IR 05000321/1990022

Insp Repts 50-321/90-22 & 50-366/90-22 on 901015-19.No Violations or Deviations Noted.Areas Inspected:Design Mods & Changes & Followup on SSFI Open Items

ML20062G623
Person / Time
Site:	Hatch
Issue date:	11/19/1990
From:	Jape F, Casey Smith, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20062G620	List: IR 05000321/1990022 ML20062G617
References
50-321-90-22, 50-366-90-22, NUDOCS 9011290355
Download: ML20062G623 (12)

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UNITED STATES

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[ ' ",o NUCLEAR REGULATORY COMMISslON s ,[ REGION 11-g- g 1 101 MAnlETT A STRE ET, i

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e ATL ANT A, GEORGI A 30323

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t Report Nos.: 50-321/90-22fand 50-366/90-22 Licensee:' Georgia Power Compan P. O. Bor 1295- ,

zr N' Birmingham, AL 35201

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isockot .Nos . : 50-321 and 50-366 License.Nos.: DPR-57 cnd NPF-5 facility Name: Hatch 1 and 2 Inspection Conducted: - October 15-19, 1990 r Inspectors: rMMmN C. Smith '

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+ 'h H k a 0lw ut M. Thomas 0 IJ~/9- %

Date. Signed L

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Accompanying Personnel: F. Jape on October 17-19, 1990  !

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Approved by:.

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ecO x' A/v7 ev

/ F. Jape, 'Section Chief fate 71gned

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y SUMMARY-i my . Scope:

This routine announced inspection was conducted in the areas of Design Changes -

l and Modifications and followup on- SSFI open-items, i; Results:

In the areas inspected, violations or deviations were not; identifie I d

Review < of : completed and partially completed design change packages revealed-

that plant modifications are made in a controlled andl technically adequate l

, manner. -Licensee management's involvement in. assuring quality is demonstrated >

by the- development .and implementation of an Engineering Quality Improvement - .!

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u Program' intended to improve the quality of engineering deliverables. ~ Review,of

, selected portions of this program. verified that the program has been effective q lin .c meeting specified objectives. .. Additionally, . a Desian Change Reques+ 1 Closecut Program has been effective in reducing the number of plant modiiicatio '

packages which were still.ope !

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REPORT DETAILS Persons Contacted Licensee Employees i

• K. Breitenback, Engineering Supervisor

. _* D. Edge, Manager, Nuclear Security

• D. Davis, Manager, Plant Administration

• P. Fornel, Manager, Maintenance

• 0. Fraser, Supervisor, SAER

• M. Gauge, Manager, Outage and Planning G. Goode, Manager, Engineering Support

• J. Hammonds,-Supervisor, Regulatory Compliance

• W. Kirkley.. Supervisor, HP/ Chemical Engineering

• J. Lewis, Operations Manager

• D. Madison, Engineering Manager, Hatch Cor *T. Moore, Assistant General Manager, Plant Support

• D. Reid, Acting-General Manager

• J. Robertson, Jr., Acting Engineering Support Manager

• S. Tipps, Manager, Nuclear Safety Compliance

~* Zavadoski, Manager, HP/ Chemistry-Other licensee employees contacted - during this -inspection included ,

engineers and administrative personne Other Organizations c

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B. Garner, Manager, Hatch Project, SCS

• G. McGaha, Design Manager Hatch, SCS

• K. Khianey, Project Engineer, Bechtel NRC Resident Inspectors

• R.'Musser, Resident Inspector

• Attended exit-interview Acronyms and initialisms used throughout this report are listed in the last paragrap .2; Design Changes and Modifications (37700) DCR 1H89-192, RCIC Low Speed Bypass Line This DCR was implemented to provide an orificed bypass test line with a motor' operated valve around the steam admission valve in order to reduce the severity of the RCIC start transient, thereby reducing che possibility of ' an overspeed trip. This design change was -divir'ed into two part The first part was completed under DCR 1H89~192 which involved the mechanical portions and the routing of electrical cables. The second part will be completed under DCR IH90-072 which involves completing the electrical logic chanbes and cable terminations to make the system operabl DCR l O-072 had not been completed at the time of this inspectio ..

