Text

Partially Deleted Ack of Receipt of 850812 Concerns Re S&W Practices.Summary of Concerns Encl,Focusing on Unacceptable Specs for Station Svc Transformer & Incorrect Specs for 600-volt Power Cables
	ML20214P234
Person / Time
Site:	River Bend
Issue date:	09/30/1985
From:	Emerson M; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	; AFFILIATION NOT ASSIGNED
Shared Package
ML20214P025	List: IA-86-687, Forwards SER Re Licensee 860917 & 19 Ltrs Proposing one-time Change to Tech Specs,Removing Operability Requirements from Automatic Isolation Valve in 3/4 Line That Bypasses RCIC Inboard Isolation Valve.Salp Input Also Encl; ML20214P022; ML20214P084; ML20214P101; ML20214P121; ML20214P143; ML20214P154; ML20214P168; ML20214P178; ML20214P206; ML20214P234; ML20214P615;
References
	FOIA-86-687 NUDOCS 8612040122
	Download: ML20214P234 (3)
	v • d • e

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J'/ " %,'*6 UNITED STATES NUCLEAR REGULATORY COMMISSION

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SUBJECT:

RIV-85-A-0093 This refers to your letter of August 12, 1985, to the Office of Inspection and Enforcement in which you expressed concerns related to Stone and Webster's practices at the River Bend Nuclear Power Plant.

An enclosure to this letter documents your concerns, as I understand them, based on your letter.

We have initiated actions to develop and examine the facts and circumstances of these concerns; therefore, if the enclosure does not completely and accurately reflect all of your concerns, please call me collect as soon as possible at (817) 860-8100, so that we can assure they are adequately addressed during our review.

If you should call and I am not available, please leave a message and I will return your call.

In closing, let me assure you that every effort will be made to resolve your issues. The U.S. Nuclear Regulatory Consnission is responsible for protecting the public in the uses of nuclear facilities and materials. We engage in major efforts on a daily basis to fulfill that responsibility. We appreciate conscientious citizens bringing legitimate concerns to our attention.

Should you have additional questions, or if I can be of further assistance in this matter, you can call me collect at (817) 860-8100 during normal business hours.

Sincerely,

// /

Emerson Investigations / Allegations Coordinator, Region IV

Enclosure:

Statement of Concerns 8612040122 861125 PDR FOIA GUSTE86-687 PDR q

M-Statement of Concerns RIV-85-A-0093 u

1.

Unacceptable speci-fications for station service transformer.

2.

Incorrect specifications for 600V power cables.

3.

Stone and Webster is not responsive in correcting deficiencies, slow in correcting deficiencies, and makes incorrect deficiency corrections.

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~2 Luclear Regulatcry Connission Yashington,

.C 20555 Cer.tlemen:

'. ave had occacier.

review various specifications prepared by Stone & Webster f:

.he Iiiver her.d : : clear rower 112nt and have made extensive con ents on then, prtaining to otvic;. o.ierions, lack of techni:12 cm r.i inconsistencies in t,e texts. !ha tecnnical quc2it;. of their product i-for,clow any acceptable level.

"'nese co=entr hnve been discussed with Stone & Webster r.d problem areas were peinted out to ther.. :iewever, the situation has not ir. proved enti there exists seriour concern concarnir.c 2'one ?, Webster's performance en the 2iver Jend project. Here

=.:e a few ext--ler restrding the quality of their werk:

. "'he specific.ti:n fer the nor al station service tr:nsferer was technically unacceptable wp c 2.;7.1'1), ss follows:

a. *he EII fer ' ne ;ransf::- er secondary windincs (..1(k[ was specified the sa e as that fer th prirri windings (13 0kV). :hir war. Tong.

b.

No tap en the hign side of the transforner vs mpecified, although one was required.

c. The sane ph r: dirplacement for the existing trr.sfe:- er (delta-delta) was specified for the na ' tr nrformer (delta-wye / wye). Thi:. war _n error.

d. No neutral buu.ir.g w;s specified; one was re,uire?.

e. Thd rimpehnce h wc er t".e two low winding * (,

. s r. - + specified. This is T. ;uirer..

f. The short cire..i. capability was not included.

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6. The spec.dh cenditiona cf service (norm.1 - acciden'il) for outside the c ont ainr.c n'....csc -:o not apply to outdom t

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2.

he specific.tirn f r f007 power cables (241.230 ' t.'.

.rher of errors, to wit:

a. The centrar' Ir- *irien vis deleted which al?cvr fer c'.: ncin6 the quantities of cable prior to rele 2c for fabrication.

i.. The de:cript10r. Of the three conductor cabic u.d c ble ri es 3/0 58 and #10 was delete / althcud; these sizes were specifically re 2 ired.

c. The requirei que.ntities o r cable were chan6ed,drs.stically changed 3 despite the fact that the contract did not make provision for such chan6es.

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The 6as turner test section was changed fron 400,000 31T to 70,000 BTU

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-2 despita th2 fact that tha clicnt, Gulf Stc.t:a Utilitico, specifically required

.he 400,000 BTU test and paid extra for it (non-standard test).

e. The whole test section was unnecessarily revised, and conflicted with purchase

rder requirements.

f. Cable reel lengths were specified without considention for the weight and physical handling that would be required, e6,1/c - 500 hCE copper cable was specified witr a reel length of 9100 feet, which would wei6h about 15,000 lbs.

3. icen there and nunerous other deficiencies were called to Stone & Webster's attentier, cor ections were sicw in forthcoming, often not at all, or altered so as to add new prehlems to the spac.

I teleive that these problens warrant some very serione concern by the I;hC as,to the technical adequacy cf Itone & 'nebster's performance en the F.ivee Bend project.

Sincerely, L

Arthur E. Getr.c v,

r., I rinciple Frice,1;orvelle and >tney e

GULF STATES UTILITIES COMPANY RsvER SEND STATION POST OFFict 80x 220 ST FRANCl$ VILLI.LOviSweA 70775 AREA CODE SO4 635-GON 344 test August 4, 1986 RBG-24147 File Nos. G9.5, 244.700 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.

S. Nuclear Regulatory Commission Washington, D. C.

20555

Dear Mr. Denton:

River Bend Station-Unit 1 Docket No. 50-458 Attached are proposed revisions to the River Bend Station Operating License as referenced by your ietter to Mr.

W.

J.

Cahill, Jr. dated July 16, 1986 (Safety Evaluation Report).

The operating license (NPF-47) for River Bend Station (RBS),

Attachment 3,

Item 7, specifies that operation beyond the first refueling outage is subject to NRC Staff approval based on the Staff's final review of the (TDI Diesel Generator) Owners Group generic findings and the overall design review and quality revalidation program at River Bend (for the TDI diesel generators).

Gulf States Utilities proposes to implement the recommendations of Revision 2 of Appendix II of the TDI Diesel Generator Owners Group Design Review and Quality Revalidation (DRQR) Report (submitted May 1, 1986).

The change being requested herein concerns the requirement of to NPF-47, "TDI Diesel Engines Requirements".

GSU agreed to implement the Maintenance and Surveillance (M/S) program identified in Appendix II of the River Bend DROR

Report, Revision 1 by letter dated May 17, 1985.

Appendix II of the DRQR report presents a

schedule of maintenance and surveillance i

procedures recommended by the owners groups for implementation at River Bend.

The NRC Staff's review of this program is documented in Supplement 3 to the Safety Evaluation Report.

The proposed change involves no significant hazards consideration.

The proposed change does not involve a

l significant increase in the probability or consequences of an i

accident previously evaluated because the TDI Owners Group DRQR l

Report requires inspections that are more thorough than the manufacturers recommendations.

GSU's commitment to the DRQR I

Report is designed to increase reliability of the Division I and II diesel generators.

The proposed c ar. e does not create the 0l su.cm on27e 860004 0

4 g( $YS ADOCK 0500 B

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i Mr. Denton 2-August 4, 1986 possibility of a

new or different kind of accident from any accident previously evaluated because the change provides a more accurate reference to existing commitments that are presently being adhered to.

The proposed change does not involve a

significant reduction in the margin and safety and further establishes the engine design adequacy and manufacturing of the TDI diesel generators for nuclear standby service as required by General Design Criteria (GDC) 17 and in accordance with

10CFR50, Appendix B.

GSU therefore requests that the requirements of Attachment 3 to NPF-47 and Technical-Specification 3/4 8.1.1 be revised to reflect the recommendations of Revision 2.

In accordance with 10CFR170.21 also find enclosed a

check for

$150.00 for the application fee.

Pursuant to 10CFR50. 91 (b) (1),

the State of Louisiana, Department of Environmental Quality -

Nuclear Energy Division has been notified of this amendment request.

Additionally the Regional Administrator of Region IV and the Senior Resident Inspector have been provided copies of this application.

Your prompt attention to this application is appreciated.

Sincere y,

-e i

. C.

Deddens i

Vice President

,24 g River Bend Nuclear Group JCD/ER@JEP/j e t

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UNITED STATES OF AERICA NUCLEAR REGULATORY C(MilSSION STATE OF Lt?JISIANA

)

PARISH OF WEST FELICIANA

)

In the Matter of

)

Docket No. 50-458 50-459 GULF STATES UTILITIES COMPANY

)

(RiverBendStation, Unit 1)

AFFIDAVIT J.

C.

Deddens, being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

/

v J.[.Deddens i

Subscribed and sworn to before me, a Notary Public in and for the State and Parish above named, this 4

day of Avo os4-

,19 se.

s fL.

. Middrebrpoks l

an i

otary Public in and for j

West Feliciana Parish, l

Louisiana My Commission is for Life.

ATTACHMENT 3 TO-NPF 47 TDI DIESEL ENGINES REQUIREMENTS GSU shall comply with the following requirements related to the TDI diesel engines.

1.

Changes to the maintenance and surveillance program for the TDI diesel engines, as identified in (

),

shall be l

subject to the provisions of 10 CFR 50.59.

The frequency of the major engine overhauls referred to in the license conditions below shall be consistent with Section IV.1.

" Overhaul Frequency" in revision 2 of Appendix II of the Design Review / Quality Revalidation report which was transmitted by letter dated May 1,

1986, from J.

