ML20044C989

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LER 93-003-00:on 930407,scram Time Testing on All Control Rods Not Completed Prior to Exceeding 40% Rated Thermal Power.Caused by Personnel Error.Personnel Counseled & Procedures Will Be revised.W/930503 Ltr
ML20044C989
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 05/03/1993
From: Beckham J, Tipps S
GEORGIA POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-3269, LER-93-003-01, LER-93-3-1, NUDOCS 9305140271
Download: ML20044C989 (6)
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     .       George Power Com;cny
  'Y
  • Co invernets Cemer Parctway Post CMice Sex 129s BirrrungNim. Nabama 3s201 Telephone 205 B77-7279 L

J. T. Beckham, Jr. Georgia Power 

             %ce President Nadear Hatch Profect                                                      If+ 50uf'*"' C E!'C V uf'"

May 3, 1993 Docket No. 50-366 HL-3269 005299 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant - Unit 2 Licensee Event Report Personnel Error Results in a Condition Prohibited by the Technical Specifications Gentlemen: In accordance with the provisions of 10 CFR 50.73(a)(2)(1), Georgia Power Company is submitting the enclosed Licensee Event Report (LER) concerning a personnel error which resulted in a condition prohibited by the Technical Specifications. This event occurred at Plant Hatch - Unit 2. Sincerely, G wdA [ J. T. Beckham, Jr [ v JKB/cr

Enclosure:

LER 50-366/1993-003 cc: Georoia Power Company Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS U.S. Nuclear Reculatory Commission. Washinoton. D.C. Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Regulatory Commission. Reaion 11 Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch sp? 9305140271 936503 1 PDR S ADOCK 05000366 I PDR k) 

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LICENSEE EVENT REPORT (LER) 

    +Acain hAMt (i)                                                                                                    vacu i huxetk v1                +w          m      l PIRU E. I. HATCH, INIT 2                                                           05000366                      1 l0d5               I IITLE (4)                                                                                                                                                             l PERSOMEL EPROR RESULTS IN A CONDITION PROH1BITED BY THE TECHNICAL SPECIFICATIONS                                                                                    $

I EVEhl DATE (5) L E fa hbMbER (6) EEP0kT Call (7) CINER F ALILIIIE5 Ihv0LbED (6) M0h1N LAY TEAR TEAR 5EQ huh REW M0hIn DAY YEAR FACIllIY hAME5 DOLLE1 huMBER(5) i l 05000  ; t i i 1 04 07 93 93 003 00 05 03 93 05000 . OPERAT1hG S MI M 10 M @mM M M UR (11)  ! N00E (9) 1 20.402(b) _ 20.405(c) 50.73(a)(2)(iv) 73.71(b) POWER - 20.405(a)(1)(1) ~ 50.36(c)(1) ~ 50.73(a)(2)(v) 73.71(c) i LEVEL 038 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) 

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20.405(a)(1)(iii) T 50.73(a)(2)(1) 50.73(a)(2)(viii)(A) Abstract below) , 20.405(a)(1)(iv) - 50.73(a)(2)(ii) - 50.73(a)(2)(viii)(B) t 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) ilCEh5E L C0hTACT FOR THI5 LER (12) i NAME TELEPncht humbER LREA CODE . STEVEN B. TIPPS. MANAGER NUCLEAR SAFEIY AND COMPLIRJCE, IRTCH 912 367-7851 l LOPPLETE DhE LlhE FOR LACH FAILbkt DESCRIEED lh Th15 kEP0kT (13) CAUS[ SYSTEM COMPONEhT jAhgt-g gFgT g CAUSE SYSTEN COMpDhENT jU ( FAhkUFAC-PORT I i 5bf PLEMEh1 AL REPORI LAFELIED (14) M0hlh DAY TEAR i EXPECTED C SUEMISSION  ; 

    ] YE$(If yes, complete EXPECTED SUEMISSION DATE)                           % NO                                      DATE (15)                                         j AE5TRALT (16)                                                                                                                                                         !

On 04/07/93 at 0103 CDT, Unit 2 was in the Run mode at a power level of 938 CMWT 3 (38.5% rated thermal power). At that time, licensed shift personnel were l performing Control Rod Drive (CRD) system scram time testing in accordance with Unit 2 Technical Specifications section 4.1.3.2.a which requires the testing to  ! be completed prior to exceeding 40% rated thermal power. Per procedure  ! 34GO-OPS-001-2S, "PIANT STARTUP," reactor power was being maintained below 40%  ! in order to comply with the specification. When licensed personnel prepared to j test control rod 26-31, they expected that fully withdrawing the rod might cause j thermal power to increase to a level very close to 40%. Therefore, they slowly > withdrew the rod a few notches at a time while observing the Average Power Range  : Monitors. The monitors, which have strip chart indication, appeared to indicate l approximately 40%. By the time the rod had been fully withdrawn, however, the plant's Process Computer indicated that thermal power had actually reached 40.71%, violating the requirement of the specification. Licensed personnel then , scrammed the rod, and thermal power decreased to approximately 38.3%. j The root cause of this event is personnel error on the part of the licensed j superintendent of shift. The plant startup procedure stated that the testing should normally be performed at 35% rated thermal power. However, the plant was operated at a higher power level (38.5%) with the result that when rod 26-31 was , fully withdrawn, core thermal power increased to 40.71%. t Corrective actions for this event included counseling the licensed superintendent of shift. Also, the plant startup procedure will be revised by  ; 7/16/93 to provide improved administrative control of reactor power.  : 1 