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The : inspector performed a field walkdown to verify that DCR 1H89-192 ,

was installed in.accordance with applicable design document l

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b.- DCR 1H89-249, Diesel Generator Check Valves c ,

The modification irvolved replacing the energency diesel generator (EDG) starting air compressor discharge check valves and the check u '

valves installed in each supply header to the air receivers. The old check- valves were replaced with ones that are believed to provide greater assurance of being leak tight. This modification -

also involved installing a test connection ,to allow venting of the header during check valve testin The inpsector performed field walkdowns of EDGs 1A, IB, and 10 to i verify' that the. check valves were installed in accordance with l design document This modification is scheduled to be implemented for Unit 2.during the next Unit 2 refueling outag '

While reviewing- DCR 1H89-249, the inspector observed thht the acceptance cr.iteria specified in the post modification test special purpose procedure 34SP-111689-DC-1-0S,- DG Air Start Check iValve Functional Test, was greater than the total four hour leakage

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specified by desian' engineering. Design engineering specified that the maximum leakage for four hours should be less than 75 psi while ,

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the special. purpose procedure acceptance criteria specified 95 ps l The inspector discussed this issu

with licensee personnel who stated

that the 95 ~ psi 4value stated'in -the special purpose procedure was used in error. The licensee initiated procedure revisions to. correct the special purpose procedure. prior xto-implementation for Unit 2, and'

to correct surveillance procedure 34SV-R43-014-0S, Diesel Generator

' Air Start Check Valve Functional Tes ,

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The inspector reviewed the completed test results for EDGs IA,1B,.

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and 1C and found' that the . maximum. four hour leakage for any of the:

T_, check valves tested was 16 psi, which was well below the maximum leakagt specified by either design engineering or the value specified

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Lin the special purpose procedure. This demonstrated that the check valves-were operating properl !

DCR 1H90-079, .LPCI . Injection Valves-

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' This modification . involved replacing the valve stem and valve disc

'for RHR LPCI injection valves IC11-F017A and 1E11-F017B due to damage to ~ the -valve stem and valve disc of both valves. The exact replacement materials of the' original . stem and disc were unavailable,

.so this change was processed by design engineering who evaluated the adequacy and acceptability of the replacement materia .g

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4 DCR 1H90-153, Diesel Generator Undervoltage Relay Time Dial Setting-t This = modification involved changing the time dial setting for - _

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undervoltage relays CV-7 for each Unit 1 EDG in order for the EDG to it

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be able to energize their emergency busses in less than or equal to ,

c 12 seconds on an automatic start signal. This issue was discovered by site emergency personnel during efforts to develop the surveillance test procedure to meet this new proposed TS requiremen [

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The inspector discussed this item with licensee personnel w 4 stated

'that this issue has been discussed in detail with NRC Regional and Headquarters management. The actions taken to resolve this' issue are l y  : discussed in greater detail in Licensee Event Report (LER)

50-S?1/1990-017 and NRC Inspection Report 50-321, 366/90-2 e.; 'DCR 2H W-154, Diesel Generator Undervoltage Relay Time Dial. Setting y This modificu+1on'~ involved changing the time dial setting for the undervoltage re'ays for the Unit 2 EDGs as discussed above under DCR

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2H90-15 DCR No.88-294, 600 Volt Switchgear RER' No._ 87-514. was ' written on September 25, 1987, to request evaluating the installation of micro-verse. trip elements to replace .

,; existing.EC type trip elements.-.An attachment to the-RER documented

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severe problems with AK_ type breakers equipped with EC trip devices-and: provided a cost-benefit analysis which favored %se of the -

micro-verse' trip devices. The final disposition of GER resulted in

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.the' preparation of DCR No.88-294 with stated design objectives 'of . ,

(1) improving coordination between circuit:breake;s installed in_GE-600V switchgears- 1R23S0034and '1R23S004 and downstream loads, and-(2) reducing maintenance cost associated with,the present: trip device.1 g The' inspector reviewed the; DCP .and verified that a 10 CFR 50'.59

? safety ' evaluation . screening. had been: performed with . no USQ ' , t

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identified.- Design! basis calculations numbers SNC-85-098, Seismic _

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Evaluation' of: Micro Versa Trip, Revision 3, and SEN 89-010, Micro -

Versa Trip Coordination Study, - Revision' 2, was verified as ; having been completed.' Selected portions of calculation, SENt 89 010 were reviewed and verified to be technically adequate. . Additionally, d

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post-modification , test requirements -specifled in surveillance

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, procedure 52SP-111688-RW-1-05 were: verified as having been

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' complete Refurbishment of the - circuit breakers and installation of the micro-versa trip units was performed off-site by the OE The work was accomplished 'under PO No. P-02757 which was reviewed by' the inspector to verify that a applicable technical and a quality '

. requirements had been imposed on the vendor. A field walkdown of the

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installed circuit breakers was performed by the inspector to verify agreement between plant switchgear lineup and design drawings, and calculated trip setpoints versus actual setpoints of selected circuit breaker ,

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" g .L :DCR'No 88-349T, 100 Percent Stator Ground Fault Protection Relay DCR'No. 88-349T was written to remove the 100 percent stator ground l fault relay, 64 GSX, from service to facilitate troubleshooting the D relay. This temporary plant mcdification- is non-nuclear safety l related . and was administered under the licensee's temporary modification progra .