George, Owners Group, to H. Denton, NRC.

2.

Crankshafts shall be inspected as follows:

SD 1B: During the first refueling

outage, inspect the fillets and oil holes of the three most heavily loaded crankpin journals (Nos.

5, 6,

and 7) with florescent liquid penetrant and ET as appropriate.

SD 1A and IB: During the second and third refueling l

outages, inspect the fillets and oil holes of two of the three most heavily loaded crankpin journals in the manner just mentioned.

SD 1A and IB: During each major engine overhaul, inspect the fillets and oil holes of the two main bearing journals between crankpin Nos.

5, 6,

and 7,

using fluorescent liquid penetrant and ET as appropriate.

This inspection is in addition to the crankpin inspections.

3.

Cylinder blocks shall be inspected for " ligament" cracks,

" stud-to-stud" cracks and " stud-to-end" cracks as defined in a

report

by Failure Analysis Associates, Inc.

(Fa AA),

entitled, " Design Review of TDI R-4 and RV-4 Series Emergency Diesel Generator Cylinder Blocks" (TaAA report no.

Fa AA-8 4 11.1) and dated December 1984.

(Note that the FaAA This report was transmitted to H.

Denton, NRC, from C.

L.

Ray, Jr.,

TDI Owners Group, by letter dated December 11, 1984.

report specifies additional inspections to be performed for blocks with "known" or

" assumed" ligament cracks).

The inspection intervals (i.e.,

frequency) shall not exceed the intervals calculated using the cumulative damage index model in the subject FaAA report.

In addition, inspection method shall be consistent with or equivalent to those identified in the subject FaAA report.

In addition to inspections specified in the aforementioned FaAA

report, blocks with "known" or

" assumed ligament cracks" (as defined in the FaAA report) should be inspected at each refueling outage to determine whether or not cracks have initiated on the top surface exposed by the removal of two or more cylinder heads.

This process should be repeated over several refueling outages entil the entire block top has been inspected.

Liquid penetrent testing or a similarly sensitive nondestructive testing technique should be used to detect cracking, and eddy current should be used as appropriate to determine the depth of any cracks discovered.

4.

The following air roll test shall be performed as specified

below, except when the plant is already in an Action Statement of Technical Specification 3/4.8.1, " Electric Power Systems, A.C. Sources":

The engines shall be rolled over with the airstart system and with the cylinder stopcocks open prior to each planned start, unless that start occurs within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of a shutdown.

The engines shall also be rolled over with the airstart system and with the cylinder stopcocks open after 4

hours, but no more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after engine shutdown and the rolled over once again approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each shutdown.

(In the event an engine is removed from service for any reason other than the rolling over procedure prior to expiration of the 8-hour or 24-hour periods noted above, that engine need not be rolled over while it is out of service.

The licensee shall air roll the engine over with the stopcocks open at the time it is returned to service.)

The origin of any water detected in the cylinder must be determined and any cylinder head which leaks due to a crack shall be replaced.

The above air roll test may be discontinued following the first refueling outage subject to the following conditions:

All cylinder heads are Group III heads (i.e.,

cast o

after September 1980.

o Quality revalidation inspections, as identified in the Design Review / Quality Revalidation report, have been completed for all cylinder heads.

o Group III heads continue to demonstrate leak free per formance.

This should be confirmed with TDI prior to deleting air roll tests.

5.

If inspection of diesel generators lA and/or 1B should reveal cracks in the crankshaft of in the cylinder block between stud holes of adjacent cylinders, this condition shall be reported promptly to the NRC staff and the affected engine (s) shall be considered inoperable.

The engine (s) shall not be restored to " operable" status until the proposed disposition and/or corrective actions have been approved by the NRC staff.

6.

Thefollowingactionsarerequj{pdif SD 1A or SD 1B is operated in excess of 3130 KW a)

For indicated engine loads in the range of 3130 ggy to 3200 KW for a

period less than two hours

, no additional action shall be required.

b)

For indicated engine loads in the range of 3130 KW to

}3p0 KW for a

period equal to or exceeding two hours

, a crankshaft inspection pursuant to Item d

below shall be performed at the next refueling outage.

c)

For indicated engine loads in the rjgye of 3200 KW to 3500 KW for a period less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

,a crankshaft inspection pursuant to item d below shall be performed for the affected engine at the next refueling outage.

d)

For indicated engine loads in the range of 3200 3500 KW for periods equal to or exceeding one hourKW (2f, and for engine loads exceeding 3500 KW for any period of

time, (1) the engine shall be removed from service as soon as safely

possible, (2) the engine shall be declared inoperable, and (3) the crankshaft shall be inspected.

The crankshaft inspection shall include crankpin journal numbers 5, 6, and 7 (the most heavily loaded) and the two main journals in between using florescent liquid penetrant and eddy current as appropriate.

7.

Periodic inspections of the turbochargers shall include the following:

o The turbocharger thrust bearings should be visually inspected for excessive wear after 40 non-prelubed starts since the previous visual inspection.

(1) Momentary transients (not exceeding 5

seconds) due to changing of bus loads need not be considered as an overload.

(2) If there are multiple overload events within a given load range since the previous crankshaft inspection, then the time period criterion applies to the total accumulated time in that load range.

m Turbocharger rotor axial clearance should be measured at o

each refueling outage to verify compliance with TDI/Elliott specifications.

In addition, thrust bearing measurements should be compared with measurements taken previously to determine a need for further inspection or corrective action.

o Spectrographic and ferrographic engine oil analysis shall be performed quarterly to provide early evidence of bearing degradation.

Particular attention should be paid to copper level and particulate size which could signify thrust bearing degradation.

o The nozzle ring components and inlet guide vanes should be visually inspected at each refueling outage for missing parts or parts showing distress.

If such are noted, the entire ring assembly should be replaced.

a ELECTRICAL POWER SYSTEMS SWVEILLANCE REQUIREMENTS (Continued) 3.

By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-81 are met when tested in accordance with ASTM 0975-81, except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

At least once every 31 days by obtaining a sample of fuel oil from e.

the storage tanks in accordance with ASTM D2276-78 and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTA D2276-78, Method A.

As**

f.

At least once per 18 months 4, during shutdown, by:

1.

Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2.

Verifying the diesel generator capability to reject a load of greater than or equal to 917.5 kw for diesel generator IA, greater than or equal to 509.2 kw for diesel generator 18, and greater than or equal to 1995 kw for diesel generator 1C whfie maintaining engine speed less than nominal plus 75% of the dif-ference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is less.

3.

Verifying the diesel generator capability to reject a load of 3030-3130 kw*** for diesel generators 1A and 18 and 2500-2600 kw***

for diesel generator 1C without tripping.

The generator voltage shall not exceed 4784 volts for diesel generator 1A and 18 and 5400 volts for diesel generator 1C during and following the load rejection.

4.

Simulating a loss of offsite power by itself, and:

a)

For divisions I and II:

1)

Verifying deenergization of the emergency busses and load shedding from the emergency busses.

w 60804 PDR ADOCK 05000458 P

PDR

For any start of a diesel, the diesel must be operated with a load in accordance with the manufacturer's recommendations.

- - Momentary transients due to changing bus loads shall not invalidate the test.

uns E*xc E4PT rot

daeke! genera b rs 14 and 16 Wbalb Ort do be. Per[oeM avery refueh nq ou+.3e.

RIVER BEND - UNIT 1 3/4 8-6

e.

O

's In Reply Refer To:

Docket: 50-458 AUG 121906 Culf States Utilities ATTN: William J. Cahill, Jr.

Senior Vice President River Bend Nuclear Group P. O. Box 220 St. Francieville Louisiana 70775 Centlemen This letter responds to your letter, serial RBC 24143, dated August 4,1986, which requested an extension to the 18-month surveillance interval of the turbine overspeed protection system specified in paragraph 3/4.3.8 of your Technical Specificatfons.

Your request for an extension of the aforementioned Technical Specification should be directed to NRR as a request for an amendment to the Technical Specifications.

Therefore, you are advised to submit an amendment request to the requirements of 10 CFR Part 50.

pursuant Sincerely, e.1,= ri a d

), g.

a..

. i Robert D. Martin Regional Administrator cc:

Culf States Utilities ArrN:

J. E. Booker, Manager-Engineering. Nuclear Fuels & Licensing P. O. Box 2951 Beaumont Texas 77704 Louisiana State University, Government Documents Department Louisiana Radiation Control Program Director bec:

(see next page)

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GULF STATES UTILITIES COMPANY mivER SEND STAnom Post OFFICE som 220 57 FRANCisvtLLE. LOVi$ LANA 70776 ARE A CODE SO4 835 8094 346 SSSI August 15, 1986 RBG-24227 File Nos. G9.5, 244.700 Mr. Harold R.

Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

River Bend Station - Unit 1 Decket No. 50-458 Gulf States Utilities (GSU) submitted by letter dated August 4, 1986 (RBG-24147) a technical specification change request for your review and concurrence.

It has recently been brought to my attention that there was a transcription error on the last page of our August 4, 1986 letter.

Please note the attached corrected page.

Sincerely, W. J. Cahill, Jr.

Senior Vice President River Bend Nuclear Group 3

WJC/ ERG /JEP/je Attachment i

nae" OgGC^-860815 80 DR~ADOCK0500g4]B g

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1

UNITED STATES OF AMERICA EUCLEAR EEGULATORT CGeiISSION STATE OF IMISIAEA 5

FARISE OF WEST FILICIAEA In the Matter of I

Docket Bos. 50-458 CULF STATES UTILITIES CCMFAIT I

(River Bend Station, Unit 1)

AFFIDAVIT W.

J.

Cahill, Jr.,

being duly sworn, states that he is a Senior Vice President of Gulf States Utilities Company that he is authorised on the part of said Company to sign and file with the Nuclear Regulatory Coussis sion the documents attached hereto and that all such documents are true and correct to the best of his knowledge, information and belief.

A W.yCahill,Jrr Subscribed and sworn to before me, a Notary Public in and for the A

State and Parish above named, this 8 day of Ipeus4

, 19 &.

W

/

/ Notary Public in and for oan W. Middlebro'oko West Feli:iana Parish, Louisiana My Copssission is for Life, i

l l

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.

