05000302/LER-1998-001, :on 980104,SSCs Were Not Protected from Dynamic Effects of Loca.Caused by Failure to Fully Satisfy Design Criteria as Described in FSAR for Dynamic Effects of LOCA Inside Containment.Completed Jco for Targets
| ML20199K875 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/03/1998 |
| From: | Peterson P FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20199K815 | List: |
| References | |
| LER-98-001, LER-98-1, NUDOCS 9802090044 | |
| Download: ML20199K875 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) |
| 3021998001R00 - NRC Website | |
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NRC f oRM 3e6 u s. NUCLEAR REGULAToRf cot
- Mission APFnovse SY Oese No. 31664 44 (4 96)
SAPIR86 64/200
, LK:ENSEE EVENT REPORT (LER) gy Q"g*o Efii'" 'lw.TJ?"f**'t"' g (See toverse for required number of hf '
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CRYSTAL RIVER UNIT 3 05000302 1 OF 6 flf L t (4)
Systems, Strudurer Ano Components Were Not Protected From The Dynamic Effeds Of A Loss of Coolant Accident EVENT DATE (6)
LER NUMBER (t)
REPORT DATE (7)
OTHER P ACILITIES INVOLVED (0) sEgu9(N AL MONTH DAY YEAR YEAR y
MONTH DAY YEAR 01 04 98 98 001 00 02 03 98 OPERATING THIS REPORT IS SUSMITTED PURSUANT TO THE REQUIREMENTS OF 16 CPR $: (Check one et more) (11) l MODE p) 5 20 2201(b) 20 2203(eX21v) 60.73(e)(2)(i) so.7)(a)(2)(wi)
POWER 20 2203(a)(1) 20 2203(a)(3)(1)
X L'.73(a)(2)(61) 60.73(a)(2)(x)
J LEVEL (10) 20.2203(a)(2,(1) 20 2203(a)(3)(6t) 60 73(eX2)(ni) 73 71 g
20 2203(eX2)(61) 20 2203(a)(4) 60.73(eX2)(N) oTHER
[,,,,
f' 20 2203(e)(2)(lll) 50 36(c)(1) 50 73(e)(2)(v) speyin warm m 20 2203(a)(2)(lv) 60 36(c)(2) 60 73(a)(2)(vti) tw in NRO Form 306A LICENSEE CONI ACT r0R THIS LER (12)
P4AMt:
TE4.LPHONE NVMisLR (incWoo Area M)
Patrick M. Peterson, Sr Regulatory Specialist (352) 795-6486 CoMPLET E ONE LINE FOR EACH COMPONENT FAILURE DESCRISED IN THIS REPORT (13)
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03USE SYSTEM COMPONENT MANUFACTURER
CAUSE
SYSTEM COMPONENT MANUFACTURER 0
0 DS V
SUPPLEMENT AL REPORT EXPECTED (1el EXPECTED MONTH ~~
CJAY YLAR YEg sUSMISs10N (If yes. complete EXPECTED sVBMISsloN DATE).
X No DATE (16)
ABSTRACT (LM to 1400 speces. I e., approximately 15 singne-spaced typewntten lines) (16)
On January 4,1998, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR 3) was in MODE 5 (COLD SHUTDOWN), FPC performed an evaluation of the impact a high energy line break (HELB) would have on systems, structures and components (SSCs) inside the containment. FPC has identifiec several SSCs that did not meet the original design criteria for protection from the dynamic effects of a Loss of Coolant Accident (LOCA), This condition was determined to be reportable pursuant to 10CFR50.73(a)(2)(ii)(B), as a condition outside of CR.
3's design basis. The evaluation resulted in the relocation of two Emergency Feedwater Initiation and Control (EFIC) instrumentation lines. The apparent cause of this event was a failure to fully satisfy the Design Criteria as described in the Final Safety Analysis Report for the dynamic effects of a LOCA inside containment. FPC reported two LERs regarding the failure to consider the design criteria for protection of equipment from the dynamic effects of a LOCA outside containment.
9002090044 900203
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PDR ADOCK 05000302 8
I NR S ORM M6A u s. NUCLEAR REGULATORY CoMMlSSloN VA5)
LICENSEE EVENT REPORT (LER) l TEXT CONTINUATION j
F AclLITV NAME (1)
DOCKET LER NUMBER [6)
PAGE 0) l sEoVE NTtAL pgyisioy i
YEAR NUMBER NUMBER 2 CF 6 CRYSTAL RIVER UNIT 3 05000302 98 001 -
00 1C1.htf mots space is reqwed use ndhonal evins clNRc f omt 3%A) l1rl QMCRIPTION Ori January 4,1998, Floride Power Corporation's (FPC) Crystal River Unit 3 (CR 3) was in MODE 5 (C0! C $kUTDOWN). FPC performed an evaluation of the dynamic effects of a loss of waiaat wu;Nnt (LOCA) originating from a break in the high energy portions of Reactor Coolant Syf.em (RCS) [A6) piping on systems, struires, and components (SSCs) inside the containment [NH).
