ML20044A279

Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Table Indicating Status of Each Generic Safety Issue Encl

ML20044A279
Person / Time
Site:	Vogtle
Issue date:	06/25/1990
From:	Hairston W GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TASK-***, TASK-OR ELV-01726, ELV-1726, GL-88-14, GL-90-03, GL-90-04, GL-90-3, GL-90-4, NUDOCS 9006280290
Download: ML20044A279 (7)

Text

' Georgia Power Conuiany .

aw *.

, 333 Piedmont Avenue

(- AtttAta, Gwg-a 30308 -

Teleptone 404 526 314s Madang Address o<,t .c ox 1295 Birmingham, Alabama 35201 Telephone 205 868 5581 June 25, 1990 " '"* "' * ""

W G. Hairston,lil sena vce r,esident ELV-01726 Nuclear Operanons 0411 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

V0GTLE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 90-04 NRC Generic letter 90-04, " Request for Information on the Status of Licensee Implementation of Generic Safety Issues Resolved with Imposition of Requirrinents-

-or Corrective Actions", was issued on April 25, 1990, to all holders of operating licenses and construction permits for nuclear power reactors. The generic letter requested-licensees to provide information concerning the implementation status of GSI requirements for which a final technical resvlution has been achieved. The information is requested to assist the NRC staff in the review of GSI implementation status.

As requested, the completed table indicating the status of each GSI,-as referenced in' Generic Letter 90-04, is provided as. an enclosure. Supporting documentation is available for review upon request. The status indicated is consistent with the instructions provided in Generic Letter 90-04. The attached notations regarding the status of individual items are necessarily brief and cannot address all of the details of the more complex items. Considering these constraints, the enclosed information is correct to the best of my knowledge.

If you have any questions in this regard, please contact this office.

Sincerely, C

h.h . hn W. G. Hairston, III WGH,III/PAH/gm Enclosure xc (see next page)

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-0 , . - 1 Georga PoEer d i

U. S. Nuclear Regulatcry Comission ELV-01726 Paae Two xc: Georoia Power Company ,

Mr. C. K. McCoy Mr. G. Bockhold, Jr.

Mr.-R. M. Odom Mr. P. D. Rushton  !

NORMS U. S. Nuclear Reaulatory Comission Mr. S._D. Ebneter, Regional Administrator Mr. T. A. Reed, Licensing Project Manager, NRR Mr; B. R. Bonser, Senior Resident inspector, Vogtle I

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, - c, ENCLOSURE 4

STATUS OF LICENSEE IMPLEMENTATION OF GENERIC SAFETY ISSUES RESOLVED WITH IMPOSITION OF REQUIREMENTS OR CORRECTIVE ACTIONS Docket llam Numbers Status Comments l l Safety Concerns Associated 50-424 NA l_

with Pipe Breaks in 50-425 NA i

the BWR Scram System BWR Scram Discharge 50-424 NA I Volume System 50-425 NA Reliability of Air Systems 50-424 C Licensee's response to i;

50-425 C Generic Letter 88-14 l

-submitted in letter l ELV-00197 dated 2/17/89.

Resolved per NRC letter dated 3/5/90.

Improving the Reliability 50-424 I- Resolved per NRC 1

of Open-Cycle Service 50-425 I letter dated 5/16/90. J Water Systems ]

.l Improved Accident 50-424 C Resolved. See Appendix l

! Monitoring 50-425 C M, NUREG-1137 Supplement  !

L No. 2. l 1

l 1

li Item 1.1< Post-Trip 50-424 C Resolved. See NUREG-ll37 Review (Program Description 50-425 C Supplement No. 2.

! and Procedure) 1 Item'l.2-Post-Trip 50-424 C Resolved. See NUREG-ll37 l Review-Data and 50-425 C Supplement No. 3.

Information Capability Item 2.1-Equipment 50-424 C Resolved per NRC letter

[ Classification and 50-425 C dated 7/29/87 and L Vendor Interface (Reactor NUREG-1137 Trip System Components) Supplement No. 7.

Item 2.2.1-Equipment 50-424 C Resolved per NRC letter Classification for 50-425 C dated 11/3/89. i Safety-Related Components l 1

Docket 1t1m Numbers Status Comments Item 2.2.2-Vendor 50-424 E To be resolved by Interface for 50-425 E submittal addressing Safety-Related Components Generic Letter 90-03; due 9/90.

Items 3.1.1-Post 50-424 C Resolved per NRC letter Maintenance Testing 50-425 C dated 7/7/87 and (Reactor Trip System NUREG-ll37, Supplement 1 Components) No. 7.

' Items 3.1.2-Post 50-424 C Resolved per NRC Maintenance Testing 50-425 C letter dated 7/7/87 and (Reactor Trip System NUREG-ll37, Supplement Components) No. 7.

Item 3.1.3-Post- 50-424 C Resolved. See l Maintenance Testing - 50-425 C NUREG-1137, Supplement- l Changes to Test Requirements No. 4.

