M

1. Serial No. 85-13

2. Tecnnical Specification References 3*7*2 3/4 i*3

3. Technical Specification Page No. _

4

Subject:

Control Loon Energency Filtration System (CBETS) Operability

.:....t; i.!! u t s ; ;

1;;7 --

I

6. Description of kequest (concise detailed descriptien of requested inter;tetation er problem area):

When chlorine or fire detectors are out of service or when these detectors are tripped so as to previde their required input to the C6tTS. is the i

Citts considered ersrable?

J. K. harrell (LESU) 11/27/65

7. Originater:

Dates J. P.. k r om

f. Reviewed By:

),g,,

11/27/65

9. Interpretation: The control building eeergency filtration syste is c;( rat it.

BAS:S The design criteria fer the CBETS is 10CTR$0 Appendix A Criterion 19.

s "Criterier 19-Control roce. A control root shall be provided fror. which a:ti:ns c E tt taken to crerate the n$ clear power unit safely under nortal cenditiois and to raintain it in a saf e condit ten under accident conditions.

~

Inclucang less-cf-coolant accidents. Adequate radiation protection shall be previded to territ arress and occupanev of the control roce under accident 6575E37>

10) Technical Specification

.ange Required:

Yes X(6575E24/ No

!!) Prepared Byt

/

Ad Date_

///f/sf

12) Concurrence

/

(~1

'l Date f ?/ ?/FJ' Directer - 1stulatory Corp 11ance isl>lTI

13) Approved:

t $

Date

~

PNSC 0-ECI-02.3 Rev. 1 Tage $ of 6

9)

Interpretation (cont.)

conditions without personnel receiving radiation exposures in excess of $ rea whole body. ~or its equivalent to any part of the body for the duration of_the accident."

s' : 44uirement in Design Criteria k9 for either fire or chlorine There 1:

detect <

i fsese sub-systems are a part of the plant specif fe design j

of the s c-o at Brunsviti, as described in the T&AR; however, they are i

not requtrea to meet Design Criteria 19.

I it-t'* eve *? et e attuativ.t of the fire dett:ti :. systtu, the CLEli

\\

is placed in the same operational status as vould be required by a high radiation signals therefore. there is na question of operability.

If a fire deteetor vere to fail such that it vould not provide an automatic start of the CitFS. system design vould still allow an autocatic start on a high radiation signal. Therefore, neither a fire detector f ailure not initiation can create a problem with CBET$ operability.

.In the event of a chlorine detector failure or actuation, the CnET$ -

vould align itself into the recirculatten mode of operation.

In this m>de. normal:sakeup of outside air (1000 s 's) to the_ system is isolated.

i With this loss of makeup air. the reduction in positive pressure within i

the control roce allows an increase in the in-leakage of unfiltered i

air ~ from approximately 275 scfm (CatTS running) to approximately 1375 stic (CitTS isolated).

a A Control Re:t Habitsbility-Evaluation petformed on BSEP by NUS-(per TM1 111.D.3.4) evaluated the dose which would be received byJcentrol room personnel during a LOCA with different values i

of unfiltered in-leakage.

This evaluation determined t'at the dose t

f rom airborn radioactivity in the. control root vill peak.and. level L

cf f at 2.5 ret thyroid and 0.004 ret whole cody for unfiltered in-leakage i

of 100.000 scfs or greater. When combined with the other sources of radiation to the control room. the sue totals are 0.41$.ree whole body and 0,6 ret-thyrcid. Based on this, neither a chlorine detector f ailure nor a Cntr$ isolation from the chlorine detection systen vill create a problem with CBEFS operability per Design Criterien 19.

t t

t

4 ATTACHMENT 1 TECINICAL SPECITICATION INTERPRITA!!0N RIQt'IS7 T0F#.

l. - Serial No.

86-01

2. Technical Specification,Referencer 3*1'3'3

'3. Technical Specification Fa8e No. __ 3/I4

f.

Subject:

Control Rod Scram Accumulators

}. TM

M Mft:ttt!

Unit 2 and t'r.it 2

6. Description of Request (concise detailed description of requested interpretation or probles area): What action is required when two or more control red scram accumulators are declared inoperable in Conditions 1 or 2?

4

7. Originator:

S. L. Russell STA Datet 1-28-86

6. Reviewed lyt M. R. Foss. SOS 1-28-86 p g,g_

9. Interpretatient With-two or mere control rod scram accumulators inoperable in hdes 1 and 2. operation is outside the defined boundaries of the ACTION statement and 7/S 3.0.3 stust be entered. To get out of T/S 3.0 3, the number of-Anoperable accueulators must be reduced to less than or equal to one.

This t

may be' de'ne by either cor.*ecting the problem or by inserting the associated control rod (s) and either electrically or hydraulically dfsarming the control rod (

10) Technic'al Specificatto Change Required:

Yes X(86TSB02) g,

11) Prepared Byt_-

Date _ p /a /f(,

y

12) Concurrencet_

W Date 2.

J 6

Director - Regulategy Ccepliance

/

1

13) Approved:__ h,,.. (p ( A,3.--

Date. 4//3/f [,

/

PSSC

/

0-RCI-0.3 Rev. 1 Page 5 of 6 c

1

The f inal s t a t emen t c f ACTION a s t a t e s "o t he rvi s e, b e in a t least BOT SEVTDOWS vit hin t he nex t 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s. " This statenant only applies if ACTION itens a.1 or a.2 are not coupleted within the required 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

This statenent has no bearint on the ACTION to be taken if mere than ene accutulator is ineperable.

BA515:

The purpose of the atran accuralators is to ensure that sufficient energy is availatie to insert the control rod in the ecst unf avorable condition.

This conditien vould be a hDCA vith a loss of of f-site power where the CLD systen and vessel pressure vould both be lost.

In this condition, the only available seurce f er control rod insertion vauld be the accumulator.

Insertion prevents a pattern of inoperable accuculators that vould result in less reactivity insertion on a scran than has been analyzed.

Inserting and electrically or hydraulically isolating such a control rod recoves it f rom the ineperable accuculator specification and places you in the incterable centrel red 0.1.3.1) specificatien.

Ik h

w _ _

,w

ATTACEMthi !

IECENICA1. SPICITICAT CN INTIK7kITATION RIfeUEST TCF#.

1. Serial No.

86-02

2. Technical Specification References _ section 3/4 3.2 Table 3.3.2-1

3. Technical $pecification Page No.l. 3/4 3-13 and 3/4 3-16

4. Subject Secondary Containment Requirement in Condition S

......., a. a e s. e u ESIl it.n (cr.t) ard boat 2 (tvo)

6. Description of Request (concise detailed descrietion of reeuetted interpretation er problee area) _

li D12-F.M-NOIOA and/or 80105 are inoperable while in Mode $, cost Secondary Containment Integrity be established with the $3Gi's operating within one hour? (Action statement 23)

7. Originators __

Ricky D. Tart

ate 1*26-86

8. Feviewed By:

A. Hegler Da t e s__I*27-86

9. Inter;retatient_ If the N010A and/or N0103 is inoperable in Mode $. the A !!!?: Wich should be taken is "Istablish Secondary Containment Integrity vith the standby gas treateent system operations within one hour or enter the AC710N state:ents for 3.6 45.1, 3.6.5.2, and 3.6.6.1 as applicable.

10) Te chnic'al Specification

.e Required:

Yes_ X(867S803) yo ll) Pre;ared ly:

L Date....pg//e/cc

12) Concurrence h(

s.n ~

Diretr5r - AMgu atorpoepliance__

Date_

S /1

/

13) Ap p roved i__.

h.,_ (t) Chaw Datej)/3/ / /

?.NS C

\\*

0-K01-02.3 Kev. I Nge 3 of 6

__.,_.__,m--.

- - - - ~ ' ~

BA51$1 The intent initiated by N010A(B) are established as the initiating instrof A those functions inoperable.

and to start the SBC7's on a high reactor bu13 ding exha umentation is Each of the functions required by ACTION 23 [$1CONDARY CONTAINF5

$bC7's, secondary containment dacters (inplied) requirenents, and all are required in the san.e m) odes.have their ovn unique AC710N not be able to be ic11oved (any of the three functions cannot be perforned)

Should ACTION 23 the implication vould be that 7/S 3.0.) sust be entered; however in Cold shutdevn.

, you are already A07? M ?3 d'

- tt!

t (t h: t ; set the required AC710N's for those functions (T/sof functions should N010A(B) be It 3 6 5.1. 3 6 3 2..., 3.6 6.1) are taken when they are inoperable, then the intent of ACTION 23 is net

~

RECEIVED ATTACXXENT 1 0\\)N 21 YA9 TECHNICAL SPECITICATION Ih'7ERPRITA710N RIOUEST F0F.M 87-01 (Rev. 1)

BNP DOC. CONTROL 3,

,,,g,3 g,,

2. Technical Specification Ref erence t

3. Technical Specification Page No.t

4. Subjectt_ Recuired Off Site AC Power Sources

5. Unit (s) Affected:

B$rr Both

6. Description of Request (concise d t,ailedjeggtigogtggs,ged f

interpretation or problem area):

int.etendent circuits between the of f site transmission network and the ensite Class It distributten system," as it applies to OPEhASILITY and ACTION state:ent s a,b, g c M M.

g s

7. Originatort leb Poulk, Jr.

Datet_

t/19'!9

8. P.svieve d By:

Earl Enter Dates _

6/19/89

9. Interpretatient Tor this LCO statetent to be satisfied, the fellevint cust te c.a in ta ined f or ea ch unit i g

1) carat {.e ei su;;1yint h:th the VAT and $AT.erable inc ceing t ran stis si:n line s to 1+4b*.Nvitc hyardf[e two :

8.-

u'

./ an CTerabe e distribution system f rca both the l'A7 and SAT to BOP buses 1(2)C

, arf 1(2)D.

a ar. cr etatie distribution syst er te supply power f rom the IG)C ar.d 1(2)D to it s respet tive etertency bus.