The inspector verified by review of objective evidence that (1) disabling of the relay and removal of the NW-511-2 module was performed in a controlled manner via MWO No. 1-88-7849, (2) the TM while still installed was periodically

reviewed' in accordance with the TM program controls, and 1 '

(3)' reinstallation of the NW-511-2 module and enabling of the relay x af ter ' repairs by the vendor was adequately controlled via MWO 1-89-4815. . Successful completion of post installation test was verified .by review of: document number 571T-S32-002-1N, Brown Bover Type GIX 103. Relay '100 Percent Stator Ground: Fault Protection,

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Revision s DCR -No.82-173, RHR Time Delay Relays SILE No. 230, Revision 2 dated July 1981, provided information C concerning the: time delay of GE CR 2820 time delay relays used in-

the core' spray,;RHR, and automatic depressurization system _ The SIL ,

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stated that the time . delay of the CR 2820 relay tends to increase af ter .long periods > in the de-energized state and recommended .their !

replacement. with Agastat time- delay, relays. The 'above DCP was ,!

prepared in response to DCR .82-173 dated May 25, 1990, and had a-stated design objective of providing the RHR system logic with time '

delay { relays that will maintain their catalog &ct'ied accurac ,

- The scopetof the hardware changes involved the changeout of 11 time delay relay .The inspector revewed the DCP n and determined that part;ai f k implementation of de plant modification ohad been completed with o:

changeout of. relv LIEs 11-K93A and The replacement relays were i'

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Struthers , Dunn,- model- 236ABX A : documented basis for use

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of. the' Struthers cDunn relays was notL part of .the ,DCP. Licensee

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management stated that, ~ based on the poor experience with the

^d Agastat. time ' delay relays, a decision -was made to use Struthers J Dunn relays. The replacement relays are used in the logic for the V RHR' Heat Exchanger -Bypass Valves E11-F048A and B. The inspector verified that'-the10 CFR 50.59 safety evaluation bounded the scope of the -designiactivities,' that the drawings accurately reflected the

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hardware' changes,= and the elementary diagrams correctly implemented-the required logic. with specified time delay. Post modification ,

testing ~ was verified to have been successfully completed in accordance with 'special purpose procedure 42SP-052990-PW-1-1S, -

1-K93A yand 8, Functional Test for DCR 82-173. One minor weakness .

related to the ' absence of installation instructions for the ;

replacement-relays was identifie l

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.- This information was added by FCR 82-173- A walkdown. or the

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installed relays verified that installation was - completed in accordance with design document <

.No violations or deviations were' identified in the areas inspecte ,

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< Engineering-Support Activities Engineering Quality Improvement Program p Licensee management . is presently implementing an Engineering Quality Improvement Program intended to improve the quality of engineering ,

. deliverables. The following selected areas were reviewed to ascertain

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the program effectivenes '

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(1) FCR Tracking FCRs are the administrative process used for making onsite changes to DCPs prepared by offsite engineering organizations. The licensee has

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developed and implemented 'a program to evaluate and trend FCRs

ganerated (1) during DCP installation and (2) identified during plant / equipment walkdow A uniform list of reason codes has been prepared;for use by SCS and Bechtel pursuant to the analysis of field change; data 4transnitted to the engineering organizations. The w inspectorEdetermint d that existing work practices related to FCR j tracking needs has lot been incorporated in SCS procedure .

V  : Licensee management .has established a goal of not more than two FCRs

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per DCP with. assignable cause to design-engineering activities.~.