(Reactor Trip System Components)

Item 3.2.1-Post- 50-424 C Resolved per NRC letter Maintenance Testing 50-425 C dated 7/7/87 and (All Other Safety-Related NUREG-1137, Components) Supplement No. 7.

Item 3.2.2-Post- 50-424 C Resolved per NRC Maintenance Testing 50-425 C letter dated 7/7/87 i (All Other Safety-Related and NUREG-ll37,

, Components) Supplement No. 7.

Item 3.2.3-Post- 50-424 C Resolved. See Maintenance Testing- 50-425 C NUREG-ll37, Supplement Changes to Test Requirements No. 4.

.(All Other Safety-Related Components)

Item 4.1-Reactor 50-424 C Resolved. See Trip System 50-425 C NUREG-1137, Supplement Reliability (Vendor-Related No. 2.

Modifications)

Items.4.2.1-Reactor 50-424 C Resolved. See Reactor Trip System 50-425 C NUREG-ll37, Supplement Reliability-Maintenance No. 2.

and Testing (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers) 2

4 Docket ligm Numbers Status Comments  :

~

Items 4.2.2-Reactor 50-424 C Resolved. See Trip System 50-425 C NUREG-ll37, Supplement Reliability-Maintenance No. 2.  !

and Testing (Preventive #

Maintenance and Surveillance Program for Reactor Trip Breakers)

Item 4.3 - Reactor Trip 50-424 C Resolved. See System Reliability - Design 50-425 C NUREG-1137, Supplement Modifications (Automatic No. 3. '

Actuation of Shunt Tri)

Attachment for Westing 1ouse and B&W Plants)

Item 4.3-Reactor 50-424 C Resolved. See Trip System 50-425 C NUREG-1137, Supplement Reliability-Tech No. 3.

Spec Changes (Automatic Act'Jation of Shunt Trip Attachment for Westinghouse and B&W Plants)

Item 4.4-Reactor 50-424 NA Trip: System Reliability 50-425 NA (Improvements in Maintenance and Test' Procedures for B&W Plant)

Item 4.5.1-Reactor 50-424 C Resolved per NRC

. Trip System Reliability - 50-425 C letter dated 7/7/87 and l

Diverse Trip Features NUREG-ll37, Supplement l (System Functional Testing) No. 7.

Item 4.5.2-Reactor Trip 50-424 C Resolved. See l System Reliability - Test 50-425 C NUREG-ll37, Supplement Alternatives and Intervals No. 4.

L (System Functional Testing)

Item 4.5.3-Reactor Trip 50-424 C Resolved per

l. System Reliability - Test 50-425 C NRC letter dated 7/3/89.

l Alternatives.and Intervals l (System Functional Testing)

Long Range Plan for 50-424 NA Dealing With Stress 50-425 NA Corrosion Cracking in )

BWR Piping ,

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Docket

-lits N.umbiti Status Comments Steam Binding of Auxiliary 50-424 C. Resolved per NRC Feedwater Pumps 50-425 C letter dated 5/13/88'

. Confirmation letter '

(GN-1519) sent to NRC for Unit 2 on 12/12/88. ,

RCS/RHR Suction Line Valve 50-424 C -See letters ELV-01092- 3 Interlock on PWR's 50-425 _1 dated 11/30/89, ELV-00186 dated  ;

02/02/89, and ELV-0109 dated 12/29/88. -

Auxiliary'Feedwater 50-424 NA System Reliability 50-425 NA Snubber Operability 50-424 C Resolved.

Assurance-Hydraulic 50-425 C See Technical Snubbers- Specification 3/4.7.8.

Snubber Operability 50-424 C Resolved.

Assurance-Mechanical 50-425 C See Technical Snubbers Specification 3/4.7.8.

Steam Effects on 50-424 NA BWR Core Spray Distribution 50-425 NA ,

. Adequacy of Offsite Power 50-424 C Resolved. See Systems 50-425 C NUREG-ll37, Section 8.4.1 and Supplements 5, 6, and 8.

i Behavior of BWR 50-424 NA Mark III Containments 50-425 NA Develop Design, Testing 50-424 C Resolved. See and Maintenance Criteria 50-425 C -NUREG-ll37, Sections for Atmosp_here Cleanup 6.5.1, 9.4.1, 9.4.2, System Air Filtration and 9.4.3 and 9.4.5 Adsorption Units for and Supplements 4, 7, Engineered Safety Features and 8.

and for Normal Ventilation L Systems.

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Docket lifs Numbers Status- Comments-Isolation of Low Pressure- 50-424 C Resolved. See Technical Systems Connected to the 50-425 C Specification Table Reactor Coolant 3.4-1 System Pressure Boundary _

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1 Status

, NA Not applicable- ,

NC- Applicable, no changes necessary to implement resolution '

Applicable,. submittals made or changes complete-C I Applicable, changes are necessary, such changes not fully implemented E Implemented guidance for GSI recently-issued, under evaluation i 5 L

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