H!gh energy piping is defnied as having an operating temperature of 2 200 degrees Fahrenheit and operating pressure of 2 275 pounds per square inch gauge (psig).
FPC identified several SSCs that did not meet the original design criteria for protection from the d7MmlC effGCts of a LOCA. This condition was determined to be reportable pursuant to 10CFR50.73(a)(2)(li)(B), as a condition outside of CR 3's design baw,s. FPC evaluated the High Energy Line Break (HELB) originating from the pressurizer surge line, core flood lines, decay heat drop line, makeup and purification letdown line, and the pressurizer spray and auxiliary spray line. The evaluation scope included those SSCs that could be affected by pipe whip and jet impingement from the lines identified above. Jet impingement was considared to be emanating from a longitudinal or circumferential break and assumed to occur at any location within the piping with three exceptions. Longitudinal ruptures for lines less than four inches in diameter are not assumed in CR 3's licensing basis, the RCS primary loop piping (hot legs and cold legs) was exempted by a license amendment in 1986, p.nd longitudinal ruptures in the pressurizer surge line were eliminated from consideration based on Generic Letter (GL) 8711 methodology.
EVALUATION FPC identified, by field walkdown, SSCs that were potential targets of pipe whip or jet impingement from a high energy break of the lines described above. These represented potential deviations from the original design criteria for protection from the dynamic effects of a LOCA.
A Deficiency Report was developed which included a Safety Assessment /
Unreviewed Safety Question Determination for the affected SSCs.
These SSCs were dispositioned using one or more of the following criteria:
- 1) The target is not required for LOCA mitigation.
A subset of this category includes targets whose failure will result in additional RCS leakage, urgets required for post accident monitoring in accordance with Regulatory Guide 1.97, and targets which are identified in the Improved Technical Specifications.
A Justification for Continued Operation consistent with GL 91 18, Revision 1, has been completed for these targets. These targets will require physical modification or a change to the Licencing basis to achieve full qualification. One of these options will be implemented prior to start up from Refueling Outage 11.
NRC FoHM 366A U.s. NUCLcAR REGULATORY COMMIS5loN
(*/')
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION
- ACILjiY NAME (1)
DOCKET LER NUMBER lt)
PAGE (3) sEoVEN11AL REyisioy WR NUMDER NUMBER 3 OF 6 CRYSTAL RIVER UNIT 3 05000302 001 -
00 98 iens vimore space os regauro. use nosoonet copes orNec f orm assA) tsis
- 2) The target required modification to remove it from the path of the pipe whip and/or the jet impingement envelope.
Two Emergency Feedwater Initiation and Control (EFIC) [JB) instrumentation lines were determined to require modification. The EFIC instrumentation lines could be impacted by pipe whip from a rupture of the pressurizer surge line and become inoperable. The instrumentation lines are associated with the Once Through Steam Generator (OTSG) level transmitters [LT) for EFIC channels B and D.
The level transmitters are used for detection and actuation of the Emergency Feedwater (EFW)
[SJ) system in the event of a OTSG low level condition. A failure of an EFIC instrumentation line would cause the associated level transmitter to fall low. This would result in an automatic actuation of the EFW system for the OTSG.
- 3) FPC applied the guidance in Generic Letter (GL) 9118, Revision 1, " Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," to disposition nonconformances as follows:
a) The break location was eliminated using the methodology of GL 8711
" Relaxation in Arbitrary Intermediate Pipe Rupture Requirement."
b) The target was eliminated by implementing NUREG/CR 2913 " Two Phase Jet Loads," methodology for redefinition of the jet impingement envelope.
As part of items 1 and 3, final corrective actions in accordance with GL 91 18, Revision 1 are required. Specifically for l tem 3, GL 8711 snd NUREG/CR 2913 are not part of the current licensing basis for CR 3. The use of those documents as the basis for accident analysis involves an unreviewed safety question as defined in 10CFR50.59. As part of a final corrective action, FPC will submit a license amendment request providing justification for these conditions and/or propose additional plant modifications that will be completed prior to start up from Refueling Outage 11. This will be provided in a separate submittal by May 29, 1998.
CAUSE
The apparent cause of this event was a failure to fully satisfy the Design Criteria as described in the Final Safety Analysis Report (FSAR) for the dynamic effects of a LOCA inside containment.
l Additionally, the high energy line break licensing basis contains conflicting requirements concerning protection of redundant, safety related equipment against the dynamic effects of a LOCA inside containment. Section 4.2.6.4 of the FSAR implies that the initiating LOCA event can cause consequential failure to safety related equipment as long as the failure is limited to only one train of redundant safety related equipment. This statement conflicts with the Design Criteria as described in FSAR Section 1.4.