10) Techtteal Specification

.snas Requiredt Ye s_

No y

11) Prepared Byt p/

Date

/f//949

/7!d

12) Concurrence Date Direc tor - Pplatory Cotrpitance

13) Approvedt.

M./

w-Date 6 /IP / fi y

PNSC 0-RCI-02.3 Rev. 2 Page 6 of 7

[

l

Q BASISt-The orir,ina) interpretations provided by TSI's 85-08 (alternate / normal) and 87-01 (two independent off-site power.

sources) was based on_the design information provided in the r$AR as to how BNP satisfies the requirements of General-Design criteria 17..

The plant design document took credit for two off-site powerm feeds to the switchyard as meeting this criteria, rollowing a technical-specification submittel and subsequent discussions with the NRC on this matter it-bas been determined that-the requirement for two independent =off-site power sources i s met by the Unit Auxiliary Transformer-[(UAT) alternate) and

- the Startup Auxiliary Transformer ((SAT) normal).

This was addressed by-the--NRCrin their response to the technical speelfication change request dated May 25. 1989.

ft is recognited-that the UAT can not normally supply the enerr.ency; buses when the generator is off-line.

Credit for UAT operability is taken by the ability to backfeed through-the UAT-following the removal of the generator disconnects.

4

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I C '."I! ! 70 7.*

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Cieratten with inererable F.HL roet coalers i

3. 7,,,Peth

,<r.

. Tercit:t'. *. :f Fe:ves

'::ncita detai'ed descif:tten cf re: tested

tter retatt: * ?! t t : b ' # :: ares'i hat actien, if any, should be taken if k*

er.e er be-th F.HK rec:. coolers are inoperable?

t 1r.,.,,,,.4...t_

lob T:ulk, Jr.

8/27/67 3 ate,

!. F.evteved 3ys,_

tarl Enter Oate: 8/27/87 i

9. ht er;te ta ti:n (See Attached) 10' *e:hni:a*. frecift:atien Ch. e Recuired:

Tes 8775519 se

11) Frepa' red Byt.

dd,_ s M. _

rate _

g/21/E7

10) Cencurrence:

pa g e_ cf> g *9

~

Directer - kertdatory Cect'tance 13% Appicved:

C -- b. s tate

?) Lply,

Jse-1 e

7 RECEIVED i

l 0-1c:-02.3 Rev. 2 SEP 1 5 87 ret, 6 er -

L ENF DOC, CCt TROL

t

9) Interpretatien Ltt; ting Cen11tten fct 0;eraticn 3.5.3.0 c.

Ct.e LPCI roce coeler 3.6,0.2.

Ths suppression pool cooling mode of residual heat removal (ENK) syf ten shall be OPERAILE vith two independent cooling loops, ea:h loep censisting of tve pumps, one heat exchanger, and or.

root cooler.

A??'. 023:L:rY:

Cer.dit s en 1, 7. 3. 4*, and $*

ACTION (f or roe: cooler ineperability only):

a.

In Condttien 1, 2 et 3:

1.

With ene LPCI subsyste: room cooler inoperable:

Restere the inoperable LPCI subsystem roon cooler to a.

OTELABLE status within 7 days, or; b.

Le:enstrate the 0?thA1:LITT of the retaining redundant LPCI subsyste roce cooler by performing Surveillance lequirement 4.X vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least ence per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter for the next 7 days; Festere the incterable LPCI subsystem roce cooler te c.

CTIEALLI status vathan 14 days or be in at least HO; SEU!DOWN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHV7DOVN within the f ollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With both LPCI subsystem roet coolers inoperable, be in at least HOT SHUIDOWS within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUILOWN vithin the f ellowing 04 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

In Condition 4* and 5* vith one or more LPCI rutsystens roce co:lers it.cperable, take the ACTION required by Specificaticn 3.5.3.1.

The provisiens of Specification 3.0.3 are not applicable.

SL'KVIILLANCE FlQ"IF1MENTS 4.X Each LPCI subsyste: roen cooler shall be demonstrated OPEF.ASLE:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when required by ACTION a.1.b above by:

a.

1.

Manually starting the LPCI subsystem room cooler and verifying air flov and service water flov f rom the cooler unit.

Not appli'. 4tle when two CSS subsystems are OPERABLE per Specification 3.5.3.1 BAS!Et See EER 66-0460 m

3

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+-

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'T+M'-TM-

ATT;.J#.ENT 1 4

TICFSICAL SPECIPICATION !b"E7JAITA7:CN RIC'.157 FCT?.

1. Serial No.__

87-03

2. Technical Specificatten Referencet_

3.6.6.3

3. Technical Specification Page No. _

3/4629

4. Sub,4ect Oxygen Concentration

1. Unit (s) Affected:

BSE?.

both

6. Description of Recuest (concise detatled descriotten of raquested interiretation or prebles area)t_ seecificatien 3.6.0M reeu!re "en ke-24 heurs prier to a scheduled reduct ten of thereal rever te lest then 15'. ef _

ra te d t hettal pcive r."

Hev is this specification te be arrlied to " scheduled reducticn"?

?. Originater: _

t-i t_"

Date:

11/0! /! ?

2. Revieved By:

K. I. Enter Dates _

11/Of/B7

c. Interpretatien This spe:!ficatien it irtended t: allev fer the de ine rt ant of the dryvell prior to decreasing power to less than 15 er scheduled 7:ver reductiens requtring dryvell entry.

This is applicable for both planned and e_xiSent pcVer de:reases.

The require:ent r e::.ain s t ha t powe r cu s t be less than 15*. er the dryvell exygen concentration cust be less than 4%

vithin the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the action statement entered.

This has been concurred 10)ith tv Rerien 11 and NF1. Tech'nica7 spectrication O nee Required:

Yes 8775B25 No v

11) Prepared 3 2 7

e Date

////.f[87 1

12) Concurrence:

w Date

///6!/'?

Director - {jgulat.y Compliance

13) Approved:_

d.-- *~

Date.

I t[17 l T7 PNSC e-RECDMD

.. 1

.. <7-0-RCI-02.3 Rev. 2 pare 6 ef 7 g y. h, # C....,

u..~.,

a

i l

./-

RECEIVED ATTACEMENT 1 23ON l

TECENICAL $PECir! CATION INTE7.PRETATION RIQPEST TOPM DIIIEOl f

1. Serial No.

838-C/

Table 4.3.1-1 (la, Ib, 2a) Note "d"

2. Techuteti $pecification

Reference:

Table 4. 3. 4-1 (ld. 3a.b.e.d. 4a.i. e.d 4 Not a "c !

t

3. Technical Spectitcation Page No.s. 3 /4 3-7. 3 /4 ' 3 8 s 3 /4 3-43. 3 /4 3-4 3e

4.

Subject:

Nrt*TPOS E*; tier.IN: st'tMittN: t 'rtot'IRfm r I

3. Unit (s) Affected:

BSt?- 1 AND 2

6. Description of Request (concise detailed description of requested interpretation or probles area):

SEE ATTACEED SHE!? TOF. ADDITIONAL INT 0Fy,A!!ON Note d" would indicate that the referenced survefilences arust be perf ormed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter entering OPERATIONAL COND1710N 2.

Can the surveillances be perf err.ed in et tjoPEFAj{0NA1. CONDI!!ONS and still acet T/S7

7. Originator _

i ty;, 2704 Date

/Of MkI[/'9 P O M j w.~)

'Iljh', fhsv's?

e. Rutev.d 3 :

eate 7

9. Interpretation SEE ATTACHED SHEET TOR ADDITIONA1 INTOPyATION inese survet;;ances identified above s.ay be perf ormed in conditions other that

"...vit tin 10 heirs a f t er ent ering OPERATIONAE CON 0!*:0N 2".

i

10) Technical Specif tention Cinte Re,utrod:

Te e_ s e s 3J 2, No

11) Prepared By:

/2_

Date S//fM6 bate.

74 f'

12) Concurrence

%{es&

Dir et,o

ulatory Co pliance r

/ '

j j

. 13) Approved:

Y l

4,.

Date C-FO!-00.3 Kev. 2.

Taro 6 of i

..,_,,_.,__._,m.._,,,,Jy,...,__-_,_.,__,,,,,._,

,___.,___._._.__m.__.

l Attachment to Tech Spen interpretation request fors iten No. 6:

The tech spec surveillance requirements listed above refer to a Note

'd*

which states that the APM, IM, and SM trip functions arust be checked v'. thin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering operational condition 2 fres operationa'l Condition 1.