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Quarterly reports with data presented in bar chart form is prepared-A along with FCR tracking forms delineating corrective actions for each i

^ deficiency. At' the time of the' inspection, a summary report documenting. the analysis 'of data gathered up to September 1,1990, j g

had been; prepared for management revie '

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? .An-evaluation of the'effectiveress of the FCR tracking program could not be made given the relatively short. time it has been in effec Baseline ' performance data is presently being gathered and analyse The effectiveness of the program will be demonstrated- after it has-been'in effect for some reasonable tim ' ' l

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,(2) ABN Program no f '

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- ABNs are the administrative process used to document -discrepancie *

identified .between ; approved design drawings / documents and as-found plant installation. If required by plant administrative procedures, a deficiency card is prepared to document the discrepanc Alternatively, an explanation will be provided on the ABN form as to why a deficiency car 3 is not require The inspector determined that Hatch design configuration project had

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a dedicated staffing leve' of 61 equivalent people during 1989 and 45 equivalent people for 1990. The processing of ABNs were performed in

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m accordance with approved procedures and the results of the drawing

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update activities are as follows:

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End of 1988 I, 82' percent of total drawings had no outstanding ABNs i '

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94 percent of total drawings had no outstanding ABNs

Current 1990 99 percent of total drawings had no outstanding ABN ,

. Additional statistics presented by the.-licensee demonstrated that the-

, program has been extremely effective.in ensuring plant configuratio All category 1. drawings, which require 30 days to be updated, are current. .There hasE been a significant reduction of Category 3-drawings. . These drawings are updated 90 days after. receipt- of the -

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third ABN or upon request, and was the largest population that contributed.to lack of configuration contro The results of the drawing; update are as follows:

1989 Category 3: 79,276 1990 Category 3: 457 Updated Category 3 drawings are now controlled as Category 2 drawings

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which ensure a more- timely revision than.that for Category 3. The inspector- concluded that the ABN-~ program has been effective in maintaining consistency between as design drawings;and actual plant i -installatio '

b.- DCR Closeout'. Program A formalized DCR closeout prog'am was initiated in January 1990, to reduce a backlog of 186 open. DCRs. Phase:1 of the program consisted of (1.'

L defining the DCR closcout-program scope, (2) review cf imp'erisuted het not closed out DCRs for completeness,-(3) verifying accuracy of impimentation ;

by .walkdown, (4) taking corrective acticas as ducrepancies are

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identified -(5) updating plant ' documents, and (6) closee:t of the DCR ,

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The program was implemented in accordance with adminis'.rative controls

. delineated.in the fell:.;ir.g procedures / department in=tructions:

Special Purpose Procedure N SP-112989-0B-1-0S, DCR Processing / Voiding Det6iled Instructions, Revision *

C partmental Instruction DI-ENG 42-1189N, DCR Closcout/ Voiding Guidelines, Revision Proced ne 42EN-dG-001-05, DCR Processing, Revision .

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The inspector verified by review of objective' evidence that, at the time of the inspection, the DCR closeout program had accomplished the following:

122 DCRs had been closed ou DCRs had been voided in accordance with written criteri five %Rs had.been transferred out of the scope of the closecut '

program for.various reason DCRs were still' outstanding awaiting closure ,

Substantial resources were expended by licensee management to accomplish the above tasks via a dedicated group assigned to the DCR Closecut Group. Phase 2 of- tra DCR Closecut Program is presently in the planning stage and is being dev0 loped to. process and close out partially

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implemented DCRs. Phase 2 will miclude DCRs that are outside the scope of

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Phase 1, and is intended to assest and complete all required actions for 4 implementing and closeout of approximately 50 partially implemented DCRs having various classification ,

T'he inspectors concluded that-the DCR Closeout Program has been effective sin reducing the number of.open DCRs identified in Phase 1.

L Deficiency Cards and Significant Occurrence Reports

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Deficiency cards' are used to identify and document problems found in

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the plant.. The DCs.are initially reviewed for significance and potential impact - on plant. operations' by the STA and applicable unit shift i

. supervisor. ~1tems which.are determined to be significant are given a L significant ~ occurrence. report number by the Nuclear Safety and Compliance Department and are assigned: to various plant departments' for ,

followup and resolution. The percentage of DCs written thus far in 1990

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< which have resulted in SORS'is approximately three percent. The licensee trends SORS and DCs on a' quarterly basis .

- The inspector reviewed selected SORS which have been assigned to the site Engineering Support Department.- Nearly 40-percent of the 1990 S0Rs were .

assigned to Engineering Support for actio Enoineering Support '

nnagement stated that meeting the SOR response due dates is continually emphasized to engineering personne The inspector. noted that the response dates for SORS assigned .to engineering were met in nearly all

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instances. The inspector noted a few instances where the SOR responses prepared by engineering were rejected by the PRB and sent back to

' engineering -for a variety of reasons. Subsequent responses for the applicable SORS were resubmitted and accepted by the PR The inspector

.noted that, while a .Sw of the engineering responses to SORS were considered inadequate by J.a PRB, the subsequent review and approval demonstrated that the licensef s overall review process for SORS (including the PRB) was effectivt. in ensuring that adequate resolutions for SORS were being provided by the various department ; 1 r, .-.