By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM 0975-81, except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

At least once every 31 days by obtaining a sample of fuel oil from e.

the storage tanks in accordance with ASTM 02276-78 and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A.

f.

At least once per 18 months ring shutdown, by:

l 1.

Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2.

Verifying the diesel generator capability to reject a load of greater than or equal to 917.5 kw for diesel generator 1A, greater than or equal to 509.2 kw for diesel generator 18, and greater than or equal to 1995 kw for diesel generator 1C while maintaining engine speed less than nominal plus 75% of the dif-ference between nominal speed and the overspeed trip setpoint i

or 15% above nominal, whichever is less.

3.

Verifying the diesel generator capability to reject a load of i

3030-3130 kw*** for diesel generators 1A and 18 and 2500-2600 kw***

for diesel generator 1C without tripping.

The generator voltage shall not cxceed 4784 volts for diesel generator IA and 18 and 5400 volts for diesel generator IC during and following the load rejection.

4.

Simulating a loss of offsite power by itself, and:

a)

For divisions I and II:

1)

Verifying deenergization of the emergency busses and load shedding from the emergency busses.

$N 8

P

For any start of a diesel, the diesel must be operated with a load in accordance with the manufacturer's recommendations.

I

- - Momentary transients due to changing bus loads shall not invalidate the test.

W sGeset 4.8.1.l. Z.f. I lo bt Pc'4ormed every refuelm3 o(darje. 3 for-basef Seaer=4*r3 iA and i e only.

RIVER BEND - UNIT 1 3/4 8-6

t 6

GULF STATES UTILITIES COMPANY ArvfF SENO STAflON Post OFFICE SQX 220 St 8RANCISWit LOvesaANA M775 AntA CODE $04 835 6064 344 8651 August 29, 1986 RBG-24285 File No. G9.5 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Dear Mr. Denton River Bend Station - Unit 1 Docket No. 50-458 Gulf States Utilities (GSU) hereby files an application for an amendment to the River Bend Station Unit 1

Technical Specifications, Appendix A to Facility Operating License NPF-47, pursuant to 10CFR50.90. contains the information for Staff review of this request. provides the requested revisions to the River Bend Station Technical Specifications.

Pursuant to 10CFR170.12, GSU has enclosed a

check in the amount of one-hundred and fifty dollars ($150.00) for the license amendment application fee.

Pursuant to 10CFR50.91 (b)

(1),

the State of Louisiana, Department of Environmental Quality - Nuclear Energy Division has been notified of this amendment request.

Additionally the Regicnal Administrator of Region IV and the Senior Resident Inspector have been provided copics of this application.

Your prompt attention to this application is appreciated.

J.C.

Deddens Vice President River Bend Nuclear Group Enclosures j

JCD/JEB/E J

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-4NyeM434G4e-660029 J

ai PDR ADOCK 05000458 O

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(

UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ptilSSION STATE OF LOUISIANA

)

PARISH OF WEST FEl.!CIANA

)

In the Matter of

)

Docket No. 50-458 50-459 GULF STATES UTILITIES COMPANY

)

(River Bend Station, Unit 1)

AFFIDAVIT J.

C.

Deddens, being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, infonnation and belief.

/

Y/w J.#IDeddens C

Subscribed and sworn to before me, a Notar Pub ic in andfortheStateandParishabovenamed,this$y9 ay of NUGMSf'

, 19b.

.J hlte WW n W. Middlebr'ooks otary Public in and for West Feliciana Parish, Louisiana My Commission is for Life.

ENCLO6URE 1 I.

PROPOSED REVISION Technical Specification 3.5.3, Suppression

Pool, establishes the Limiting Conditions for Operation for operability of the suppression pool.

This change request adds the Suppression Pool Pumpback System (SPPS) to Technical Specification 3.5.3 to ensure it is considered as required equipment for suppression pool operability.

II.

JUSTIFICATION The Suppression Pool Pumpback System (SPPS) is designed to assist in the identification and control of post-LOCA leakage from ECCS piping in the auxiliary building crescent area at el 70 ft 0 inches.

The SPTS maintains suppression pool inventory while leak detection and isolation efforts are in progress.

The SPPS also protects ECCS electrical components in this ares from flooding from any other source of leakage following a LOCA.

During the development of the Technical Specifications for the full power license, GSU committed in a letter dated November 18, 1985 (RBG-22622) to include the SPPS as part of the River Bend Technical Specifications to clarify that SPPS is a

necessary subsystem to ensure operability of the suppression pool.

The NRC staff requested that we include limiting conditions for operation, surveillance requirements, and bases.

The application for amendment is to satisfy the GSU commitment to include the SPPS as part of the River Bend Technical Specifications.

III.

SIGNIFICANT HAZARDS CONSIDERATIONS A.

The proposed change does not include a significant increase in the probability or consequences of an accident previously evaluated because the change only clarifies the intent that the SPPS is a necessary subsystem to ensure operability of the suppression pool where as these relationships were previously considered required subsystems.

This change does not involve a design change or physical change to the plant.

Thus, there is no increase in the probability or consequences of any accident previously evaluated.

B.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because this change only provides explicit clarity that the SPPS is an integral part of suppression pool operability.

This change does not involve a

design change or physical change with respect to new or modified equipment, or does it involve a change in the mode of operating existing equipment.

Thus, no new accident scenario is introduce d by this clarification of requirements for suppression pool operability.

C.

The proposed change does not involve a significant reduction in the margin of safety because this clarification of requirements for suppression pool operability significantly reduces the possibility of not considering SPPS as part of suppression pool operability, which would enhance safety rather than reduce the margin of safety.

4 b

ENCLOSURE 2 PROPOSED REVISED RIVER BEND TECHNICAL SPECIFICATIONS

EMRGENCY CORE COOLING SYSTEMS o

3/4.5.3 SUPPRESSION POOL LIMITING CONDITION FOR OPERATION 3.5.3 The suppression pool shall be OPERABLE:

)

a.

In OPERATIONAL CONDITION 1, 2 and 3 with a contained water volume of at least 137,571 ft3, equivalent to a level of 19'6".

b.

In OPERATIONAL CONDITION 4 and 5* with a contained water volume of 3

at least 94,000 ft, equivalent to a level of 13'3", except that thd suppression pool level may be less than the limit or may be drained provided that:

1.

No operations are performed that have a potential for draining the reactor vessel, 2.

The reactor mode switch is locked in the Shutdown or Refuel

position, 3.

The condensate storage tank contains at least 125,000 avail-able gallons of water, equivalent to a level of 11'1", and 4.

The HPCS system is OPERABLE per Specification 3.5.2 with an I

OPERA 8LE flow path capable of taking suction from the conden-sate storage tank and transferring the water through the spray sparger to the reactor vessel.

C.

Sea. Arrawen Tussar l'

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2 or 3 with the suppression pool water level less than the above limit, restore the water level to within j

the limit within I hour or be in at least HOT SHUTOOWN within the j

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

In OPERATIONAL CONDITION 4 or 5* with the suppression pool water level less than the above limit or drained and the above required conditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode switch in the Shutdown position.

Establish PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.

Sr.Atracws L uar "The suppression pool is not required to be OPERABLE in OPEPATIr'lAL CONDITION 5

rovided that the reactor vessel head is removed, the cavity is flooded, the i

upper containment fuel pool gate is open, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

N bbCI!So!!58 P

PDR RIVER BENO - UNIT 1 3/4 5-8

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression pool shall be determined OPERABLE by verifying the water level to be greater than or equal to, as app'.icable:

a.

19'6", at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , in OPERATIONAL CONDITION 1, 2 and 3.

b.

13'3", at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , in OPERATIONAL CONDITION 4 and 5.

4.5.3.2 With the suppression pool level less than the above limit or drained in OPERATIONAL CONDITION 4 or 5*, at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

a.

Verify the required conditions of Specification 3.5.3.b to be satisfied, or b.

Verify footnote conditions

to be satisfied.

4. 5.3. 3 Set AnAcuto ruaar l

The suppression pool is not required to be OPERABLE in OPERATIONAL CONDITION 5 eq provided that the reactor vessel head is removed, the cavity is ficoded, the V

upper containment fuel pool gate is open, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

RIVER BEND - UNIT 1 3/4 5-9 l

+

INSERT LIMITING CONDITION FOR OPERATION 3.5.3 c.

With an OPERABLE suppression pool pumpback system (SPPS) consisting of:

1.

At least one OPERABLE crescent area sump pump and 2.

An OPERABLE flow path to the suppression pool.

ACTION:

c.

With the SPPS inoperable, restore SPPS to OPERABLE status within 7 days or:

1.

In OPERATIONAL CONDITION 1, 2, or 3 be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDONN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In OPERATIONAL CONDITION 4 or 5* provide at least one alternate pumpback method and demonstrate the OPERABILITY of at least one alternate method at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, otherwise suspend CORE ALTERATIONS and all operations that have a

potential for draining the reactor vessel and lock the reactor mode switch in the shutdown position.

Establish PRIMARY CONTAINMENT INTEGRITY

~ FUEL HANDLING within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.3.3 At least once per 92

days, the SPPS shall be demonstrated OPERABLE by a.

Verifying each crescent area sump pump develops 50gpm, and b.

Verifying the flow path can be remote-manual aligned to the suppression pool.

t 3/4.5 EMERGENCY CORE COOLING SYSTEM BASES SUPPRESSION P0OL (Continued)

Repair work might r:cuire making the suppression pool inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression pool must be made inoperable, including draining. in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CONDITIONS 4 and 5 the suppression pool minimum required water volume is reduced because the reactor coolant is maintained at or below 200*F.

Since pressure suppression is not required below 212'F the minimum required water volume is based on NPSH, recirculation volume, vortex preven-tion, and a 2' 6" safety margin for conservatism.

l

\\

RIVER BEND - UNIT 1 B 3/4 5-3

INSERT for Page 83/4 5-3 The Suppression Pool Pumpback System (SPPS) is a necessary subsystem to ensure suppression pool level and therefore the OPERABILITY of the Suppression Pool can be maintained in the event of a passive ECCS failure.