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NRC F oRM 306A u s. NUCLEAR REGULATORY CoMMisSloN (4 D6)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION F ACILITY NAME (1)
DOCMET LER NUMaER 14)
PAGEpl l
SEQUENTIAL REVislod YEAR NUMBER NUMBER 4OF6 CRYSTAL RIVER UNIT 3 05000302 98 001 -
00 1sKT (Itmore spote Is requeed. use 000honalcopes of NRc Form 366A) (11)
IMMEDIATE CORRECTIVE ACTIONS
CR 3 was in MODE 5 (COLD SHUTDOWN) with the Reactor Coolant System pressure less than 100 psig, therefore no immediate corrective actions were necessary, f&B,RECTIVE ACTION FPC performed an extent of condition review and identified SSCs that were potential targets of pipe whip or jet impincoment from a high energy line LOCA inside containment. A Deficiency Report was devek. ped which dispositioned the nonconforming SSCs.
A Justification for Continued Operation was completed for targets whose failure will result in additional RCS leakage, targets required for post accident monitoring in accordance with Regulatory Guide 1.97, and targets which are identified in the improved Technical Specifications.
The EFIC instrumentation lines were modified to remove them from the pipe whip envelope.
FPC will submit a license amendment request and/or propose additional plant modifications that will be completed prior to start up from Refueling Outage 11. This will revise the affected FSAR discussions to clarify and/or correct the conflicting licensing basis requirements.
ACTIONS TO PREVENT RECURRENCE Nuclear Engineering Procedure (NEP) NEP 230,
- High Energy Line Break Design Considerations," will be revised to include high energy lines inside the cor.tainment by May 29, 1998. NEP 230 is applicable to design and design related activities performed by FPC Nuclear Engineering.
PREVIOUS SIMILAR EVENTS
FPC reported two Licensee Event Reports (LERs) regarding the failure to consider the design criteria for protection of equipment from the dynamic effects of a LOCA outside containment.
LER 50 302/88-016-00 discussed safety related modifications made to the plant without the effuct of a HELB outside containment being considered in the design. LER 50 301/97-018-00 discussed the potential for losing both control complex chillers following a HELB in the Intermediate Building.
ATTACHMENTS Abbreviations, Definitions, and Acronyms - Commitments me ronu mA ew
NN FORM 306A u s. NUCLEl.REoVLAToRY Commission LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION F ACILITY NAME (1)
DOCKET lsR NUMeER lt)
PAGE (3)
SEQUENTIAL Revision YEAR NUMBER NUMBER 5 OF 6 CRYSTAL RIVER UNIT 3 05000302 98 001 -
00 tsxt ormore space se ronwoo. use oceoons comes atrmo form assA) (11)
ATTACHMENT 1 ABBREVIATIONS, DEFINITIONS, AND ACRONYMS 10CFR Title 10 of the Code of Federal Regulations CR 3 Crystal River Unit 3 FPC Florida Power Corporation FSAR Final Safety Analysis Report LER Licensee Event Report HELB High Energy Line Break LOCA Loss of Coolant Accident EFIC Emergency Feedwater Initiation and Control psig pounds per square inch gauge OTSG Once Through Steam Generator EFW Emergency Feedwater System SSC System, Structure, and Component GL Generic Letter NEP Nuclear Engineering Procedures RCS Reactor Coolant System Note: Improved Technical Specifications terms appear in capitalization in the text of the LER.
Ells Codes appear in square brackets. Defined terms / acronyms / abbreviations appear in parentheses when first used, me romu m e es)
_._m.
NRC f oRM 306A U s. NUCLE.AN AEGULAioRY CoMMis&loN (4 M)
LK:ENSEE EVENT REPORT (LER)
TEXT CONTINUATION F ACILIT Y NAME (1)
DOCMET Ls4 NUMeER 18)
PAGE 01 MQU NTLAL pEVisioN YEAR N
ER NUMBER 6OF6 CRYSTAL RIVER UNIT 3 05000302 98 001 -
00 iExt ormore spue os requna un nosoonst copee or tmc Form assn (11)
ATTACHMENT 2 1
RESPONSE
COMMITMt.NT DUE DATE SECTION Page 3 FPC will submit a liceAse amendment request May 29,1998.
Evaluation providing justificatiors for these conditions end/or propose add tional plant modifications that will be completad prior to start up from Refueling Outage 11.
Page 4 FPC will submit a license amendment request During the current
Corrective Actions
and/or propose additional plant modifications FSAR revision cycle, that will be completed prior to start up from Refueling Outage 11. This will revit a the affected FSAR disciassions to clari' and/or correct the conflicting licensing ba s requirements.
Page 4 Nuclear Engineering Procedure (NEP) NEP.
May 29,1998.
Actions to Prevent 230, *High Energy Line Break Design Recurrence Considerations," will be revised to include high energy lines inside the containment.. _