However, the note does not take into account switching back and forth from condition 1 and 2 vithin a seven day period.

As an example, it is possible the above listed tech spec trip functions are in surveillance because the applicab1: MST's were perform in the previous seven days to satisfy going to Operational Condition 2 from Operational condition 3 or 4 The Unit is then taken to Operational Condition 1 and immediately back to Operational condition 2 still within the seven day period.

functions af ter an extended period of being in Operat

'd* vas intended to require checking these trip greater than seven days; in which case the trip functions should be checked if the intent is to reasin in Operational Condition 2 for any extended period of t i.n e.

It should not be a requirenant to perforts the trip function surveillance again within the sase seven day period if it is intended to switch Operational Conditions while these trip functions still fall within the seven day surve111ar.ce requirenent.

the trip functions have been checked by the applicable MST

'd' required surveillance if ner al tech s;ec frequency ad the unit was taken from Cperational Condition 1 to Operational Condition 2.

T t e: Nc. 9:

Footnote "d" as ref erenced above was provided to allow the perfornance of those identified surveillances within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter entering OPEMTIONAL CONDITION 2. due to extensive technical requirements to perform them prior to entering MODE 2.

The intent of that footnote is not to prevent the performance of those surveillances when in OPEMTIONAL CONDITION 1 where it is possible to de so.

Theref ore, if a surveillance test is performed in OPEMTIONAL CONDITION 1 and is still within it required scheduling periodicity, the test need not be pertcreed again "within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter entering OPEMTIONAL CONDITION 2."

Tootnote "d" was requested by BNP as a T/$ change in 1985 to allow the perfornance of these suveillances af ter entering OPEMTIONAL CONDITION 2

-due to difficulties with performing thee, in OPEPATIONAL CONDITION 1 p r io r t o_ s hu t t ing down.

Before Acendment 96 and 121 vere approved on 3/86, perf or:ance of this surveillance was required in OPEMTIONAL CONDITION 1.

A TSC vill be issued to clarify the intent of Tootnote "d" as reflected

. k in Anadtent 96 and 121.

i p

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w wwmeg,e erw - c.er - -,- -

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4 4

A!TACMD'T 1 7!CKNICAL $PICIT12 A7!CN IN7tRPRETA!!CN RICtT$7 FCM

!. Serial No. 89-01

- i i

0. Technical Speeffication References.

"III'""

i

3. Technical Specification Page No.:

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Subject:

Core Alteration I

!. D..t:s; Affected 33 r? -

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6. Description of Request (concise detailed descriptirn of requested

. j[j in t e r p r e t a t ica. cr preb;en ares)l l at constitutit a "eere alteration" an ref erenced in the definitions?

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7. Originatst: _ k. H. Poulk. Jr.

Datet 2/17/89

!. Reviewed-3y K. E. Inter pag,,

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9. Interpretation:

A core alteration is the addition removal, relocation.

cr er.6Ser.1 ci fuel. Sources. incore instruments. or reactivity controls in the reactor core with'the head removed and fuel in the vessel by other than its ncrmal contro11ng nechanism.

Exattles: 1) An SRN or_ Control Rod repcsttiened or moved by its normal drive techaritst in its norr.a1 operating channel is not a core al' eration.

2) An SRM or a control blade assembly (cont.)

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10) Technical Spect!! cation Chense Required:

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Date.

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cA / 7/r9 Director - AervCs:ory Cte6tance

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Interpretation (cent.)

tenoved f rcm or relocated in the reactor vessel is a core alteration mechanism is a core alteration.3) The addition, removal. or relocation o

4) Removal of an LPM string or an IM. is a core alteratien; however, withdrawal or insertion of the 1M using it s nors.a1 drive system is not a core alteration.

of a centrol rod with the CAD system and then the removal of the drive 5) Withdr unit is not a core alteration.

blade guides is a core alteration.6) insertion, removal, or relocation of I

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.i Neither the DEFINITIONS section nor the REFUEL

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the Frunswick Technical Specificat eetten of adequate guidance f or thi s conceen. ions (BTS) pr ovi des NUREG 's, discussi ons wi th other BWR'sinterpretation, !

approved requirements that could not be met and a review ed a di e d ** e-*

the t v. s r p r e t a *. n o n,

with the cw erent bi5 wi th were conducted.

The original STS contained a definitio the Reactor Core" which was the icereun n ice "Alteretion Of Etendard Technical nee of Specifications the current con trol rod envenen t ut ththat definitt en is the dellowin (STS).

Contained within catient "Norea) defthed as a Core siternation.the con t!ol rod dri systee is not earenent of tr. core instrunentatten Hareal alteration."

is not de'thed as a cure in 1977, Brunswi ck began operati on under i

definition of CCEE ALTERATION as pro id d the STS.

The worced di f f erentl y from that i

v e

in the STS was

however, as the STS had to be accepted as presentedin the there was no a ncication that the intent operation at and-Brunswick continued with the understan rod movement was different, control s

was not considered a CCRE ALTERATIO ng that Espport that the inten' a;prevec NURE3's $141 was not changed is reflected in NRR (Fsemi)

NUREGs for and 1862 (Ferry),

STS*s eithervarious BWR's CORE A'TERATION which mirrorprovide wording (pl a i

These in the CEFINITICN section for the original the wording contained.within BTS or they provide allowed exceptions within the REFUELIN3 section of technical specifications.

Be unswick conducted a survey.cf other utili ti tneir es to determine hydraulic system as a CORE ALTERATION mal ancicates that the Brunswick The initial input with the industry practice interpretation is consistent far).

(seven utilities responded th s

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Ne basi s f or the term CORE ALTERATION i s provided i

however, bases are provided f or the individual REFUE spectitcations.

n STS:

"... all con tr ol r ods beThe basis for L

inserted during CORE ALTERACONT (sic, which implies that 3

that the two are no fue2 uill not be loaded 2n to a ce))t synonymous)7!ONS ensures without a control 4

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reactivity changes froe CCRE ALTERAT!ON, movement of control rods and SRMs/1RM o be considered a interlock and SRM operabilitysurveillances reautred by STS f or out to the requirement to move these componcould not be perf ormed du ents.

Therefore, approved methodology f orbased on the available history and c interpretation defines a consistent and CORE ALTEAATION, this n

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Appendix $

tvelt t ticr. of Ert.r,twici Technical Specification Interpretations

CVALUATION OT BRUNSWICK TECHNICAL SPEClflCATION INTERPRETATION A.

Technical Specification Intcrpretation Request forrn-Serial No. 84-06 TUTisicra>TLoth units)

ThesubjectofthisTechnical$pecificationInterpretation(TSI)is Service Water System Operability.

B.

Background

Technical Specification (TS) 3.7.1.2 states that the-service w6ter systen nuclear header shall be operable with at least three operable service water pumps.

However, the TS does not specify what service water pumps rnust be operable in order to setisfy the T$ requirerrtnts.

The 151 states that the service water systera for each unit has five pumps.

Two are designated nuclear service water pumps and three are designated conventional scryice water pumps.

The two nuclear pumps service the nuclear header and the three conventional pumps service the conventional header.

Each conventional service water pump can service the nuclear heaoer by closing a salve associated with the conventional headtr and opening a valve associated with the nuclear header. All ten service water pump'. arc located in the servict water building.

The licensee rade the interpretation that the Unit 1 and 2 service water y s terr sta l

L t cperabic witti et IG!t:

a.

Four optrable nuclear service water pumps for the site capable of suff ying service water to the nuclear headers; and in additier.,

l b.

Two operable Unit 1 (2) conventional service water pumps, each powered from a different division and capable of supplying both the Unit 1 (?) service water nuclear header and conver.tionel header.

The licensee perforred a f ailure analysis for the TS! to support the inte rprets ticr..

The licensee evaluated the assur.ed failurt cf each of thc four safet related 4160 volt buses / diesel generators to ensure cooling water to (y) essential components of the unit experiencing the loss of a

coolant accident and (b) the unit experiencing a safe shutdown condition.

The licer see concluded that the safety function would be fulfilleo.

The original TSI was approved by the plant nuclear safety corraittee on March 12, 1964 The 151 was modified a number of times since that time.

A Diagnostic tvaluation of the Brunswick Plant was conducted early in 1969.

One of the systems evaluated by the team was service water.

The team found serious deficiencies with the system.

Examples included single failure vulnerability, nuclear header to conventional header service water leakage, pump motor reliability and unavailable preoperational/startup test data.

The licensee developed a justification for continbec operatic.n (EER No. 69-0135 Rev 0, S/4/89) until such time that design changes could be implemented in the plant.

The T5I discussed here is Revision 5 which was approved by the Plant huclear Safety Coramittee on May 5,1989.

IMs revisio. reflects ita licensee's JCO.

2 1he licensee recogni2ed that the TS required a 15 change. One was submittte on hoverttr, 1966. As a result of the LE1 evalu6 tion and tht licensee's subsequent evaluation, the licensee requested the staff to rut the 15 ctange requtst on hold.

The licensee subsequently withdrew the application on October 30, 1989, on the basis that a new arplication would t'c sutritted it. February 1990, at which titte the lictrsee's evaluations would be completed.

C.