The ECCS piping components may experience passive failures such as the suction valve packing failing.

The system design basis is for a maximum unisolatable leak into the auxiliary building crescent room of 50 gpm.

The manually operated SPPS consists of two crescent room sumps each with two 100% capacity pumps.

Each pump is capable of pumping 65 gpm and is sufficient to ensure OPERABILITY of the Suppression Pool.

e f

GULF STATES UTE.ITIES COMPANY 9

BEAUMONT, TEXAS (713) 838-6631 h

oats cwCx Not av T>eS MOUNT 9-4-86 700787

150.00**

my To ne ORoER oF A s s

m U.S. Nuclear Regulatory COnunissic, Office of Nuclear Reactor Regulation f

Washington, D.C.

20555 L_

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"'*"*b crry om sam 4TuRE REaumEO ssooo on LEss oF SEAUMONT.NA-M AFTER 60 DAYS eEAuMONT.Trx4S L

e ?OO 7B ?e :: L L 3100 21.Oi:

O G= 51 857 Le CETACH SEFORE DEPOSITDeG e

T>eS VOUCER IS Dd FULL SETTLEMENT OF ITEMS USTED SELOW GULF STATES UTUTES COMMNy l

l DATE TERMS DAOCE NUMBER OR EMPLANATION OfSCOUNT ET 9-4-86 070017 Application Fee - Facility Operating License Change Request #86-018 150.00 P

Wjwjwjwg jwjwjwjdi O

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GULF STATES UTELETIES COMPANY arvin st%D Station pos+ osfiCf 90 220 SF taANCISviLLI LoutSiAha 70776 an. coo so.

en.o

u...si September 10, 1986 RBG-24358 File Nos. G9.5, G9.8.6.2 Mr. Harold R. Donton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

River Bend Station - Unit 1 Docket No. 50-458 Enclosed for your review is Revision 2 to the River Bend Station Pump and Valve Inservice Testing (IST)

Plan.

This revision contains two minor changes to the Pump and Valve Program.

ITEM 1:

(page 9 of 14):

The Technical Specification limit for discharge pressure for Reactor Core Isolation Cooling System (RCIC) pump 1E51*PC001 was added.

This had been inadvertently omitted from the Appendix A Table.

ITEM 2:

(page 14 of 14 and Relief Request Nos. 8 & 9):

The Suppression Pool Pumpback System pumps 1DFR*PSA, B,

D E

were deleted.

These pumps are Class 4 pumps and do not fall within the defined scope of ASME Section XI, IWP-1100.

The ASME code provides

...the rules and requirements for inservice testing of Class 1,

2, and 3

centrifugal and displacement type pumps that are installed in light water cooled nuclear power plants..."

Although there is no specific IST requirements for these

pumps, Gulf States Utilities (GSU) has filed an application for amendment to the j

River Bend Station Unit 1 Facility Operating License NPF-47 (letter form J.C. Deddens to H.R.

Denton dated August 29, 1986; RBG-24285) to include testing of these pumps as part of the River Bend Station Technical Specifications surveillance testing program.

Pursuant to 10CFR170.21, GSU has enclosed a check in the amount of one-hundred and fifty dollars

($150. 00) for the license application fee.

Sincerely,

[*

4 b 9

Si" 140^20:060910

$DR ADOCK 0500 8

. E. Booker Manager-Engineering, Nuclear Fuels & Licensing gg River Bend Nuclear Group ho[

JED/ ERG /JEP/je gMcf whgg f /SCO Enclosure

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O HOLDER #

AUG 281966 PUMP REQUEST FOR RELIEF NO. S w

n..,

, si.ie

'Do c u.m o~~***

SYSTEM

. COMPONENT CLASS FUNCTI 4et.E & 8 12 D/G-Fuel Oil IEGF* PIA,B,C 3

Fuel oil transfer from the storage tank to the day tank Standby ISWP*P2A,B,C,D 3

Service Water Supply emergency cooling water to various safety related plant heat loads TEST REQUIREMENT:

Per IWP-3100 and Table IWP-3100-1, Note (1),

inlet pressure shall be measured before pump startup and during test.

BASIS FOR RELIEF:

The suction section of the pumps are submerged in the pumping fluid.

There is no method to measure the inlet pressure before and after pump startup.

In all cases, the fluid level above the pumps will remain unchanged during the test.

ALTERNATE TESTING:

Calculate the pump inlet pressure from the hydrostatic head of fluid above the pump inlet at the time of test.

l l

l

sM'W %

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0 Hott*R #-

AUG 88 1946

%, e...

.... [

c.ai..n.4 om...i 0

PUMP REQUEST FOR RELIEF NO. 9 ay a g6 SYSTEM COMPONENT

_ CLASS FUNCTION I2 D/G-Fuel IEGF* PIA,B,C 3

Fuel oil transfer oil from the storage tank to the day tank TEST REQUIREMENT:

Per IWP-4510 and Relief Request No. 5, at least one vibration amplitude shall be read during each inservice test.

BASIS FOR RELIEF:

The pumps are deeply submerged in the pumped fluid and are inaccessable to measure the vibration.

ALTERNATE TESTING:

Vibration will not be measured.

1 p

I

Cs TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

1-1 2.0 INSERVICE TESTING OF PUMPS 2.1 General Information 2-1 2.2 Plan Information 2-1 2.3 Requests for Relief 2-3 3.0 INSERVICE TESTING OF VALVES 3.1 General Information 3-1 3.2 Preservice Testing 3-1 3.3 Plan Information 1-1 3.4 Requests for Relief 3-5

~

Appendix A Pump Inservice Testing Plan Appendix B Pump Instervice Testing Plan Requests for Relief Appendix C Valve Inservice Testing Plan Appendix D Valve Inservice Testing Plan Requests for Relief AUG 19 548

""S4[ W/

3 LE & RC

)

n.

1.0 INTRODUCTICIE The Inservice Inspection Plan for River Bend Station pumps and valves was developed in compliance with 10CFR50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition thru Winter 1981 Addenda.

Where compliance was determined to be impractical, specific relief has been requested.

Section 2.0 discusses the Inservice Testing Plan for applicable Class 1,

2, and 3 pumps.

Section 3.0 discusses the Inservice Testing Plan for applicable Class 1, 2, and 3

valves.

fk

, p;tE.S UTq,,

b sfY

& marr %

AUG 19126 E

m.

O C**t'Otl.4 Docume

.a s sc RECEIVED JUL 0 31986 soc 1-1

C.

. 2.0 INSERVICE TESTING OF PUMPS 2.1 GENERAL INFORMATION The Inservice Testing Plan for ASME Class 1, 2, and 3 pumps was developed in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWP, 1980 Edition thru Winter 1981 Addenda.

l The Inservice Testing Plan for pumps will remain in effect through the first 10 year inservice inspection interval commencing at commercial operation.

Appendix A identifies the Class 1, 2, and 3 pumps that are

tested, along with the applicable parameters that are measured.

This listing is arranged by GSU system number, then by P&ID number and finally by pump number.

2.2 PLAN INFORMATION The following information is provided in the computer generated Pump Inservice Testing Plan (Appendix A):

2.2.1

' SYSTEM NAME':

The name of the system which the pump is a part of.

2.2.2

'SYS NO':

The GSU number of the system which corresponds with the System Name.

2.2.3

'Pt,ID NO':

The Process and Instrument Diagram number on which the listed pumps are shown.

2.2.4

' PUMP INFORMATION':

1.

' MARK NUMBER':

This is the pump identification number.

2.

' DESIGN AND DESCRIPTION':

Listed are important design characteristics and descriptive information for the pump. general 3.

' NORMAL STATUS' STDBY - Standby l

RUN'NG - Running pts urg, p

M 6

g RECEIVED 4

HOLDER #

Aug 19 isas JUL 0 31986 w

sDc

%g.............

2-1

I

. 3.0 INSERVICE TESTING OF VAINES 3.1 GENERAL INFORMATION i

The Inservice Testing Plan for ASME Class 1, 2,

and 3

valves was developed in accordance with the requirements of ASME Boiler and Pressure Vessel

Code,Section XI, Subsection IWV, 1980 Edition thru Winter Addenda.

The Inservice Testing Plan for valves will remain in effeet through the first 10 year inservice inspection interval commencing at commercial operation.

Appendix C identifies the Class 1, 2, and 3 valves that are tested.

This listing is arranged by GSU system number, then by P&ID number and finally by valve number.

3.2 PRESERVICE TESTING Each valve was tested after installation and prior to service.

These tests were conducted under conditions similar to those to be experienced during subsequent inservice tests.

Safety and relief valves which will be removed and bench tested during subsequent inservice i

tests were not necessarily installed prior to their

(~

preservice test.

A 3

3.3 PLAN INFORMATION The following information is provided in the computer generated Valve Inservice Testing Plan (Appendix C) :

3.3.1

' SYSTEM NAME':

The name of the system which the valve is a part of.

3.3.2

'SYS NO':

The GSU number of the system which corresponds with the System Name.

3.3.3

'P&ID NO's The Process and Instrument Diagram number on which the listed valves are shown.

3.3.4

' VALVE INFORMATION':

1 1.

' MARK NUMBER':

This is the valve identification number.

(A few dampers are listed which are required to be tested

[s[6S "'l/rfg pursuant to the ASME Code Section XI,

_o Subsection IWV by Tech Spec 4.6.5.3.C even HOLDER #

though they do not fall within the scope of IWV itself.)

AUg Ig $86 RECEIVED

% 4~~J!M'O [

~

JUL 0 31986 detc a a SDC 3-1

=

APPENDIX B PUMP INSERVICE TESTING PLAN REQUESTS FOR RELIEF f*

s t

gxbTES Ultg Yg

o MG 191966

+r l

g.*

RECEIVED JUL 0 31986 sac

4.

a

,. ~

~

0 0

APPENDIX C VALVE INSERVICE TESTING PLAN I

(

gts urg, 4

0 2

O HOLDER #

9 AUG 19 IM6 O

""'8D**"'"*

O

[

BLE & FL

/~

RECEIVED JUL 0 31986 SDC

m I

a_

m-e APPENDIX D VALVE INSERVICE TESTING PLAN REQUESTS FOR RELIEF D-9

/pr.s urg,1*

5 h ~ HOLDER #

AUG 191986

% c". L '*!*d'.* O OLE s

s 4m RECEIVED.