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. Temporary Modifications

The inspector reviewed the licensee's temporary modification program and the- efforts to reduce the TM backlog. A team consisting of

' engineering, maintenance, and operations personnel was formed in mid-1989 to evaluate TMs that were greater than 90 days old. The team recommended removal of those TMs- that could be removed, implementation of DCRs for

.some, and incorporation of repetitive TMs into procedures. This process resulted in the number of TMs greater than 90 days old being reduced from approximately 55 in June 1989 to 18 in September 1989. At the beginning of 1990,' ten TMs greater than 90 days old were open, nine of which were related to the Unit 1 outage. As of October 1990, 15 of the 33 open.TMs were greater than 90: days old. Licensee personnel stated that the main p' . reason for reducing the number of TMs greater than 90 days is to reduce the paperwork load on the unit shift supervisor The goal is to reduce the number of. non-outage TMs to zero. The status of TMs greater than 90 days old are discussed monthly with- the Plant General Manage :The. inspector reviewed selected open TMs and the licensee's administrative controls covering'TMs. .All safety-related TMs are required to be reviewed

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, year. All TMs 'that remain installed for greater than 90 days must be reviewed by Engineering' Support. The unit shift supervisors review the TM logs monthly and identify to Engineering Support those TMs which require review because the TMs are greater than 90 days old. The inspector

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review ':all the open safety-related TMs that were greater than 90 days e old a verified that both the ~ unit :;hift supervisors and Engineering

, Support reviews > were performed -in accordance with the applicable administrative controls. The following safety.-related TMs were reviewed:-

1-89-0751 Temporary Clamp -to Block OffLAir Leak on Transfer Canal'

i Seal-' Assembly, K '2-90-005 Refueling Bellows Leak Detection Alarm Remove .006 Annunciator for 2E41-F006 0verload Remove ,

2-90-015 Points 14, 16, and 20 : Causing. Ar.nunciation for Recirculation Temperature. Recorde Clamp Placed on Three . Pin Hold. Leakss of PSW Discharge ~

Header Minimum F1ow.

I! The inspector determined from reviewing the' licensee's TM program that the licensee's efforts to control TMs have been effective in reducing the

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number of outstanding TM I No' violations or deviations were identified in the areas inspecte I  !

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7. . ActiononPreviousInspectionFindings(92701)

c a, .(0 pen)IFI 50-321,366/89-08-03, PSW System Design Pressure, j

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h ' A' question was raised during the SSFI with respect to pressure experienced 'by various pieces of equipment in the Reactor Building which are serviced by Plant Service Water System. Specifically, ;

during an accident wherein the Turbine Building plant service water t

flow isolates, the PSW pump backs up its performance curve, resulting .

y": in system pressure slightly higher than normal, The piping i

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specifications. cover the higher pressure by specifying design and maximum pressure at 180 and 190 psig, respectively. However, several I components have been identified as hning a design pressure lower {

6 than that expected during an acciden ;

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.The licensee's action, upon discovery of this item, included'a i justification for continued operations and a' commitment to resolve

- this' issue in the long term. The JC0 was reviewed and accepted by the NRC and work has progressed on the long-term'fix. 'The long-term fix includes several modifications. Design change packages have been

, prepared and approved, but the installation has not been-don Final ,

resolution of this issue will be completed when all hardware fixes l

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are installed. These are scheduled for future planned outage .(Closed)IFI 50-321,366/89-08-04, Common 10-Inch PSW Discharge Lin This; issue: relates- to a safety concern- raised during the SSFI l

• , 'regarding the common cooling water discharge line for the EDG j

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.o 'and the ' possibility. of losing the cooling water for the'EDG- due j

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to line blockag . . a

' aThe licensee responded to this question by reviewing the drawings of j

- the piping . system (following Lthe inspection). The-design drawings, H-11600 and H-11146, show a rupture disc, installed downstream of all branch lines connecting to the common discharge' header. The rupture disc _ is designed to fai_1 at 95 psig,'which is a realistic pressure ,

that would be experienced if such'a catastrophic failure occurred, j 1=

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In addition, the licensee performed a probabilistic evaluation to: I

determine the risk significance of this: failure. The evaluation :

concluded that pipe collapse contributes less than one ' percent' to l1 the frequency of station blackout occurring at Hatch due to al ~

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Therefore, considering the above two actions, this issue is closed.- (Closed) IFI 50-321,366/89-08-07, Emergency Diesel Generator ,

CARD 0X Syste !