Lvaluation The present 15 lacks definition as to which three purps rnust be cre rat it.

This it ir fertent retuese the nuclear service rater purps stcrt on an act cent signal (if not already running) ano the coruntional i

purps do riot.

The 15 bases do not provide clarity.

The prestrit 15 would per:1it the two riucle'ar service water purps for each unit to be out of service, and the licensee would r ot be in an action stater.cnt. This is not appropriate.

15 that reflect the design basis of the system are necesscry.

The licensee's TSI is a step in the right direction.

The requirer,ents trctified iri tht irterpretetien ensur(s that the pitnt's desigt. tesis can be net.

However, the 151 Revision 5 is not a final substitute for the pre sent TS.

Tre licenset ccn chance tFr 151 at anytir,0 without sta f g r provai.

it shcsic t>e f.cted that this 151 is a revision to the 151 reviewed by the del.

1he ability cf the sers1ce water sy! tem to fulfill its design functico vn 4 r.c l e r cer ccm icer, tit ied by ttc l'iogr ostic b e luttict T(tr curing e Spring 19E9 inspection.

The liter see was notified that a violation is under censiceratior, for escalettd enforcer.ent action iri inspection Report 89 34, dated h'c vuber 30,19f 9.

An Enforcerent Conference was held on becerter 15, 1989.

b.

Concluticn 1here is no irrediate safety cor,cern at this time, as long as the TSI P(vision 5 remains in-effect.

There rady have been an imrediate saf ety cor.cern prior to Pevision 5.

If a further revision is acopted, it should have at least the sare level of safety. A 15 change that reflects the desigr t, asis of the system is necessary.

Escalated Enforcerent is pending.

E.

Feference As stated f.

Frincipal Contributor:

E. G. Tourigny

D ALUATI0li CF BRUNSWICL 1ECHNICAL $PEClflCAT10N IkTERPT: ETAT!0N A.

TECHNIC AL SFIClflCA110N lh1E ATEC1 AT10N Ff 0 VEST FORM SERI AL NO. 64 C8

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The subject of this Technical Specification Interpretation (TSI) is ccncerned with the Core irrey Tire F:elay B.

BACLGPOUND Brunswick Technical Specification (TS) Table 3.3.3 2, Jtem Id iridicates that the trip setpoirit f or the E21-K16A,8 tire delay relay is 14 f 1416 sec.

Tht TS1 statts tFat this trisy is cett' ally a $ see. reley.

This TS! was approvec by the plar.t f.uclear Saf t.ty Cor.nittee on May 24, 1905.

The licensee justified this 151 on the basis that the K16 relay is in series with the Sik 1A2,1E2 (2. A2, 2.02), which is a 10 second tine r61ay, to provide the 15 second delay for core sprey.

The correct tirr: delay value of the Ll( A,5 relay s is 4.5f t! 5.F seconds.

C.

QALUATION The staff has verified that the 15 second delay for core spray is cer.sistent with the safety aralysis of tht plant as docurented in the Brunswici l'pdateo final Sat (t;, Italy $1s l y crt (UT5I.T).

The staff rioted th6t even though the 151 value assigned to [21-K10A cr E is 4.5 t 5.5 sec., the core spray tint delay specified in the Ufi/E is 15 seconds.

Tin e celsy re lty s STE lid,102 (2. A0, T.EE) in st r ic s wi th t he E21.h16A relay provide the aeditional 10 second delay for the core spray tire delay.

Since the Uf5AR specifies the 15 second delay, the T5 should be chtrged to show ite correct rielays and/or tirnes fer the core spray systern.

The staff noted that although the 1964 T51 indicated that a TS change was

'titiated, cr.e was not sti.uitted to the NRC.

P.

CONCLUSION The staff cetermined that the ov erall delay tirne of 15 secchds for cort spray is cerisistent with the saf ety analysis of the Brunswick plant end should remain in the-15.

The staff considers the TSI incorrplete because it Qes oct address the design of the core stray tint de lay.

A T5 change should be submitted to the NPC to include the STt 1 A2,102 (2.A2, 2.B2) relays in the TS, as separa te items or contir'ed with the L16 relay, l

t 2

The staff concludes that a breakdown in the licensee's administrative controls appears to have occurred because no TS change was proposed to the liRC since the 151 was ivrrelized on May 24,198'.

s E.

[EF EREtiCE As stated T.

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EVALUAT10N OF ERUNSWICL TECHNICAL 3PEClflCAT10H INTERpKETA110N A.

Technical $gecification Interpretation Request f orm - Serial ho. 64-10 D evision 6/TIItl Units)

The subject cf this Techr.ical Specification Interpretation (151) is pressure / Temperature Limits for the Reactor Pressure Yessel.

b.

Bac)pround The licensee recognized that existing technical specification pressure-tt r;1 ra tu rt lir" figart! have not bttn revised te reflect current Generel Electric Lasis occuments.

Upon review of the documents, the licensee determined that the current 15 figures are not sufficiently conservative.

The original T51 was written in 1984 The licensee subtnitted a 15 change request by application dated October 26, 1908, as supplenented March 30, 1909 June 13, 1989, and August 4,1909.

The figures contained in the T51 (Revision 6) art to be used in the interim.

The pressure-temperature limits of the current 151 art based upon the latest NRC staff guidance (Revision 2 of RG 1.99) and 10 CFR 50, Appendix G.

The figures are ct;nteincd in tht 151.

Thit irterpretaticr was appreved by the plar,t Nuclear Safety Cornittee on June 29, 1989.

L.

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pressure terperature limits arc irrposed upon reactor coolant system ccrponents so th61 they are not overstressed during cyclic carotion.

The pressure-temperature limit figures in the 151 are exactly the sane and/or bounded ty the proposed and existirg T5 pressure-temperature limit figurts.

The staff's evaluations of the licensee's troposed technical specifications are almost complete.

No significant issues have been identified that would preclude issuance of the amendrent.

D. Conclusion The licensee's TSI does not represent an immediate safe) neern, since the TS1 figures are exactly the sare and/or bounded by the existing and proposed T5 figures.

The proposed T5 figures shoula be approved in the near future.

Although a T5 change has been submitted to the staff and should be issued in the near future, the T51 has been in-place as one revision or anothcr sir,ce 1984 This appears to be a breakdown in the lictnsee's administrative controls.

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C.

Reference As stated F.-

Pr_incipal contributor:

E. G. Tourigny

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[YALU/,T!0!i Of EEUhWICK TECHNICAL SP[CiflCATION INTLF.PkETAT10!i i

A.

TECHNICAL SFiClfICATION IN1[RPRCTAT_10N REQUt$1 TORM - SERIAL NO 8419 y

The subject of this 7tchnical Specificatior, Interpretation (TSI) is concerned with the nurt er of required Reactor Protection $) stem (RPS) - turbine ccotrol vaht f ast closure control oil pressurc Icw chantels.

L.

B AC KGROUND Brunswick Technical Specification (TS) 3.3.1, Table 3.3.1-1, litm 10 requires 4 operable channels per trip systun; however, the TS1 states that the plant design has only ? cFennels ter trip systera.

This TS! ras approved ty the Plant Nuchar Safety Cormittec cri October 29, 1984 The licensee justified this 151 on the basis that the nun.ber in the Erunswick 15 represents a typographical Jrror introduced when ti4 page was l

- retyped for TS ctange requests involving digital to analog inodifications.

1his errce wts incorporated in littnse anendnent nurber W issued by the l*C, C.

.EV A.t V AT 10!i The staf f's rivicw of sir.ilar TS for Ur it 1 showcd that there are only two channcis per try system, whict, is consistent with tre licensee's interpreta-tier, f ui Lt it i.. - h s ta f f a ls o r ev ie f.c the l'pda ted f ilia l Sa fe t) /r ely sis Report (UFSAR), Table 7.2.2 2 entitled

Reactor Protection System Mirein.um hun.ber of Charnels f(c,ut rt:0 for functional ferformance in Run Mode." The nurt er of thenrels per trip 5) sten shcwr in this tabic c' the UTSAC for the turtsrt centrul te ht f a:: t'c ture is ?.

The staff nuttd that although the 1904 TSI indicated that a TS change was netted, one was ret subritttd to the HEC.

D.

MCLUSION The staff cetermined that the correct nurter of channels per trip system j

for:the tuttine control valve fast closure is 2.

Alticugh thera 15 ro irrediate safety corcern, a TS change should te sutnitted to the URC so the Erunswict TS can be corrected.

The staff concludes that a breakdown in the licensee's administrative ccntrols arpeers to havt occurred because t.o TS change was prerosed to the NRC since the TSI was fortaltred on October 27. 1964, t

E.

REFERElg As stated T.

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EVALUATION OF BRUNSWICK TECHNICAL SPEClflCATION INTERPRETATION

-A.

Techr,ical Specification Interpretation Request form - Serial ho. 84-20 IB5thUnits)

ThesubjectofthisTechnicalSpecificationinterpretation(TSI)is Rod Sequence Control System (RSCS) Operability.

E.

_Ba c k c rou_n_d Technical Jpec:fication 4.1.4.2b requires vemonstrating the Rod Sequence Control System oprable by attempting to select and move an out-of-scqeente contrel rod in each of the otter three groups as soon as RSCS 1

1s automatically initiated when reducing thermal power.