JUL 0 31986 SDC

O GUEF STATES UTILITIES COMPANY mivEn SE ND STATION POST OFrict Box 220 St FRANCisviLLE Louis 4AAA 70775 AREA CODE 504 835 6094 346 8661 September 12, 1986 RBG-24383 File No. G9.5 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

River Bend Station - Unit 1 Docket No. 50-458 Prior to the issuance of the River Bend Station operating license, Gulf States Utilities (GSU) was requested by your Staff to evaluate the concept of concurrently removing the fire protection elements from the Technical Specifications and incorporating these elements into the Final Safety Analysis Report (FSAR).

This will, in effect, transfer fire protection program commitments, reporting requirements and amendments from the jurisdiction of 10CFR50.73 and 10CFR50.90 to 10CFR50.59 and 10CFR50. 71 (e).

The River Bend Station (RBS) Fire Protection Program will be completely described and controlled through the FSAR rather than through the combination of the FSAR and Technical Specification.

This material is consistant with the recommendations of the Committee to Review Generic Requirements (CRGR) as addressed in Generic Letter 86-10 and previously approved on the Perry Nuclear Tower Plant docket.

The Fire Protection Program currently in place at RBS meets the requirements of existing Technical Speciations.

With relatively straightforward administrative changes to the FSAR and plant procedures, the Program's commitments will be defined within FSAR Section 9.5 and Appendix 9A. provides a description of the proposed Technical Specification change.

In

addition, FSAR Section 9.5.1.4 will be amended by incorporating a

discussion of fire protection operations, testing and surveillances (Enclosure 2).

As shown in this enclosure, information provided addresses plant administrative procedures, periodic testing and remedial actions for inoperability.

There is no significant increase in the probability or the consequences of an accident previously evaluated resulting from this change because of the administrative concept of concurrently removing the fire protection elements from the Technical Specifications and incorporating these elements into the FSAR.

h(gkk I]

-86099999945660912 00(k PDR ADOCK 05000458 P

PDR

/

~

i Mr. Denton September 12, 1986 This change would not create the possibility of a

new or different kind of accident from any accident previously evaluated because no accident analysis are affected by the change.

Finally, this change would not involve a significant reduction in the margin of sa fety because the incorporation of the fire protection elements into the FSAR does not decrease the level of fire protection in the plant.

In accordance with 10CFR170.21 also find enclosed a

check for S150.00 for the application fee.

Pursuant to 10CFR50.91(b) (1),

the State of Louisiana, Department of Environmental Quality Nuclear Energy Division has been notified of this amendment request.

Additionally, the Regional Administrator of Region IV and the Senior Resident Inspector have been provided copies of this application.

Your prompt attention to this application is appreciated.

S' cerel,

6 E.

Deddens Vice President River Bend Nuclear Group JCD/6EB/ ERG /JEP/je

't l

l

UNITED STATES OF AMERICA NUCLEAR REGULATORY C0fttISSION STATE OF LOUISIANA

)

PARISH OF WEST FELICIANA

)

In the Matter of

)

Docket No. 50-458 50-459 GULF STATES UTILITIES COMPANY

)

(River Bend Station, Unit 1)

AFFIDAVIT J.

C.

Deddens, being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

J. b eddens Subscribed and sworn to before me, a Notary Public in and for the State and Parish above named, this /Shdayof

&nlemher

,192&_.

r

. Afft can W. MiddTebr(oks Notary Public in and for West Feliciana Parish, Louisiana My Commission is for Life.

-w 0

O ENCLOSURE 1

e O

GULF STATES UTILITIES COMPANY RivtR SEND ST ATION POST OFFICE 80x 220 ST FaANC15vtLLE. LouisiAhA Mnl AntA CODE 504 835 6094 346 8651 September 12, 1986 i

RBG-24385 File No. G9.5 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

River Bend Station - Unit 1 Docket No. 50-458 Gulf States Utilities (GSU) hereby files an application for an amendment to the River Bend Station Unit 1

Technical Specifications, Appendix A to Facility Operating License NPF-47, pursuant to 10CFR50.90. contains the information for Staff review of this requestion.

Enclosure 2

provides the requested revisions to the River Bend Station Technical Specifications.

Pursuant to 10CFR170.12, GSU has enclosed a check in the amount of one-hundred and fifty dollars

($150. 00) for the license amendment application fee.

Pursuant to 10CFR50. 91 (b) (1), the State of Lousiana, Department of Environemtnal Quality -

Nuclear Energy Division has been notified of this amendment request.

Additionally the Regional Administrator of Region IV and the Senior Resident Inspector have been provided copies of this application.

Your prompt attention to this application is appreciated.

Sincerely, V

l

. C. Deddens Vice President River Bend Nuclear Group l

Enclosures JCD/

/E

/J

/je g(0 (9,},h[ipoi o\\

9 o

l

-60m i

PDR ADOCK 05000458 1

l p

PDR

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION STATE OF LOUISIANA

)

PARISH OF WEST FELICIANA

)

In the Matter of

)

Docket No. 50-458 50-459 GULF STATES UTILITIES COMPANY

)

(River Bend Station, Unit 1)

AFFIDAVIT J.

C.

Deddens, being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to

~ ~

the best of his knowledge, infonnation and belief.

A

//

~

nw J.IC. Deddens 1

Subscribed and sworn to before me, a Notary Public lin and for the State and Parish above named, this /2 " day.cf Seolem b e,-

, 19 96.

fttsu b

4 an W. Middleyrooks otary Public in and for West Feliciana Parish, Louisiana My Commission is for Life.

~,

l

ENCLOSURE 1 I.A.

Proposed Revision Technical Specification 3.3.2, Isolation Actuation Instrumentation, Table 3.3.2-2, Isolation Actuation Instrumentation Setpoint, Item 2.d, Main Steam Line Flow

High, identifies a

trip setpoint of f 173 psid and an allowable value of &178 psid.

This trip setpoint and allowable value also have a ** footnote which indicates that this is a preliminary setpoint and that the setpoint be confirmed or revised based on startup test data.

The

- footnote further requires that any requested revision to this setpoint be submitted to the Nuclear Regulatory Commission.

The change request revises the trip setpoint and the allowable value from one setpoint of f173 psig and 6178 psig, respectively, for all four main steam lines to one trip setpoint and allowable value for each main steam line as follows:

Trip Allowable Setpoint Values Main Steam Line Flow Line A - High 4146 psid 6151 psid Main Steam Line Flow Line B - High 4156 psid 6161 psid Main Steam Line Flow Line C - High 6153 psid 6158 psid Main Steam Line Flow Line D - High 6164 psid 6169 psid This change request would also delete the ** footnote from Item 2.d since its required action has been completed.

II.A. Justification The

- footnote indicates the present trip setpoint and allowable value for Main Steam Line Flow High as a

preliminary value.

These are setpoints based on design Cdlculations.

Confirmation of these values are provided by measured flow data taken during the startup test program.

Results from this data require the present trip setpoint and allowable value to be revised.

GSU is proposing a change to four setpoints instead of one because of the variation between Steam Line A

and Steam Line D flow measurement data.

All of these setpoints are within the design basis specification and are conservative with respect to the assumptions of the accident analysis to ensure the isolation of the Main Steam Line system in the event of a break in that system.

If only one trip setpoint was specified, the nominal trip setpoint would be the most conservative of the trip setpoints, which is 146 psig.

If 146 psig was specified for all four main steam lines, its application to Steam Line D

would severly limit the

l

~

3 s

operational flexibility of Steam Line D high flow limits by reducing \\the effective analytical leak detection limit from 140%

of,. rated flow to approximately 130% of rated flow.

Therefore, in order to maintain the necessary operational flexibility while ensuring the setpoints are within the bounds of the design basis. and the assumptions of the accident

analysis, four trip setpoints are specified.

Since the final setpoint has been determined, it is appropriate to delete the ** footnote for this Item 2.d.

III.A.Significant Hazard Considerations A.

The proposed change does not involve a

significant increase in the probability or consequences of an accident,previously evaluated because the setpoint revibions are-more conservative than the initial setpoihts arid are based on startup test data as required by the. plant Technical Specifications.

This change does not involve a design change or physical change to the plant, and therefore, does not increase the probability of a undetectable break in the Main S' team Line.

Thus, there is no. increase in the probability or consequences of any accident previously evaluated.

B.

The proposed cha.nge does not create the possibility of s

a new or differene kind of accident from any accident

\\

previously evaluated because the setpoint revisions are more' conservative than the initial setpoints and are within the bo'unds of the design basis and the N

assumption of, the accident anlysis.

These setpoint revisions do not involve a design change or physical change, and therefore, do not alter the single failure design of the instrumentation.

i Thus, no new accident scenario is introduced by these revised and more conservative setpoints.

s C.

The proposed change does not involve a significant l

l reduction in a margin of safety because these setpoint l

revisions are more conservative and reflect actual 1

-s startup test data as required by the plant Technical

/

Specifications.

basis, setpoints are within the bounds These Q

of the design

'and the assumptions of the accident analysis.

Thus, no margin of safety is reduced.

l l

i l

i

I.B.

Proposed Revision Technical Specification 3.3.2, Isolation Actuation Instrumentation, Table 3.3.2-2, Isolation Actuation Instrumentation Setpoint, Item 5.a, RCIC Steam Line Flow -

High, identified a trip setpoint of f222" H0 and an 3

allowable value of 6230.5" H O.

The trip satpoint and allowable value also have a

footnote setpoint which indicates that this is a preliminary setpoint and that the setpoint be confirmed or revised based on startup test data.

The

- footnote further requires that any requested revision to this setpoint be submitted to the Nuclear Regulatory Commission.

The change request revises the trip setpoint from 6222" H 0 to 6127" H 0 and the allowable 9

7 value from i 230.5" H o to s135.5" H 0 This change request would also delete the ** footnote fr$m. Item 5.a.

2 II.B. Justification The

- footnote indicates the present trip setpoint and allowable value for RCIC Steam Line Flow High as a

preliminary value.