During the SSFI, the design basis of the EDG CARD 0X system was J questioned because the basis did not consider the ability of the fire detector switches to withstand a seismic event. In response to this issue, the licensee had the components in question tested at the Wyle Laboratorie All components tested met the established criteria for Plant Hatc " ,

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Wyle submitted a report, Reference No. 410338-002, to Southern Company )

Services, dated April 17, 1990. The test report described the test i requirements, procedures, and results. The test was completed with no abnormalities and concludes that inadvertent activation of the CARD 0X system would not occur during a seismic event. This closes all action on l

.this issu j (Closed)'IFI 50-321,366/89-08-08, Seismic Qualification of EDG Low ;

Lubricating Oil 3 witche The status- of this. item was reported in NRC Inspection Report No.: 89-30 and -was kept open pending receipt of the seismic qualification documentation for the Allen Bradley pressure switche These switches are: Allen Bradley type 836-C2, MPL No R43-N757A and *

R45-N:757C. To seismically qualify the switches, the licensee sent samples of- ANC0 Engineering. Inc., for testing. The test report indicated the test conditions and concluded that no chatter or bounce occurred upon-change of switch state during the seismic tes Based on these'results, this issue is close J (Closed)IFI 50-321,366/89-08-10, Electro-thermo Links on Diesel Room H Roll-up Doors,and-Fire-Damper The status of this item was reported in NR'C Inspection Report N . The item was kept open because no qualification documentation

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' was available for the fire damper electrothermal fusible links. The roll-up doors .and . fire dampers for? the EDG building are not ;

safety-related and were purchased without seismic qualificatio The electrothermal ' links are a component of the roll-up doors and j fire damper :l

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Seismic calculations Lfor the fusible links have been completed by l Southern Company Services. . It was' determined that the fusible links j are seismically adequateifor' the installationLat Hatch. Therefore,

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the fusible link should not fail'.during a seismic event, and the doors !

should ' remain. open to perform their intended task. The calculations

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. reviewed were SCN 89-029, Revision- 0, and SCN 89-030, Revision ~

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This-. completes'all action required for this ite .l ExitInterview'(30703)

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The_ inspection scope and results were summarized on.0_ctober 19, 1990, with _,

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those. persons' indicated in- paragraph 1. The inspectors described the m

areas inspected and discussed in detail the inspection results listed ;

belo Proprietary. ' information is not- contained- in this repor '

Distenting comments ~were not . received from the license a

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.The licensee was informed that the following items were closed..

I l IFI 50-321,366/89-08-04, Comon Ten Inches PSW Discharge Line L IFI 50-321,366/69-08-07, Emergency Diesel Generator CARD 0X System

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M IFI 50-321,366/89-08-08, Seismic Qualification of EDG Low Lubricating

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Oil Switches IFI 50-321,366/89-08-10, Electro-thermo Links on Diesel Room Roll-up Doors and Fire Dampers-

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.The licensee was informed that the following item remains ope IFI 50-321,366/89-08-03, PSW System Design Pressure

, Acronyms and Initialisms 1 .AB As Built Notice-DC Deficiency Card

DCP Design Change Package p iDCR Design Change Request EDG Emergency Diesel Generator FCR Field Change Request G General Electric

+' , IFI Inspector Followup Item l JC0 Justification for Continued Operation Licensee Event Report

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LER LPCI Low Pressure Coolant Injection, -

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MWO' Maintenance Work Order

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.NR Nuclear Regulatory Commission H, ,OEM Original Equipment-Manufacturer

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-P0 . " Purchase Order

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PRB' Plant Review Board L -r .

PSI Pounds Per Square Inch Gauge W , PSW' Plant. Service Water

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> :RCIC -Reactor Core Isolation Cooling i

- .RHR Residual Heat Removal ,

RER- Request for Engineering. Review <

SC Southern Company Services L

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'SIL Service Instruction Letter

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-SOR Significant Occurrence Report-l SSF Safety System Functional Inspection m STA Shift Technical Advisor TM: Temporary Modification-

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'TS Technical Specification USQ Unreviewed Safety Question

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