However, between 22% full power and 50% rod density, the RSCS has only two out-of-sequence rod groups. The RSCS design for certain conditions therefore cannot support the technical specification requirement.

The surveillar:ce requirerent is satisfied by verifying the rod block is functioning on the two out-of-sequence groups as soon as kSCS is initiated.

The remaining rod block shall be verified as soon as 50% rod density i achieuc and the I:505 is blecking 3 out-of-sequence grcups. Additicu e ly,

the group nctch control inhibit function shall be s-ified as soon as the P.SCS is autoretically initiated (as specified in GE-STS 4.1.4.2b).

The licensee's basis for the TSI is that the TS do not agree with the RSCS design or thi GE - STS guidance.

In addition, the GE-STS does not specify the nurbt, of blocked rod groups.

The TSI was approved by the Plant Nuclear Safety Committee on October 29, 1984 C.

Evaluation The RSCS restricts rcd movement to minimize the individual worth of control rods to lessen the consequences of a control rod drop accident.

Control rod movement is restricted through the use of rod select, insert, and withdraw blocks.

The RSCS-is a hardwired, redundant backup system to the rod worth minimi:er (RWM).

RSCS is required to be operable below 201 power.

According to the UFSAR, the licensee uses 221 power to ensure that the T5 of 20% is met, in addition, the RSCS is automatically in anc out of service when the preset power level is reached.

The present power level is 301. The licensee uses two ranges for the RSCS:

1001 to 50% rod density and 50% rod density to present power level.

The first range uses a four RSCS group approach (sequence control).

The second range uses a single retch approach (group notch control).

One hundred percent rod density is when all rods are in.

. ~.

2 The TS specifies that the RSCS shall be demonstrated operable by atter.pting to select and racve an out-of-sequence control rod in each of the other three rod groups:

a.

In CONDITION 2 prior to the start of control rod withdrawal for a reactor startup, and

b. - As soon as the RSCS is automatically initiated during control-rod insertion when reducing thermal power.

r The b part of the TS cannot be met in the 50% rod density to preset power 1(vel range because of the system design. The licensee's approach is ecceptable until 50t rod density is achieved, at which time the licensee fully adheres to the TS.

In addition, the rod worth minimizer adds additional assurance that the plant is being cperated safely.

Thus, there is no irrmeciate safety concern.

Althcugh the 19E4 TSI indicated that a T5 change was nceded, one was not subr.li t ted.

Other licer. sees hcve requested the deletier. cf the T5 for RSCS and received approval by staff.

This was an effort led by the BWR Owners Group.

Erunswici r.ry desire to-take a sittiler approach.

D.-

Conclusion The inter.sistency between the TS cnd the plar,t design fer the range of 50L c' red ce r t it) anc pr esent power lesei coes not tresent er ir.r..tciate safely concern, because the licensee fully complies with the TS at the 50% rod density point and the RWM is acting as a backup.

A breakdown in the licensee's administrative control appears tu have

~

occurred because no TS change has been submitted since the TSI was formalizec cr, October 29, 1984 The staff believes that the TS should be corrected or deleted.

E.

Reference:

As stated

'f.

Principal Contributor E. G. Tourigny

EVALUATION OF ERUNSWICK TECHNICAL SPEClflCATION INTERPRETATION A.

Technical--Specification Interpretation Request form - Serial No. 85-01 (Revision 1)(Both bnits)

ThesubjectofthisTechnicalSpecificationInterpretation(TSI)is Primary Containment Isolation Valves.

B.

Background

Technical S cificatiun (TS) Table 3.6.3-1 lists the primary containant isolation valves and their associated isolation tius tnd group numbers.

A question was asked by plant personnel as follows "Are these the only valves to which this specification is epplicable". A TSI was written to address this which states "The valves identified in SD-12 Tables 2.4.2 anc 2.4.3 are considered primary containment isolation valves per TS 3.6.3".

50-12 is a plant procedure.

Revision 1 of the TSI was approved by the Plant Nuclear Safety Committee or, flovember 17, 1957 A TS change was submitted to NRC in February 1988, which, among other rec,uests, wculd delete Teble 3.6.2-1 valves frcra the TS.

The applicaticn was completely superseded in September 1989.

C.

Evaluation Table 2.4.I of SD-12, F e.ision 13, itay 8,1989, er. titled "Contair. ment

( ;t :ici t;. i cr c* rat ict.,'

ccr.tcirs t listirr (f ill penetrations, are for each penetration, valve numbers and containment isolation valve group information are illustrated.

Table'2,4.3 of 50-12 entitled " Containment Isolation by Valve Number" contains a listing of all valves; and, for each valve, the penetration associated with it.

TS Tcble S.6.3-1 contains a listing of autornatic primary containment isolation valves, their isolation group nuinber, and their isolation time.

A corr.pariser was made between the automatic isclation valves listed in 5D-12 and the valves listed in Table 3.6.3-1 of the TS.

It was noted that there were automatic isolation valves listed in the TS that were not listed in 50-12 and there were automatic isolation valves listed in 5D-12 that were not included in the TS.

For example, the following TS valves were not found in 50-12: E-41-F041; E11-F040; E-11-F079A and B; E11-F080A and B.

Likewise, there were more than thirty automatic isolation valves listed in 5D-12 that were not included in the TS table.

Because there were TS valves rissing from SD-12 and SD-12 was being used to meet the TS, the operability of the valves was not demonstrated.

Likewise, since the plant j

has more automatic isolation valves in 5D-12 than covered by the TS table, the TS table is incomplete.

=

'l 2

This TSI was also reviewed by the Diagnostic Evaluation Team during a spring 1989_ inspection, ard this evaluation confirms the Team's concern.

Inaodition,thelicenseefiledaLicenseeEventReport(LER)(1-8916)

- on July 14, 1989 which also confirms the results of this evaluation.

Lastly, a violation was identified on this subject in Inspection Report No. 89-34, dated November 30, 1989.

The NRC identified violation was not cited tecause criteria specified in Section V.A. of the NRC Enforcerent Policy were satisfied.

- An immediate safety concern does not exist bec6use the licensee is presently following the TS. However, the licensee violated the TS when certain valves were not included in 50-12 and the TSI specified followinc SD-12 instcad of the T!.

D.

Conclusion There is no immediate safety concern at this time. There could have been en inraediate safety concern for that period of time that certain valves were not included in 50-12. A TS change is under review to, anong other things, remove the valve listing from the IS and include the listing in a plant controlled document subject to 10 CFE 50.59.

E.

Rcference As statec F.

-Principel Contributor:

E. G. To ri rj 5

EVALUATION Of BRUNSW1CK TECHhlCAL SPEClflCATION INTERPRETATION A.

Te.hnical Specification Interpret 6 tion Request form - Serial No. 85-04 (Both Units)

ThesubjectofthisTechnicalSpecificationInterpretation(TSI)is Standty Liquio Control (SLC) system Uperability and relationship tr Reactor Water Cleanup (RWCU) Operability.

B.

Background

Brunswick Technical Specification 3.1.5 requires that the SLC system be c:erable ir ccrditior.s 1, 2, and E. Technical Specificatien 3.3.2 requircs tie Reactor Water Clearup system isolation due to SLC initiation be operable in condition 1, 2, and 3.

The licensee's interpretation is RWCU isolation is required in conditions 1, 2, 3, and 5.

This interpretation was approved by the Plant Nuclear Safety Committee on May 2,1965.

The licensee justifies the interpretatior as follows. As SLC is required in conditions I, 2, and 5, instrumentation required to maintain it operable must be in service; therefore, instrumentation required to iscletion RWC'J cr SLC iritiation tost be operabic in conditicns 1, 2, 3, end 5.

C.

Evaluation RWCU should be isolated when SLC is actuated because the ion-exchange resins in the RWCU syster wrtid remove boran and not give the ccsired reactisit3 effect. The lico tte is adoing a condition v.her. RWCU ruit isolated if SLC is initiated, namely 5.

In this case, the licensee is going beyond the technical specifications requirements.

The staff believes that RWCU isolation should occur when SLC is acutated in Conditier.s 1, 2, and 5.

In the case of standard technical specifications, it should be noted that condition 5 aise has the caveat "where a control rod is withdrawn".

Although the 1985 TSI indicated that a TS change was needed, one was not subritted to the hRC.

D.

Conclusion The staff has no objection to the licer.see's interpretation. Adding RWCU isolation during condition 5 represents no safety concern.

A breakdown in the licensee's administrative controls appears to have occurred. No TS change was subn.itted since the TSI was formalized on Itay 2, 198E.

2 The staff believes that a TS change is necessary in this case. The licensee should have consisterity of conditions between the specifications.

E.

Reference As stated F.

Principal Contributor:

E. G. Tourigny

EVALUATION Of BRUNSWICK TECHNICAL SPEClflCATIONS lHTERPRETATION A.

Technical Specificatior.. Interpretation Request Form - Serial No. 85-10 (Both Units)

ThesubjectofthisTechnicalSpecificationInterpretation(TSI)is-Fire Protection Instruments.

B.