These setpoints are based on design calculations.

Confirmation of these values is provided by measured flow data taken during the startup test program.

Results from this data require the present trip setpoint and allowable value to be revised.

The lowering of this setpoint will further ensure the isolation of the Reactor Core Isolation Cooling (RCIC) system in the event of a break / leak in that system.

Since the final setpoint has been determined, it is appropriate to delete the ** footnote for this Item 5.a.

III.B.Significant Hazard Considerations A.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the setpoint revisions are more conservative than the initial setpoints and are based on startup test data as required by the plant Technical Specifications.

This change does not involve a design change or physical change to the plant, and therefore, does not increase the probability of a

undetectable leak in the RCIC system.

Thus, there is no increase in the probability or consequences of any accident previously evaluated.

B.

The proposed change does not create the possibility of a

new or different kind of accident from any accident previously evaluated because the setpoint revisions are more conservative than the initial setpoints and are within the bounds of the design basis and the

i-assumptions of the accident analysis.

These setpoint revisions do not involve a design change or physical

change, and therfore, do not alter the single failure design of the instrumentation.

Thus, no new accident scenario is introduced by these revised and more conservative setpoints.

C.

The proposed change does not involve a

significant reduction in a margin of safety because these setpoint revisions are more conservative and reflect actual startup test data as required by the plant Technical Specifications.

These setpoints are within the bounds of the design basis and the assumptions of the accident analysis.

Thus, no margin of safety is reduced.

l l

l 1

I.C.

Proposed Revision Techt.ical Specification 3.3.2, Isolation Actuation Instrumentation, Table 3.3.2-2, Isolation Actuation Instrumentation Setpoint, Item 5.1, RHR/RCIC Steam Line Flow High identified a

- footnote on each the trip setpoint and allowable value.

The ** footnote specifies:

" Initial setpoint.

Final setpoint to be determined during startup test program.

Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion."

The change request revises the ** footnote to specify:

" Initial setpoint.

Final setpoint to be determined during testing prior to operation in the steam condensing mode following the Nuclear Regulatory Commission's approval to operate in that mede (Reference License Condition 5.a to NPF-47).

Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion."

~

II.C. Justification The River Bend Station Facility Operating License (NPF-4 7) prohibits the use of the residual heat removal system in the steam condensing mode without prior written approval of the Nuclear Regulatory Commission Staff.

Because these setpoints would have to be confirmed during actual operation in the steam condensing mode, GSU would request that a similar condition be maintained in the Technical Specifications.

The current setpoints are based upon design calculations.

III.C.Significant Hazard Considerations A.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the change is purely administrative in nature and reflects a

condition already established in the operating license, l

B.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the setpoints are based

[

upon design calculations which have not been changed.

i C.

The proposed change does not involve a

significant reduction in a margin of safety because the current j

setpoints are within the bounds of the design basis and t

the assumptions of the accident analysis.

t

- - - - - - ~

ENCLOSURE 2 Proposed Revised River Bend Technical Specifications

I TABLE 3.3.2-2

=

ISOLATION ACTUATION INSTRLMENTATION SETPOINTS

?i

=

ALLOWABLE g

TRIP FUNCTION TRIP SETPOINT VALUE i

E 1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

c 2

Low Low, Level 2 1-43 inches

l-47 inches w

w b.

Drywell Pressure - H10h 1 1.68 psig i 1.88 psig c.

Containment Purge Isolation Radiation - High 5 1.3 R/hr 1 1.57 R/hr I

2.

MAIN STEAM LINE ISOLATION a.

Reactor Vessel Water Level -

Low Low Low, Level 1 1-143 inches

1-147 inches b.

Main Steam Line Radiation - High 13.0 x full power background 1 3.6 x full power y

background c.

Main Steam Line Pressure - Low t 849 psig 1 837 psig l

d.

Main Steam Line Flow - High i 173,,,;d i 17^, j..id e.

Condenser Vacuum - Low 1 8.5 inches Hg.

> 7.6 inches Ng.

vacuum vacuum l

f.

Main Steam Line Tunnel Temperature - High 1 135'F i 142.5'F g.

Main Steam Line Tunnel A Temperature - High 1 51*F 1 55'F

1. l.ine. A 6 f% pJJ

& 151 esM N ERT

2. Line B d 156 nid 4161 esid.

3. Line C

153 e.'d

+ 158 esid

4. Lin e. O 6 Iroy pi.d 4 s toq g3;d.

"W899E9944% 860912 gDR ADOCK 05000y B

TABLE 3.3.2-2 (Continued) h ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

a, ALLOWABLE Q

TRIP FUNCTION TRIP SETPOINT VALUE O

4.

REACTOR WATER CLEANUP SYSTEM ISOLATION (Cont'd) d.

Equipment Area A Temperature - High E

1.

Heat Exchanger Room 5 33*F

$ 36.5'F Z

2.

Pump Rooms A and B 1 78'F 5 82*F 3.

Valve Nest Roon 1 46*F 5 49.5'F g

4.

Domineralizer Rooms 1 and 2

< 46*F

< 49.5'F 5.

Receiving Tank Roon 546*F 349.5*F e.

Reactor Vessel Water Level -

Low Low Level 2 1 - 43 inches

1 - 47 inches f.

Main Steam Line Tunnel Ambient Temperature - High 5 135'F i 142.5*F w

g.

Main Steam Line Tunnel

)

A Temperature - High 1 51*F 1 55'F Y

h.

SLCS Initiation NA NA U

5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line Flow - High i eeese H0M

$ 999r6" H 0 +

l 129 135.E b.

RCIC Steam Line Flow - High Timer 1 3 seconds 5 13 seconds c.

RCIC Steam Supply Pressure - Low 1 60 psig 1 55 psig d.

RCIC Turbine Exhaust Diaphragpn Pressure - High 1 10 psig i 20 psig e.

RCIC Equipment Room Ambient Temperature - High 1 182*F

$ 186.4*F f.

RCIC Equipment Room A Temperature -

1 96*F

$ 99'F High

TABLE 3.3.2-2 (Continued)

O ISOLATION ACTUATION INSTRUMENTATION SETPOINTS-5

[

ALLOWABLE g

TRIP FUNCTION TRIP SETPOINT VALUE o

5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Cont'd) g g.

Main Steam Line Tunnel Ambient y

Temperature - High 5 135'F i 142.5'F h.

Main Steam Line Tunnel A Temperature - High 1 51*F

$ 55'F 1.

Main Steam Line Tunnel Temperature Tider 0 seconds NA j.

RHR Equipment Room Ambient Temperature - High 1 117'F i 121.1*F R

k.

RHR Equipment Room y

a Temperature - High 5 29*F 1 33.6*F 1.

RHR/RCIC Steam Line Flow - High 5 156" H,0**

$ 164.5" H 0**

m.

Drywell Pressure - High 1 1.68 psig i 1.88 psig n.

Manual Initiation NA NA 6.

RHR SYSTEM ISOLATION a.

RHR Equipment Area Ambient Temperature - High 3 117'F

$ 121.1*F b.

RHR Equipment Area A Temperature -

High 1 29*F

$ 33.6*F c.

Reactor Vessel Water Level -

Low Level 3 1 9.7 inches

1 8.7 inches d.

Reactor Vessel Water Level -

Low Low Low Level 1

> - 143 inches *

> - 147 inches

i 1

TABLE 3.3.2-2 (Continued)

=Q ISOLATION ACTUATION INSTRUMENTATION SETPOINTS

!E 1

co ALLOWABLE G

TRIP FUNCTION TRIP SETPOINT VALUE t

6.

RHR SYSTEM ISOLATION c.

(Cont'd) 5 e.

Reactor Vessel (RHR Cut-in l

Permissive) Pressure - High 3 135 psig

< 150 psig w

j f.

Drywell Pressure - High 5 1.68 psig i 1.88 psig 7.

MANUAL INITIATION NA NA l

See Bases Figure B 3/4 3-1.

I R

- Initial setpoint.

Tinei Octpeint t; b; det mined dur'n; :t: rte; test pr:;r=. Any required change to his setpoint shall be submitted to the Commission within 90 days Y

of test completion.

E$

l Faml scfPoenf }o be defermsbed during -k"$4sn' 3 PMoe b cperal-ion en 4he sleam condens.'nq <woJe Ell-eng Me Nuclear Aegul focy Co,

,34 ' oppro.; 4 oper.4 in 4ha4 mods (Reference. Luce,e Coed,laan 5.a. 4. NPF-47).

l

\\

GULF STATES UTILITIES COMPANY river BEND STATION POST OFFICE SOX 220 ST FRANCl$VILLE. LOutSIAhA 70775 ARIA CODE 504 435 6094 346 4651 September 17, 1986 RBG-24405 File No. G9.5 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

River Bend Station - Unit 1 Docket No. 50-458 Gulf States Utilities (GSU) hereby files an application for an amendment to the River Bend Station Unit 1

Technical Specifications, Appendix A to Facility Operating License NPF-47, pursuant to 10CFR50.90. contains the information for Staff review of this request. provides the requested revisions to the River Bend Station Technical Specifications.

The revisions discussed in Enclosure 1 are required to prevent impending River Bend Station shutdown and therefore represent emergency circumstances as discussed in 10CFR50.91.

Pursuant to 10CFR170.12, GSU has enclosed a check in the amount of one-hundred and fifty dollars

($150. 00) for a

license l

amendment application fee.

Pursuant to 10CFR50.91(b) (1), the State of Louisiana, Department of Environmental Quality - Nuclear Energy Division has been notified of this amendment request.

Additionally the Regional Administrator of Region IV and the Senior Resident Inspector have been provided copies of this application.

Your prompt attention to this application is appreciated.

1

(

Sincerely, e

l l

-0007220!"i_860917 J. C. Deddens i

PDR ADOCK 05000458 Vice President P

PDR River Bend Nuclear Group Enclosures JCD/

/

P/je

[)

l00(

s l

us e m

/

'Y

UNITED STATES OF AMERICA NUCLEAR REGULATORY C0fMISSION STATE OF LOUISIANA

)

PARISH OF WEST FELICIANA

)

In the Matter of

)

Docket No. 50-458 50-459 GULF STATES UTILITIES COMPANY

)

(River Bend Station, Unit 1)

AFFIDAVIT J.