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TEXT CONTINUATION l FACILITY %AME (1) DDCKET NUMEER (2)- LER NUMBER (5) PAGE (3) f VEAR SEQ hum REV i PIANT E. I. HATCH, UNIT 2 05000366 93 003 00 2 0F 5 l ILAI  ? PIANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor [ Energy Industry Identification System codes are indicated in the text as (EIIS  ; Code XX). ' l I DESCRIPTION OF EVENT , i On 04/07/93 at approximately 0103 CDT, Unit 2 was in the Run mode at a power - level of 938 CMWT (38.5% rated thermal power). At that time, licensed shift 1 personnel were performing Control Rod Drive (CRD, EIIS Code AA) scram time  ! testing per procedures 34GO-OPS-001-2S, " PLANT STARTUP," and 42SV-C11-001-2S, i 

               " CONTROL ROD SCRAM TESTING." These procedures implement the requirement to perform individual control rod scram time testing per Unit 2 Technical Specifications section 4.1.3.2.a. This specification requires that scram time                                  i testing on all control rods be completed prior to exceeding 40% of rated thermal power. Seven control rods had been tested successfully and operations personnel                                ;

were preparing to test control rod 26-31. I i Prior to rod 26-31 being tested, shift personnel recognized the rod as a , high-worth rod. Accordingly, they withdrew the rod in stages (rather than , continuously) while observing the Average Power Range Monitors (APRMs, EIIS Code  ! IG) to ensure that core power did not exceed 40% as indicated by the APRM strip - chart recorder. As the rod was withdrawn, core power increased as expected, but appeared on the APRM strip chart to remain at approximately 40%. By the time the rod was fully withdrawn, however, the plant's Process Computer (EIIS Code j IO) indicated that thermal power had reached 40.71%. This represented a l condition prohibited by the plant's Technical Specifications since core thermal power rose briefly above 40% before all control rod scram time testing was l complete. , Upon discovering that the plant had exceeded 40% thermal power, licensed  ; personnel completed the scram time test on control rod 26-31, and core thermal l power decreased as expected to approximately 38.3% of rated. The Shift i Technical Advisor then initiated a Deficiency Card to document the event in  ; accordance with the plant's administrative control procedures. Subsequently,  ! licensed personnel partially inserted one rod group to provide additional margin , below 40%, and then successfully completed scram time testing on the rest of the control rods. The total time during which the plant operated above 40% was , approximately two minutes. CAUSE OF EVENT 3 ( The cause of this event was personnel error on the part of the licensed j superintendent of shift. One of the procedures in use at the time, l 34GO-0PS-001-2S, " PLANT STARTUP," contained a note to the effect that control i rod drive scram time testing is normally to be performed when the plant is operating at approximately 35% of rated thermal power in order to avoid the i I l l 1 1