Background

TS Table 3.3.5.7-1 specifies for Zow 4 of the A0G Building the following l

minimum operable instruments: one flame detector, 6 heat detectors, and 6 snc4e detecters The licensee states in the TSI that fire detection instruments specified by Technical Specification (TS) 3.3,f.7-1 for Zone 4 of the Augnented Of fgas (A0G) building are not consistent with the instruments actually installed.

The minimum instruments operable for Zone 4 of the A0G building tre 2 flame detectors, 5 heat detectors, and no smoke cetectors. These minimura instrun.ents were specified in a request for L

a TS change submitted on September 7, 1982 to the NRC. This TS l

request vcs unapprerriately scurseded by later TS change requests

.to the NRC dated December 17, 1982 and October 17, 1983.

These later requests were based on out-dated information, contained a l

typogrcf hical trror, and were issued by the hRC as Auencrients 06 ar.c W.

A TS change has been initiated to correct the Zone 4 for detection instruments identified by Table 3.3.5.7-1.

TFit ir ' c rpreta t u r tas apprned by the fl<r*. Nuclear Safety Comr i;tce i

on May 24. 1965.

C.

Evaluation The technical specifications should reflect actual plant equipn:ent.

A. review of bfSAR-Table 9.5.1-2, entitled " Detection System Summary",

identified ionization, thermal (heat) and flame detectors as being in Zone;4 of the A0G Building. Photoelectric (smoke)detectorsarenot listed.- The type of detection instruments contained in the UFSAR is consistent'with the licensee's TSI information; however, the UFSAR does l

not identify the number of detectors by-type.

The licensee's fire hazards analysis-in section 9.5.1.5 of the UFSAR stated that the exposures present in the A0G Building are not severe L

and because of an absence of either train, none could be lost due to-a fire.

Thus, a fire in the A0G Euilding,, Zone 4 would not inhibit safe shutdown of the plant.

l u

l~

m

2 D.

Conclusion The inconsister.cy between the number of ninimum operable fire detec.

tion instruments in the A0G Builoing, Zone 4 does not present an immediate safety concern, because the AOC Building is not needed for safe shutdown. A breakcown in the licensee's administrative controls appcars to have occured because no TS change was submitted since TSI was formalized on May 24, 1985.

The staff believes th't the TS should be corrected.

The TSI is correct.

E.

Reference A: Stated F.

Principal Contributor:

E. 6. Tourigny

i EVALVATION OF BRUNSWICK TECHNICAL SPEClf! CATION lh1ERPRETAT10N A.

Technical Specification Interpretation Request form -Serial No. 85-12 TIoth unitsj The'subjectofthisTechnicalSpecificationInterpretation(TSI)is Reactor Water Clean Up (RWCU) Response Time.

B.

Background

The response time for Reactor Water Clean Up differential flow specified by Technical Specification (TS) Table 3.3.2-3, item 3.a is less than or equal to 13 seconds. However, there are 45 second timers in the logic traint tnd this appears to be an inconsistency, according to plant personnel.

The licensees interpretaticn is that the response tirne for the RWCU differential _ flow is less than or equal to 13 seconds and this time does not include the 45 second timer.

The licensee's basis is as follows.

The RKCU differential flow instrumentation is part of the steamline break protection provided for the RWCU system.

The instrumentation compares the inlet and outlet flors to ensure tht leakage fror the systero is below a specified minimum of less than or equal to 53 gallons p r minute. The system incorporates a timer to prevent spurious isolation during system evclutticns. This tirer is set et a value high enough to ellow the syften evolutions to occur, and below the time used in the GE analysis for a line break on this line.

This TS change was submitted in an admendment application dated February 28, 1988 as supnlemented September 20, 1989. The TS change is under review.

The TSI was approved by the Plant n clear Safety Committee on July 29, 1985.

u C.

Evaluation The staff reviewed UFSAR section 7.3.1.1.6.17 " Cleanup System High Differential flow." High diff erential flow in the cleanup system measured between a point inmediately outside the primary containment and points downstream f rom the filter-demineralizers could indicate a break between these points. The automatic closure of-the cleanup system isolation velves prevents excessive loss of reactor coolant and release of significant amounts of radioective material. A break downstream from the filter-demineralizers woulc be less consequentici because of the low radioactivity of the water et that-point.

The high differential flow isolation trip setting was selected high enough to avoid spurious isolations yet low enough to provide timely detection and isolation. A 45 secer.d time delay is provided to allow the RWCU system to ride through

.g.

~

normal flow transients without a differential flow isolation occurring.

This setting is.high enough to prevent spurious isolations and low enough to arevent safety problems due to having a High Energy Line Break (HELB) witicut beino isolated within design limits.

The UFSAR specifies the flow trip setpoint of f 53 gallons / minute Based upon the above FSAR statements, it appears that the RWCU response tine due to high flow conditions is less than or equal to 45 seconds.

Thus the response time of less than or equal to-13 seconds specified in the TS is incorrect or incomplete.

The T5 change is under review.

D.

Conclusion The TSI does not cause an imediate safety concern because the licensee takes into account the 45 seconds in its analysis of line breats.- In additien, the licensee has submitted a T5 change which is under review.

The TSI was formalized in Jt.ly 1985, and it took the licensee ricre than three yetrs to request the change formally.

A breakdown in their accinistrative controls appears to have taken place.

E.

Reference 1.; sicted

f. Principal Contributor: E. G. Tourigny l

l l

EVALUATION OF ERUNSWICK TECHNICAL SPECIFICATOIN INTERPRETATION A.

Technical Specification Interpr'etation Request Foro. - Serial No. 85-13 (Both Units)

The subject of this Technical Specification Interpretation (TSI) is Control fuilding Emergency filtration System (CBEFS) Operability.

B.

Background

Brunswick posed the question:

Is the CBEFS considered operable when chlorine or fire detectors are out of service or when these detectors are tripped so at to provide their required input to the CBEFS. Tht licensee believes that the CBEFS is operable under such conditions.

The licensee provides the following basis.

The design criterion for the CBEFS is 10 CFR 50, Appendix A, General Design Criteria (GDC) 19.

There is no requirement in GDC 19 for either fire or chlorine oetection.

These sub-systems are a part of thc plant specific cesign of the CEEFS at Brunswick as described in the FSAR; however, they are not reovired to meet GDC 19.

In the event of an actuation of the fire detection system, the C6EFS is placed in the same operaticnal status as would be required by a high rocietion sigi,oi; therefcre, there is no question of operability.

If a fire detector were to fail such that it would not provide an automatic start of CBEFS, system design would still allow an automatic start on a high radiatier signal.

Therefore, neither a fire cetector failure nor initit.tico car crecte a probler. with CBEFS yerabilit).

In the event of a chlorine detector f ailure or actuation, the CBEFS would alien itself into the recirculation mode of operation, in this modt, normai makeup of outsice air (1000 scfm) to the system is isolated.

With this loss of makeup air, the reduction in positive pressure within the control room allcws an increase in the in-leakage of unfilterec air from approximately 275 scfm (CBEFS running) to approximately 1375 scfm (CBEFS isolated). A Control Room Habitability Evaluation performed on LSEP by NUS (per TM1 111.0.3.4) evaluated the dose which would be received by control room personnel during a LOCA with different values of unfiltered in-leakage. This evaluation determined that the dose f rom radioactivity airborne in the control room will peak and level of f at 2.8 rem thyroid and 0.004 rem whole body for unfiltered in-leakage of 100,000 scfm or greater. When combined with the other sources of radiction to the control room, the sum totals are 0.415 rem whole body and 2.8 ism thyroid.

Based on this, neither a chlorine detector failure nor a CBEFS isolation f rom the chlorine detection system will create a problem with CBEFS operability per GDC 19.

This interpretation was approved by the Plant Nuclear Safety Committee on Decemt,cr 5,1985.

.y.

C.

Evaluation The control room habitability system at Brunswick protects control room operators against such postulated releases of radioactive materials, toxic gases, and products of cornbustion. The single control room at Brunswick services both units.

The licensee incorrectly assumes that GDC 19 only addresses radiation protection of operatcrs.

it goes beyono radiation protection concerns.

In addition, GDC 3 and 4 applies. Las tly NUREG.0737, Itern !!I.D 3.4, specifically included toxic g s releases a,s a scenario control room opcrators nust tc protected agairi:1.

The licensee stated that in the eveat of an actuation of the fire detection systern, the CBEFS is placed in the seme operational status as would be required by a high radiation signal. Therefore, the staff agrees that the CBEFS is operable.

The licensee stateo that if a fire detector were to fail such that it would not provide an automatic start of the CBEFS, system design would still aHtv er autorstic start or, a high radiation signal.

Thcrefore, the staff agrees that the CBEFS is operable, so far as a radiation type accident, such as a LOCA, is concerried.

However, the licensee would have to follow the technical spcifications for a failec fire cetector or multiple detectors.

The licensee would have to make a CBEFS operability deterrnina-tion from a fire protection perspective.

The licensee states that in the event of a chlorine detector failure or actuation, the CBEFS would align itself into the recirculation mode of ope ra t inr.. Therefore, the staff believes that the CBEf5 is operable, as fcr as a radiction type. accident, such as LOCA, is concerned.

However, the licer.see would have to follow the technical specification for a failed chlorine cetector er multiple cetectors.