C.

Deddens, being duly sworn, states that he is a Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, infonnation and belief.

/

J. [ Deddens Subscribed and sworn to before me, a Notary Pub ic in nd for th State and Parish above named, this /

day of FM ef

, 19

/ftA

. I4

/

anW.Middlebrojks otary Public in and for West Feliciana Parish, Louisiana My Commission is for Life.

ENCLOSURE 1 I.

Proposed Revision Technical Specification 3.6.4, Table 3.6.4-1, Item a.1 identifies the RHR/RCIC Steam Supply Valve 1E51*MOVF076 as a primary containment automatic isolation valve.

The proposed change adds a

one-time footnote applicable to valve 1E51*MOVF076 to remove its requirements for operability through October 4, 1986.

On September 8, 1986 with the plant in mode 1 at 1004 power, a functional test was performed on the "RCIC WARMUP LINE ISO VLV E51-F076 NOT FULLY CLOSED" annunciator.

As the operator proceeded to stroke the valve

open, the MOV76 breaker i

tripped.

Investigation determined the valve motor to be grounded and the valve to be open.

By taking the i

i appropriate actions in the Technical Specifications the operator declared MOV76 inoperable and deactivated and secured in the isolated position the outside containment isolation valve 1E51*MOVF064.

This action renders the 4

l Reactor Core Isolation Cooling (RCIC) system inoperable and requires plant shutdown on September 22, 1986 per Technical Specifications 3/4.7.3.

The MOV76 is a'3/4" bypass line valve used for prewarming the RCIC steam line and currently will not close upon manual or automatic isolation.

In order to isolate the 1KJBZ15 penetration the outboard isolation valve (MOV64) is required to be closed.

The proposed technical specification change will temporarily eliminate the requirement that the 1KJBZ15 penetration be completely isolated and will allow the RCIC system to be functionally operable.

This request would be for the time period through October 4, 1986 at which time a shutdown is scheduled and the MOV76 valve would be repaired.

Safety Evaluation The Reactor Core Isolation Cooling (RCIC) system isolation trip setpoints are identified in Table 3.3.2-2, Item 5 of the River Bend Technical Specification.

In the event of a

significant leak from the RCIC steam supply system in the auxiliary building, the RCIC system would be automatically 1

isolated due to high ambient temperature in the RCIC l

equipment room, high RCIC steam line flow or RCIC steam supply low pressure.

These automatic isolations would initiate the closure of E51*MOV's 63, 64, and 76.

For these events, only the outboard isolation valve MOV64 needs to be closed to seal off the break.

In the event the MOV64 4

did not close and MOV63 did close, the blowdown flow rate for a complete severance of the RCIC steam line would be reduced within 10 seconds from flow through a 8" main line i

i

..._,_._~__.-,__.,,_.,m__-m__.._,.,-.-

.,,.y,,,.,,,.,,,-,~.,__..,.,_,,___._,..,_,m.__.___,._,.,_,._._.__,__m_

4 to flow through a 3/4" bypass steam line corresponding to less than 40/sec.

(i.e.

30 gpm).

Feedwater flow could i

easily provide make-up for this fluid loss.

If feedwater is not available

, the High Pressure Core Spray (HPCS) system could be used to provide cooling water to the vessel thereby preventing uncovering any reactor fuel.

There is a manually i

operated 3/4" valve, V15, in the bypass steam line in the drywell just upstream of MOV76 which, upon reactor shutdown, could be manually closed to isolate the primary containment.

Additionally, the automatic MOV64, located in the auxiliary

building, could be manually closed depending on local conditions.

A potentially more serious postulated loss-of-coolant accident requiring RCIC system isolation would be the Design t

Basis Accident where the reactor core is uncovered and fuel rod failures occur, thereby releasing fission products.

For this event the airborne fission products could be 4

transported through a closed system via the main steam line, the RCIC steam line, the RCIC bypass steam

line, the RCIC i

steam line to the RCIC turbine and back to the suppression pool.

Any leakage occurring through valve packing ~ and would be minimal and processed via the Standby Gas Treatment System.

t

However, in order for the fission products to even get out of the primary containment into the auxiliary building through the RCIC system piping it would require first that the Design Basis Accident event occur.

Based on WASH 1400 table III 4-1tggprobabilityoftheDesignBasisAccident event is 1.0x10 per hour of reactor operation.

Considering that the reactor will be operated for less than 15 days with

MOV76, open and using an additional conservatism of a

factor of 30, would result in the possibility of a Design Basig Accident event to occur during l

those 15 days to be 1.08x10 Then, for any fission products to get past MOV64, it would require that MOV64 failed to close.

BasedonIEEEStandarg L

500-1977 there would be 2.5 failures of the valve in 10 i

actuations.

Assuming a

conservative estimate of 50 isolation actuations of MOV64 per calendar year and an i

additional conservatism of a

factor of 10 resulted in a probability of the failure of MgV64 to close in the 15 day time period to be 5.14x10~.

Therefore, the overall l

probability of unisolated fission product release in i

conjunction with a Design Basis Analysis into the RCIC steam i

piping outside the primary containment would be the product 1

l of the Design Basis Accident event to occur and the MOV64 failure _ygo close during the 15 day time period or 5.55x10 This probabglity is more than an order of magnitude lower than 10 which is assumed incredible.

Therefore, leaving the MOV76 valve stuck in an open position

while isolation valve MOV63 and MOV64 are open during a

15 day period is considered to be at an acceptable level of risk for continued plant operation and the LCO for the failure of MOV76 to isolate may remove operability requirements through October 4, 1986.

No Significant Hazards Evaluation The proposed amendment to the Technical Specifications a.

would not involve a

significant increase in the probability or consequences of an accident previously evaluated because the potentially unisolated penetration is limited to a 3/4" line discharging to a closed system that returns to the containment.

Thus, there is no increase in the probability or consequences of any accident previously evaluated.

b.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because no new modes of operation are introduced and no change to the plant design has been made.

Thus, there are no new or different kinds of accident from any accident previously evaluated.

c.

The proposed change does not involve a significant reduction in the margin of safety because this event is expected to have a

probability that is non-credible.

Because the unisolated penetration is limited to a

3/4" line that discharges to a closed system, the potential increase in offsite dose is insignificant.

Thus, there is no significant reduction in the margin of safety.

II.

Revised Technical Specifications The requested revision is provided as Enclosure 2.

III. Interim Compensatory Measures River Bend Station currently meets the requirements of Technical Specifications.

Therefore, no interim compensatory measures are required.

IV.

Bases for Emergency Circumstances With the RCIC system inoperable as the result of MOV64 being

closed, River Bend Station will be required to shutdown on September 22, 1986.

The closure of the MOV64 valve

precludes the use of the RCIC system as a

means of maintaining reactor water level since the use of RCIC requires the prewarming of the RCIC steam line to prevent water hammer in that line.

To operate past September 22, 1986 under the current Technical Specifications would require the MOV76 isolation valve to be manually closed and MOV63 and MOV64 to be opened.

In order to isolate MOV76 the RCIC system would have to initially be placed in standby and a drywell entry at reduced reactor power (approximately 14 -

mode 2).

It is expected that three (3) workers would be inside the drywellforapproxfmately30minutesatextreme temperatures (approximately 135 F) and would receive about 0.2 MAN-REM.

The extreme conditions with an estimated 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months unit unavailability or the resulting plant shutdown justify these emergency circumstances.

V.

Schedule for Attaining Compliance As indicated in Item III

above, River Bend Station is currently in compliance with the applicable Technical Specifications.

VI.

Notification of State Personnel A

copy of this amendment application is being provided to the State of Louisiana, Department of Environmental Quality Nuclear Energy Division.

The State has been verbally notified of this request.

VII. Environmental Impact Appraisal In the event that the RCIC outboard isolation valve MOV64 did not close upon its isolation signal there would be a

potential pathway for fission products to exit the primary containment via the RCIC steam supply line.

This pathway would be through the inoperable and open MOV76 isolation valve all the way to the RCIC turbine control valve MOV45.

However, these fission products would still be contained within a closed system.

Furthermore, leakage from the RCIC i

closed system within the Auxiliary Building would be minimal I

(i.e. valve packing) and would be diluted and filtered through the Standby Gas Treatment System.

Therefore, the I

overall change to the total Design Basis Accident fission produce release is insignificant.

The current FSAR thyroid dose for the exclusion area from the Design Basis Accident is 29 Rem (FSAR table 15.6-7) which is below the 30 Rem regulatory allowable dose.

l 1

l

)

ENCLOSURE 2 PROPOSED REVISED RIVER BEND TECHNICAL SPECIFICATIONS Nmg

~

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRWELL ISOLATION VALVES 5

MAXIMUM SECONDARY PENETRATION VALVE ())

ISOLATION TIME CONTAINMENT SYSTEM VALVE NUMBER NUPSER GROUP (Seconds)

BYPASS PATH c5 (Yes/No)

-4 H

a.

Automatic Isolation Valves 1.

Primary Containment ")

(Continued)

I RHR & RCIC Steam Sup.

1E51*MOVF063 IKJB*Z15 2

9.9 No g

l RHR & RCIC Steam Sup.

IE51*MOVF076 IKJB*Z15 2

13.4 No 1

RHR & RCIC Steam Sup.

IE51*MOVF064(3)

IKJB*Z15 2

9.9 No I

RCIC Pump Suc.-Supp. Pool 1E51*MOVF031 1KJB*Z16 2

30.5 No RCIC Turbine Exh.-Supp. Pool IE51*MOVF077 1KJB*Z17 3

14.2 No

.}

RCIC Turbine Exh. Vac. Bkrs.

IE51*MOVF078 1KJB*Z18B,C 3

16.5 No Cont./Drywell. Purge Sup.

1HVR*A0V165 IKJB*Z31 8

3 No j

J, Cont./Drywell Purge Sup.