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EXPIRES: 4/3D/92 TEXT CONTINUATION FACILITY hAME (1) DOCKET NUMBER (2) LER NUMBER ($) PAGE (3) 1 EAR 5EQ hum REV PIANT E. I. HAIM, INIT 2 05000366 93 003 00 3 0F 5 lEXT possibility of exceeding 40% during testing. However, the plant was operated at a higher power level (38.5%) with the result that the withdrawal of a high worth rod caused core thermal power to exceed 40%. REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CFR 50.73 (a)(2)(i) because the plant entered a condition which was prohibited by the Technical Specifications. Specifically, prior to completion of all control rod scram time testing as required by Unit 2 Technical Specifications section 4.1.3.2.a. core thermal power exceeded 401 for  ; a period of approximately two minutes. l The Control Rod Drive system provides control of reactivity through positioning 1 of control rods. The system consists of 137 cruciform-shaped control rods and  ! related hydraulic equipment to move them. The rods contai.n boron or hafnium ( based compounds which absorb neutrons and thus retard the nuclear chain reaction. E all the control rods are fully inserted into the reactor core,  ! sufficient ne w ive reactivity is present to shut down the reactor and maintain i it in the shutdown condition under worst case combinations of moderator density j and temperature. Each control rod is supplied with motive force by its own i hydraulic control unit. The scram function of the CRD system normally causes [ rapid, full insertion of all control rods simultaneously; however,-individual control rods may be scrammed one at a time for testing purposes. If the core l has been altered or shut down for 120 days or more, the scram function of all rods is required to be tested prior to ascending to full power operation. The  ! purpose of this testing is to ensure that the CRD system is capable of inserting l all control rods within the time constraints assumed in the plant's Final Safety  ! Analysis Report (FSAR). The method of test involves fully. withdrawing each control rod one at a time, actuating its scram function, and measuring the time interval between the initiation signal and the completion of rod travel. This = scram time test was in progress at the time of the event. f In this event, licensed shift personnel were operating the plant at a power level of 38.5% of rated while CRD scram time testing was being conducted, i However, when rod 26-31, which is located near the center of the core, was fully l withdrawn during testing, thermal power rose slightly above the 40% limit, l remaining there for approximately two minutes. When licensed shift personnel -{ observed Process Computer indications that thermal power had exceeded 40%, they . j scrammed rod 26-31, reducing core thermal power to below 40% of rated. Thus, l I the brief excursion above 40% rated thermal poser occurred as a result of the scram time testing, not as a result of a normal ascension toward full power  ! operation. The plant was in steady state operation below 40% power prior to the  ; event and returned to operat 6n below 40% power as the scram function on rod  ; 26-31 was tested. The 401 1 6 .t does not form the basis for any assumption in j the FSAR nor was it chosen to protect any safety limit. Therefore, the plant  ; did not exceed or potentially exceed any safety limit as a result of this event.  ; i The operability of plant equipment was not affected by the event, and the. l ability of the CRD system to pr3 duce a scram was not compromised in any way as  ! demonstrated by the fact that all control rods were successfully tested prior to l l i 

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TEAR !5EQ hum EEV PIRC E. I. HATCH, UNIT 2 05000366 93 003 00 4 0F 5 IEAT normal ascension to full power operation. Following the event, one rod group was partially inserted into the core, reducing reactor power slightly and providing additional margin between the operating regime and the 40% limit. Thus, there were no further power excursions above 40% of rated until all control rod scram time testing had been successfully completed. Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety. The a:falysis is applicable to all plant conditions. CORRECTIVE ACTIONS Corrective actions for this event included: ,

1. Counseling the licensed superintendent of shift. This action is complete.
2. Revising procedures 34GO-OPS-001-2S, 34GO-OPS-001-1S, 42SV-C11-001-1S, and 42SV-C11-001-2S to provide improved administrative control of reactor power during CRD scram time testing. This action will be completed by 07/16/93.

t ADDITIONAL INFORMATION

1. Other Systems Affected: No plant systems other than the CRD system were affected by this event.
2. Failed Components Identification: No failed components contributed to or resulted from this event.
3. Previous Similar Events: Events reported in the past two years in which ,

personnel arror resulted in the plant entering a condition which is prohibited by the Technical Specifications are described in the following LERs: - 50-321/1991-32, dated 01/27/92 l 50-321/1992-02, dated 02/06/92 50-321/1992-08, dated 04/20/92 50-321/1992-19, dated 08/04/92 . 50-366/1991-16, dated 06/28/91 . l 50-366/1991-21, dated 12/04/91  ! 50-366/1992-04, dated 04/30/92 50-366/1992-06, dated 06/22/92 { 50-366/1992-11, dated 08/14/92 50-366/1992-12, dated 08/25/92 50-366/1992-17, dated 10/21/92 50-366/1992-22, dated 12/07/92 50-366/1992-24, dated 12/14/92 50-366/1992-25, dated 12/21/92 P t

u.s. NWm AnMW uminiA V6MD DE WJ 3nD-lim (E-B,dem a s EXPIRES: 4/30/22 . LICENSEE EVENT REPORT (LER) 1 TEXT CONTINUATION  ; LER NUMBER (5) PAGE (3) I FACILITY %AME (1) DDCKET WUMBER (2) VEAR SEO hum kev FIANT E. I. liKIG, INIT 2 05000366 93 003 00 $ OF 5 IEAT , Corrective actions which addressed the personnel errors contributing to these events included counseling personnel, reviewing events with licensed personnel in Beginning of Shift Training (BOST), reviewing and revising procedures, issuing an Operations Departmental Directive on the purpose of daily instrument channel checks, training plant personnel on the use of self-verification techniques, removing a licensed individual from duty, f carrying out formal discipline under the Company's Positive Discipline program, and performing management observation of personnel performance in the Main Control Room. Although the event described in this LER involved  ; personnel error, these corrective actions would not have prevented this l event because they pertain to the performance of different activities. . Also, counseling or training personnel cannot completely eliminate , oversights such as the one that caused this event.  ! t i i t}}

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