The licensec-would have to make a CEEFS operability determination from a chlorine release perspective.

Tte licensee determir.ed that a TS change was necessary. One has not beer submitted since the TSI was put in effect in Decertber 1985.

D.

Conclusion The licensee's TSI does not present an immediate safety concern. However, the licensee does not hcve a firtr grasp of the applicable GDC's and others Bruriswick TS that have a bearing on CBEFS operability.

In addition, Detectors failing low versus high should be addressed if the differences are material.

A breakdown in the licensee's administrative control appears to have occurreo because no TS change ha. been submitted since the TSI was forrralized cn December 5,1985.

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-E.

-References As stated F.

Principal-Contributor:

E. G. Tourigny y

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EVALUATION Of ERUNSWICK TECHNICAL SPEClflCATION 1HTERPRETATION A.

TECHNICAL-SPECIFICATION IN1ERPRETATION REQUEST FORM - SERIAL NO. 86-01 (50TH UNITS)

The subject of this Technical Specification interpretation (TSI) is concerned with actions to be taken when more than one control rod scram accumulator is declared inoperable in OPERATIONAL HODES 1 or 2.

B.

BACKGROUND With two or more cchtrol rod scram accur:ulators INOPERABLE in OPERATIONAL MODES 1 and 2, operation is outside the defined boundaries of Brunswick Technical Specification (BTS) 3.1.3.5.

Therefore BTS 3.0.3 must be entered.

The TS! restates this fact.

It elso states that to exit the actions cf BTS 3.0.3, the nur.ber of inoperable accumulators must be reduced to less than or equal to one.

The TSI woulo allow 3.0.3 to be satisfied by (1) correcting the problen or (2) inserting the associated control rods and either electrically or hyoraulically disarr.iPo the control rod and taking the ACT10h5 required by BTS 3.1.3.1.

The licensee justified the TSI on the basis that inserting and electric 11y cr bydraulicelly isolating such a centrol rod reroves it from.the inoperable

- accumulator specification and transfers control to the control rod specification (BTS 3.1.3.1 ).

C.

EVALUATION The' current TSR Standarc lechnicel Specifications (STS) ACTIONS for raore than one cci:trc l rod accertletor liXPERASLE (STS 2.1.3.5), specifies ttat the

.ACT10flS for INOPERABLE control rods (STS 3.1.3.1) be followed.

The intent of the TS1 was to institute, administratively, actions which are equivalent to ACTIONS which are allowed by the STS.

The ACT!0liS in the STS ere safe and en an.endment in this area to the BTS to be consistent with the STS wculd likely be approved by the NRC.

Therefore, the ACTIONS which the TSI institutes are equivalent to the--STS-and do not creatt 6 safety problem.

However, the suggestion of the TSI that BTS 3.0.3 can be met by inserting and disarming the control rods associated with the inoperable accumulators is not acceptable. BTS 3.0.3 can only be met by shutting cown or restoring the inoperable equipment to operable status, meeting the remedial measures specified in the LCO's ACTION statements or entering an OPERATIONAL NODE in which the LCO is not applicable.

.g.

D.

CONCLUSION The staff hcs determineo that the licensee is misapplying the requirerents of BTS 3.0.3 with regards to ways of terminating the actions of the spec'ficatien.

Although there is oc is.ediate safety conccrn, a TS cher.ge should bc submitted to the NRC to bring BTS 3.1.3.5 into conformance with the STS.

E.

REFEREllCE As stated f.

PRINCIPAL CONTRIBUTOR OTSB

EVALUATION Of BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION A.

Techriical Specification Interpretation Request form - Serial No. 86-02 (Both Units)

The subject of this Technical Specification Interpretation (TSI) is Secondary Containment Integrity Requirement in Condition 5.

B.

Background

Technical Specificction 3/4.3.2 identifies crerability and surveillance requirements for isolation actuation instrunientation. Rediation monitor in the reactor building exhaust are addressed. An action staternent is provic'ed if the rLir,inum number of channels are not opertble.

If ont channel is inoperable, it rnust be repaired within two hours or ACTION 23 rnust be takcn.

If both channels are inoperable, ACTION 23 must be taken.

The licensee poses the question:

If [ radiation monitor) D12-N010A and/or H0108 are ir.cprable in l' ode [ Condition) 5, must Second:ry Cchteinn.ent Integrity be established with the Standoy Gas Treattnent (SBGT) System operating within one heter (Action 23). The licensee provides the following interprctatier.

If the N010A crc'/or N010E is inoperable in Mode 5, the ACTION bbich should be taken is "Establisn Secundary Containment Integrity with the standoy gas treatment systern operating within one hour or enter the ACTION state-rents for 3.6.E.1, 3.6.5.2, and 3.6.6.1 as applicable.

Trx it. tera (; ACTIO! :titcr er,t R tu 1/5 3.3.2 is to ensurc that Inuse functions initiated by N010A(B) are established, as the initiating instrumentaticn is inoperable. These instruments are used to isolate the-secondary cciaainrtent dampers and to start the SEGT's or a high reactor building exhaust radiation signal.

Each of the functions required by ACTION 23 [SEC0hDARY CONTA!hMENT INTEGRITY, SBGT's, secondary contair.nent dampers (implied)] have their own unique ACTION requirc.ents, and all are required in the same modes. Should ACTION 23 not be able to be followed (any of the three functions cannot be performed), the implication would be that T/S 3.0.3 raust be entered, however, you are already in Cold Shutdown.

ACTION 23 dictates a defined set of functions should h010A(B) be inoperable.

It,the required ACT10N's for those functions (T/S 3.6.5.1, 3.6.5.2, 3.6.6.1) are taken when they are inoperable, then the intent of ACTI0h 23 is met.

The licensee determined that a TS change was neeoed.

The licensee submitted a TS arter.cment on February 13, 1959.

This TSI was approved by Plant Nuclear Saf ety Committee on February 13,19E.6.

-. C.

Evaluation Each radiation monitor can isclate the secondary containment dampers and start the SEGT system (both trains).

There is only one channel per trip system. The inoperable channel need not be place in the tripped condition as long es it can be restored to operable status within two hours.

If if cannot be restored within two hours, secondary containnent must be esteblished with the standby gas treatrr,ent syttem operating within one hour.

If both are inoperable, AC110N 23 must be follonc.

The staff does not agree with the licensee's interpretation. The lictr,see takc4 crc of two actions: ACT:0N 23 or ente.r AC110h statements for 3.6.5.1, 3.6.5.2, and 3.6.1, as applicable 7 The staf f believes that *or" should be replaced with "anG".

all applicable action statemer.ts must be completed (isolation instrumentation, secondary containment integrity, secondary containnent isolation dampers, and standby gas treatrent system). Since ACTION statement 23 has the shorttst tir'e frarne for curipletion, it should be completed prior to conpietion of the other action statements.

The various times associated with the ACTION statements should not be taken sequentially.

The licensee submitted a TS change which is currently under review.

Althousn the liter see submittec 6 TS change request, the licensee's TSI has been in effect for a number of years; and a breakdown in administrative controls appears to have taken place.

l.

C t r c ic i e r.

The staff does not believe that an incediate safety concern exists when tre licensa u lt-s ACTI0t; H, but does telieves thct cn irrecicte safety concern could exist if the licensee takes the other ACT10N's while in Cendition 5 and ACTION 23 is not also taken while irradiated fuel is being roved. A breakcown in the licensee's admininstrative controls appears to have occurred because it took the licensee a number of years to request a TS change.

E.

Reference As stated F.

Principal Contributor E. G. Tourigny i

l

EVALUAT10h Of BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION A.

TECHNICAL SPECIFICATION INTERPRETATION REQUEST FORM - SERIAL NO. 87-01 T)ev. 1) (Both Units)

The subject of this Technical Specification Interpretation (TSI) is what constitutes "two physically independent circuits between the offsite transnission network and the onsite Class IE distribution system "

B.

BACKGROUND Brunswick Technical Specification (TS) 5.8.1.1 requires "...two physicall) _

independent circuits between the offsite transmission network and the onsite Class 1E distribution system..."

In order to meet this requirenent the TSI states that the followu.9 nust be treintained for each unit:

1) two operable incoming transtnission lines to the switchyard capable of supplying both the Unit Auxiliary Transformer (UAT) and Station Auxiliary Transformer (SAT).

2) ar. crerable distribution systen from both the UAT and SAT to Balance of Plant (BOP) buses 1(2)C and 1(2)D.

3) ar cKr able distribution sy sten to supply power fror.; the 1(2)C ar.d 1(2)b buses to the crergency (E) buses.

The license justities this TSI en the basis of the desigt, inforr..ation providec in the FSAR cescribing how Brunswick rneets the requirements of GDC 17 and c'iscussior.s and correspondence with the NRC.

C.

E'. At LAT ;0h / ;Cl:CL US:Ch The staf f has verified through FSAR figure 8.3.1 1 that the requirements c: set forth ebat assurt tue physicilly independent circuits to the C16sr IE oistribution systera.

Thus the TSI is in conformance with the current staf f positior.s with recards to the configuration aspect of GDC 17 However, as a part of the Diegnostic Tearn Inspectiori followup affort, the staf f requested by letter dated November 1, 1989, additional information on how Brunswick satisfies the requirements of GDC-17.