1HVR*A0V123 IKJB*Z31 8

3 No Cont./Drywell Purge Outlet 1HVR*A0V128 1KJB*Z33 8

3 No u'

Cont./Drywell Purge Outlet 1HVR*A0V166 IKJB*I33 8

3 No Post-Accident Samp. Sup.

ISSR*SOV130 1KJB*Z6018 10 3

No Post-Accident Samp. Sup.

ISSR*SOV131 1KJB*Z601B 10 3

No l

i i

l

1 TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRWELL ISOLATION VALVES NOTES l

[

(a) Subject to a Type C leak rate test at a test pressure of 7.6 psig except as otherwise noted z

q (b) Also isolates the drywell.

(C) Testable check valve.

(d) Isolates on MS-PLCS air line high flow or MS-PLCS air line header to Main Steam Line low differential pressure.

(*) Receives a remote manual isolation signal.

II) This line is sealed by the penetration valve leakage control system (PVLCS).

,g valves sealed by the PVLCS is not included in 0.60 La Type B and C test total.The combined leakage from (9) This valve sealed by the main steam positive leakage control system (MS-PLCS) i 0

Valves sealed by the MS-PLCS are tested in accordance with Surveillance Requirement 4.6.1.3.f to verify that leakage does i

not exceed the limit specified in Specification 3.6.1.3.c.

This leakage is not included in the 0.60 La j

Type 8 and C test total.

(h) Not subject to Type C leakage tests.

Valve (s) will be included in the Type A test.

l (3) Valve is hydrostatically leak tested at a test pressure of 8.36 psig (1.1 Pa)

I The leakage from j

hydrostatically tested valves is not included in the 0.60 La Type B and C test total.

( ) Not subject to a Type A, B, or C leak rate test.

II) Valve groups listed are designated in Table 3.3.2-1.

i (n0 Wlve IE5t* MbVFO76 is not requeied. 4e be OPEEABLE neo h Oc14,e 4,nso.

J l

i

50 M SEP 2 91986 Docket No. 50-458 O!STRIBUTION FDocket< File SStern NRC POR M0'Brien LPOR (ACRS 10)

Mr. William J. Cahill, Jr.

P0l4 Rdg.

Senior Vice President RBernero River Bend Nuclear Group Dewey,0GC Gulf States Utilities Company NThompson Post Office Box 2951 EJordan Beaumont, Texas 77704 BGrimes ATTN: Mr. J. E. Booker JPartlow

Dear Mr,

Cahill:

SUBJECT:

AMEN 0 MENT TO RIVER BEND STATION OPERATING LICENSE, NPF-47 This confirms our telephone authorization given on September 19, 1986, for the change in the Technical Specifications for River Bend Station, Unit 1, as requested in your letters of September 17, 1986 and September 19, 1986.

Facility Operating License NPF-47 is amended on September 19, 1986 by permitting valve IE51*MOVF076 not to be required to be OPERABLE through October 4, 1986, thus not requirtrg valve IE51*MOVF064 to be shut and thereby permitting RCIC to be operable.

The formal license amendment, our completed safety evaluation, and the Federal Register Notice for this change to the Technical Specifications for River Bend Station Unit 1 is being processed and copies of these documents will be sent to you in the near future.

Sincerely, wl-.I W8d bY Robert M. Bernero, Director Division of BWR Licensing cc: See next page PD# /dM P0f 4/0, t DD/DBLh D/ DBL ((

SStern:lb WButler RHouston RBernero

'l/l1/86 (I'/r/86

/ /86 9 /g /86 "i " '^'t" 060919 yDR ADocK 0:00 e

r

/

UNITED STATES O*-

NUCLEAR REGULATORY COMMISSION

{

,j tyA$HING TON. D. C. 20555 u

%,,,,,/

SEP 191986 Docket No. 50-458 Mr. William J. Cahill, Jr.

Senior Vice President River Bend Nuclear Group Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 ATTN: Mr. J. E. Booker

Dear Mr. Cahill:

SUBJECT:

AMENDMENT TO RIVER BEND STATION OPERATING LICENSE, NPF-47 This confirms our telephone authorization given on September 19, 1986, for the change in the Technical Specifications for River Bend Station, Unit 1, as requested in your letters of September 17, 1986 and September 19, 1986.

Facility Operating License NPF-47 is amended on September 19, 1986 by permitting valve 1E51*MOVF076 not to be required to be OPERABLE through October 4,1986, thus not requiring valve 1E51*M0VF064 to be shut and thereby permitting RCIC to be operable.

The formal license amendment, our completed safety evaluation, and the Federal Register Notice for this change to the Technical Specifications for River Bend Station Unit 1 is being processed and copies of these documents will be sent to you in the near future.

Sincerely, f)

-<.-,.,s..,,__

Robert M. Bernero, Director Division of BWR Licensing cc: See next page

r Mr. William J. Cahill, Jr.

Gulf States Utilities Company River Bend Nuclear Plant cc:

Troy B. Conner, Jr., Esq.

Ms. Linda B. Watkins/Mr. Steven Irving Ccnner and Wetterhahn Attorney at Law 1747 Pennsylvania Avenue, NV 355 Napoleon Street Washington, D.C.

20006 Baton Rouge, Louisiana 70802 Mr. Edward Grant Mr. William H. Spell, Administrator Director - Nuclear Licensing Nuclear Energy Division Gulf States Utilities Company Louisiana Department of P. O. Box 2951 Environmental Affairs Beaumont, Texas 77704 P. O. Box 14690 Baton Rouge, Louisiana 70898 Richard M. Troy, Jr., Esq.

Assistant Attorney General in Charge Mr. J. David McNeill, III State of Louisiana Department of Justice William G. Davis, Esq.

234 Loyola Avenue Department of Justice New Orleans, Louisiana 70112 Attorney General's Office 7434 Perkins Road Resident Inspector Baton Rouge, Louisiana 70808 P. O. Box 1051 St. Francisville, Louisiana 70775 H. Anne Plettinger 3456 Villa Rose Drive Gretchen R. Rothschild Baton Rouge, Louisiana 70806 Lcuisiantans for Safe Energy. Inc.

1659 Glenmore Avenue Baton Rouge, Louisiana 70775 President of West Feliciana Police Jury Regional Administrator, Region IV P. O. Box 1921 U.S. Nuclear Regulatory Commission St. Francisville. Louisiana 70775 Office of Executive Ofrector for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. J. E. Booker Manager-Engineering, Nuclear Fuels &

Licensing Gulf States Utilities Company P. O. Box 2951 Beaumont Texas 77704

i TAttE 3.6.4-1 (Continued) k.

Ct3tTADSOff Af8 DRTEll. I5otATIGI VALVES

\\

~

luutIptM SEUNGAitif l SYSTUt PDIETRATICE VALVE ISOLATIGII TIIE CXNITADestT VALVE Insest

_ InseER tsrae03 4

(Seconds)

BYPASS PATM Autamatic Isolation Valves (YesAlo) i a.

L _ Primary Centafament "3 (Cantfankd)

I

' WR & ECIC Stamm S g.

3E51*ICWF963 I i W e & RCIC.Stasm S g.

3E51*pOWF076 g 1KJB*ZIS 2

3.3 se

, RIE & IICIC Steam Sup.

IE51890VFD64 IL:5*I15 2

13.4 se

[

- l

aCIC Pump Sec.-5gy. Peel If51890VF031g)

ILJB*IIS 2

9.9 No i

t RCIC Tastime Exk.-Segy. Pool IE51890VfM7 IKJe*I17 3'

14.2 se 3EE*ZI6 2

30.5 No aCIC Tintime Ezh. Vac. Skrs.

IE51890VFD78 Cant./Drywell. Purge Sup.

2HWR*ADV155 2KJa*Ilfl8,C 3

16.5 to 1

' Cont /Drywell Purge Sg.

IKJa*I31 8

3 se 2HWR*A0V123

] Comt./Drywell Purge Gottet. Cant./Drywell Purge Outlet IHWE*ADV128 2EE*231 8

3-he 3KJB*Z33 8

3 Ito 3HVR*ADW166 I Post-Accident Samp. Sg.

310B*Z33 8

3 No 155t*S0V130 IKJB*I6013 10 3

IIe.

]' Post-Accfssent Samp. Sg.

1555*50V131 i

IJUB"Z6015 10 3

les 1

t l

i I (

l i.

l 1

t i

Y i

t

4 f

JAOLE 3.E.4-1 (Coatinsed)

CDNTA11eEN Age lumfEtt IS0tATION VALVES i

seTES 1

I

$4 ject to a Type C Teak rete test at a test p 03 Also iselstes the drywell.

ressure of 7.5 psig a rept as otherwise'noted k'IC3 Testable check valve.

i Edl Isolates en 96-ft.CS air line high flow or IIS-PLCS i

pressure.

1 teceives a remote manual isolation signalair 1tne header to IIsin Staae Line low differe I IOI e a

) If3 This line is saaled by the penetration valve leak i

valves sealed by the PVLCS is met included in 0 60 Lage cont ml system (PVLCs).

i\\III Its-PLE5 are tested te acc M.c with ServeillaThis valve sealed by the I

ge control system (MS-PLCS).

Type 8 and C test total. mot exceed the limit specified in Specification 3 613nce R i

last seject to' Type C leakrege tests.

H is leakage is not incinded in the 0.60 La

.... c.

'OI e does Valve is hydrostatically leak tested at a test pressValve(s) will be fxtuded in i

M3 i

hydrostatically tested valves is not included in the 0 60 iure e? a.36 psig (1.1 Pa).

k II Not satject to a Type A, 8, or C 1eek rate test n Type 8 and C test total.The leakage frem I} Valve groups Ifsted are designat d i n Table 3.3.2-1.

e g W ),e I E 5 t

M t W F 0 7 b es nel reto. Ped blee OPEABLE bf. h 4N

\\

i i

l I

I

[

l I

l

ML20214P234

Contents

Text

SUBJECT:

Enclosure:

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr. Denton:

1.0 INTRODUCTION

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr,

SUBJECT:

Dear Mr. Cahill:

SUBJECT:

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ML20214P234

Text

SUBJECT:

Enclosure:

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr. Denton:

1.0 INTRODUCTION

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr. Denton:

Dear Mr,

SUBJECT:

Dear Mr. Cahill:

SUBJECT:

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