The staff is waiting for CP&L's response, it should be noted that the above evaluated TSI is a revision to the TSI reviewed by the DET.

D.

REFERENCES As stated E.

PRINCIPAL C0flTRIBUTOR

~TSR

EVALUATION 'Of BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION t

- A.

TECHNICAL = SPECIFICATION INTERPRETATION REQUEST FORM - SERIAL-NO. 87-02 jECTHUNITS)-

~

The subject of this Technical: Specification Interpretation (TSI) is the action that should be taken if-the residual heat removal (RilR) room coolers are inoperable.

B.-

CACKGR0_ND U

Brunswick Technical Specificatior.s (BTS) do not inclucie a technical specification for the RHR room coolers.

The TSI )rovides the administrative -

controls-for the RHR room coolers in the form of..imiting Conditions for Operetion (LCO),- Action Statere.nts and Surveillance requirements, i

C.

EVALUATION The RHk roorr. coolers are support systerns necessary for the operation of the RHR/LPCI system and decisions about-the operability of the room coolers and RHR/LPCI system are the safety responsibility of the licensee. The' licensee is-cllowed to develop and implement administrative controls and procedures for the operation of support systems.

In the decision process, the licensee _ must rely

- en the definiticn of operability and.the'UFSAP.

The key qutstion -is whether -

one-RHR room cocler is capable of_ handling the heat load frort the. design basis eccident and naintaining the room traperature within the opera ting liraits for-the' space n,bich conteins both MiR/LPCI subsystens.

The HEC staf f eumir.ei the

~

UFSAR, the TSI,-and discussed the system and the RHR/LPCI system space with the '

resider.t inspector.

The available information is incensistent; therefore, _the NEC is urable to_ answer this key Questier.

If one roor cooler can haridle the accicu t heat 1rfd for both ME,lFCI setsy ster;,s, remedial actier.s line thost ir the TSI are reasonable. - If one room cooler can not handle the accident heat load fcr both RHR/LPCI subsystems, the remedial _ actions in'the TSI are inadequate because they are not cebsistent with the artichs for en ir.cperable EUR or: LPCI subsy stem -(BTS S.5.S.2).

The allowed outage time for one inoperable LPCI subsystem is 7 days; the TSI proposes 14 days for one inopera >1e. room cooler.

Also,'the-7 day allowed outage time for an inoperable LPCI subsysten relies on the cperability of. the core spray system.=

- D. CONCLUSIONS

The stoff-has determined that the licensee needs to evaluate the RHR roon cooler design to determine its capabilities to maintain RHR area tempeature within design -linits during a DFA. - Af ter this determination has been made the

. TSI'should be adjusted to conform to the results of the analysis.

This justification should be included in the TSI.

E..

REFERENCES As steted F.

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EVALUATION OF BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION A.

TECHNICAL SPECIFICATION INTERPRETATION REQUEST FORM - SERIAL NO. 87-03 T F M F DT(IT S )

The subject of this Technical Specification Interpretation (TSI) is the application of the words " scheduled reduction" in Specification 3.6.6.3.b which requires the oxygen concentration to be within a certain limit whenever thernal power is above 152 of rated thern!al power.

5.

EACKGROUND Brunswick Technical Specification (TS) 3.6.6.3 requires the primary containment oxygen ccocentratier t e Icss than 47 by volure whenever thermal powtr is greater than 15% of rated thern.a1 power (RTP). The LCO is applicable in Mode 1 during the time period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter exceeding 15% RTP to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing power to less than 151 RTP.

The TSI states that specification 3.6.6.3.b is intended to allow for the deinerting of the drywell prior to decreasing power to less thsn 15i RTP or scheduled power reductions requiring drywell entry.

This is applicable for both planned and exigent power decreases.

C.

EVALUATION / CONCLUSION The staff has reviewed this position and concurs with it. A TS bases page

. change was submitted by the licensee on the subject.

It is under review.

D.

TEFERENCES As stated E.

FRlhCIPAL CGNTRIBUTOR OTSB

EVALUATION OF BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION A.

TECHNICAL SPECIFICATION INTERFRET A110N REQUEST FORM - SERI AL NUMBER 88-01

( EDIH UNITS)

The subiect of this Technical Specification Interpretation (TSI) is concerned with the timing of the perforrar.ce of the required surveillances for APRM, IRP, and SRM trip functions upon entering OPERATIONAL CONDITION 2.

E.

E. AC KGROUND Erunswick Technical Specifica tion Table 4.3.1-1 Items la, Ib, and 2a -

Note "d" and Table 4.3.4-1 Items Id, 3a, b, c, d, and da, b, c, d - Note "d" require that when changing from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION I, the required weekly surveillar.ces sPecified in the 16t.les should it perforrned within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OPERAT10hAL CONDITION 2.

The TSI states that Note "d" was intended to require checking these trip functions af ter being in OPERATIONAL CONDITION 1 fcr an extended period of tine, i.e.,

greater than seven days; in which case the trip functions should be checked if the inter.t is to r(rain in OPERATIONAL CONDITION 2 for any extended period of tite.

C.

_E VALL't TI ON The surveillance requirerents associated with Note "d" state that the APRP, IRl', and SPJ trip functions must be crected withu: !? hours of ente.rtng OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1.

The intent of the footnote is not to prevent the performance of those surveillances when in OPERATIONAL CONDITION 1 where it is possible to do 50.

However, the note does not specific 611y address sutetig baci er.d furth betteen OPETIT:0ML C0h0IT10N 1 and ? Hihin a sever day period.

The steff's position is that as icng as a component's surveillance has been

-perforced within its scheouled frequency, the surveillances do not have to be redone upon entering OPERAT10ht.L Conditions where the component is required to be OPERAELE. Thereicrt, if the surveillance requirerent was completed within the last 7 days, it does not have to be performed again upon reentry into MOLE 2.

The Staff noted that although the 1988 TSI indicated that a TS change was needed, one was not submitted to the NRC.

D.

CO N_C_L US IO N It a surveillence test is performed prior to entering or while i_n OPERATIONAL CONDITION 1 and is still within its required 7 day surveillance interval, the test need not be perforrtd again "within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter entering OPERATICNAL C0hDITJON 2."

2 Even though the TSI states that a Technical Specification change would be-submitted, the staff does not see the need for a technical specification change.

E.

REF EREldCE As stated T.

PRINCIPAL CONTRIBUTOR OTSB

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l EVALUATION Of BRUNSWICK TECHNICAL SPECIFICATION INTERPRETATION A.

TECHNICAL SPECIFICATION INTERPRETA110N REQUEST FORl1 - SERI AL NO. 89-01 IWlh UhlT51 The subject of this Technical Specification Interpretation (TSI) is what cor.stitutes a

CORE ALTERATION" as referenced in the definitions.

E.

B ACKGF OUND The Erunswick Technical Specifications uses the definition of CORE ALTERATIOL a s pros iced in the STS.

The TSI states that the original Brunswick Technical Specifications (STS), prior to the BTS change in 1977 to the STS, contained a definition for CORE ALTERATION with the following clarification; " Normal control rod roverent with the centrol rod drive hydraulic syster is not defintd as a core alternation.

Norrel reven.ent cf incore instrunentation is not defined as a core alteration." Since there was no indication that the intent was different, operation et Brunswick continued with the understanding that control rod movement was not considered a COP.E ALTERAT100, in additicn the TSI classified the following items as a CORE ALTERATION or not a CORE AL1 ERAT 10f.:

1) An SRM or Control Rod re?ositior,ed or movec by its normal drive nechanism in its normal operating c1annel is not a CORE ALT E RATION. 2) An SFP cr a contrel blade assertly removed from or reloca ted in the rtector vessel is a CORE ALTERATION.

3) The addition, removal, or relocation of sources and/or fuel by any r'echanism is a CORE ALTEff. TION.

4) Reroval of er LFRM string cr en IRP it a CChE ALTERATION; howev(r, witr,drawal or insertion of the IRM using its normal drive system is not a CORE ALTERATION.

E) Withdrewal of a control rod with the CRD system and then the terroval of the drive unit is not a CORE ALTERATION. C) Insertion, renoval, or relecetice of blace guidet it e CORE AITEF AT10tl.

C.

EVAgATION In a merorandum from Gus C. Laines Assistant Director for Regior,11 Rea ctor, Civision of F(actor Projects 1/11, NRR to Albert F. Gibson, Division cf Peactor Safety Region 11, dated July 7,1907, NRR stated thut based on its review of the Technical Specifications for a number of plants, the NRR staff found that the definition of CORE ALTERATIONS in TSs for rnore recent plants (such as Susquehanna 2, fermi 2, River Bend and Perry) and recent license etcendrents for some plants (such as Hatch 1) includes a staterent that nornal movenents of the SRMs, IRHs, TIPS, or special moveable detectors are not cor.sidered CORE ALTERATIONS.

D.

CONCLUSION The staft has deterrtined that the interpretation of CORE ALTERATICh provided in the TSI is in conformance with current staff positions as cited ir the July 7,1987 Laina s rnermancum.

E.

REFERENCE As sttted F.

FRiff NAL C0hTRIbuTOR OTSE l

l

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