B16740, Reissues LERs 97-031-00 & 97-014-02 Which Includes Rept Dates That Were Inadvertently Ommited in Previous Rept
| ML20216H651 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/09/1997 |
| From: | Hinnenkamp P NORTHEAST NUCLEAR ENERGY CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| B16740, NUDOCS 9709170010 | |
| Download: ML20216H651 (3) | |
Text
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. Nuclear Energy unioor wi,or po er si.is,o
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'Ihr Northcant Utilitics Systern SEP 91997 Docket No. 50-245 B16740 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No.1 i Licensee Event Reports 97-031-00. 97-014-02 On August 29,1997, Northeast Nuclear Energy Company (NNECO) submitted two Licensee Event ReportsWS (LERs) for Millstone Nuclear Power Station, Unit !
No.1. For both LERs, NNECO inadvertently omitted the report dates (Block 7).
- This letter forwards LERs 97-031-00 and 97-014-02 with the appropriate report dates. These LERs supersede those previously transmitted.
There are no commitments containpd within_ this letter, Very truly yours,
- NORTHEAST NUCLEAR ENERGY COMPANY A -
/\ M em A P. D. Hinnenkamp ' V Director, Unit Operations r
Attachment:
LER 97-031-00 i LER 97-014-02 g3-) ;
N P. D. Hinnenkamp letter to U.S. Nuclear Regulatory Commission, "LER 97-031-00," dated August 29,1997.
- P. D. Hinnenkamp letter to U.S. Nuclear Regulatory Commission, "LER 97-014-02," dated August 29,1997. !
97091700 9 PDR- ADOCK 05000245:
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U.S. Nucl:ar R:gulatory Commission B16740\Page 2 cc: H. J. Miller, Region I Administrator S. Dembek, NRC Project Manager, Millstone Unit No.1 T. A. Eastick, Senior Resident inspector, Millstone Unit No.1
l
. Docket No. 50-245 816740 Attachment 1 Millstone Nuclear Power Station, Unit No.1 Licensee Event Reports (LER) 97-031-00 and 97-014-02 September 1997
to C #5siTA6 U.S. NUCLE AR REGULATONY COMMIS$10N APPaon,D[ gp00604 tff#.'w*y*/A"4Wy' PWM.*""iMJo'$tM; LICENSEE EVENT REPORT (LER) ,,e%*%W**
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- 10tuf V NAMI III opCtifN W DIh Gl PA04 tal Millstone Nuclear Power Station Unit 1 05000245 1 of 3 fittt 446 Standby Liquid Control Pump Discharge Pressure Surveillance LVIa DATE(s) LtR NUMstR ist Rf PORT DATE (71 OTHER FACILIT!!s INVotVED (s)
MONTH DAY itAR YEAR $EQUEN11AL REVISION MONT H DAY YEAR # Actutv NAM Docatr N w ein NUMBER 97 '"""*" '*"""
07 l30 97 031 00 08 29 97 OPER ATING III THis REPORT is susMITTED PURSUANT TO THE REQUIREMENTS OF 10 CrR la (Check one or enote) III)
POWER 20.2201tbl 20.2203(aH H 20 2203(aH2){vi X 60aat H2Ho 60.73<aH2Hvna 20.2203(aH3)l0 50.731aH2H10 6033(aH2Hx)
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NAMt 1(LtPHONt NUM614 Dnclude Assa Codel Hobert W. Walpole, Mp1 Nuclear Licensing Manager (8601440 2191 CAubt COMPLETE 6vsttM COMPONLN1 ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
MANUF AC1Uf4R HLPQHT ABL t CAubt BY$f tM COMPONtNT MANUf ACTUHtR HlPOM1 ADLE T0 NPHDS TO NPRDS sVPPLEMENTAL REPORT EXPECTED (14l EXPECTED MONTH DAY vtAR Yts susMisslON
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(if yes, complete EXPECTED SUBMISSION DATil. f NO AssTRACT R6mit to 1400 spaces, i.w,, approumately 16 single-spaced typewntten lines) (16) i On July 30,1997, at 1400 hsurs, with the plant in COLD SHUTDOWN, while performing a 50,54(f) design and hcensing basis review of the Standby Liquid Control (SBLC) system,it was determined that the surveillance test for the SBLC pumps did not meet Technical Specification (TS) 4.4.A,1 requirements. The TS requires that the pumps must have a capacity of at least 40 ppm at a system head corresponding to a reactor vessel pressure of 1225 psig, Th3 current test procedure requires a pressure of 1225 psig at the pump discharges. This acceptance criteria value for the pump discharge pressure does not consider the system line losses to the vessel, any differences in elevation b3 tween the SBLC pumps discharge pressure gauge md the vessel nozzle, or instrument inaccuracy.
Th3 cause of this condition was development of the surveillance procedure with incorrect acceptance criteria due to an in:dequate procedure review process Contributing causes to this event are the failure to identify this discrepancy in previous TS compliance reviews, and the inadequacy of the TS to capture plant operating requirements. The proctdure acceptance criteria did not include the system pressure losses identified in the system design description.
No immediate corrective actions were needed since the SBLC is not required to be in service. The system design bisis will be reviewed to determine the correct pemp discharge pressure acceptance criteria and the Standby Liquid Cor: trol system pumps will be tested using the revised proceduto prior to operating cycle 16.
NEC toRM 366 to 9M 0b Y '
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NHC f;RJ 3f*6A U.S. nucle AR RtouLATORY commission i
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACIL11Y NAML til DOCKET NuMsER (2) if R NuMstR 161 PAof 13) i[AH SEQulNTIAL RE VISION Millstone Nuclear Power Station Unit 1 05000245 wovst n NuMBf H 2 of 3 97 -
031 -
00 TLJT (11more space is requurd use oddaroonalcopies of NRC form 366A) t11)
- 1. Dagn!stion of Event On July 30,1997, at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, with the isnt in a Cold "r.utdoW1 condition, while performing a 50.54(f) d: sign and licensing basis review of the Standby Liquid Control (SBLC) system IBRl, it was determined that the surveillance test for the SDLC pumps did not meet Technical Specification (TS) 4,4.A.1 requirements. TS 4.4.A.1 states, in part, that
- Pump minimum Cow rate of 40 ppm shall be verified against a system pressure corresponding to a reactor vessel pressure of 1225 psig.* While potforming a 50.54(f) review of the Standby Liquid Control (SBLC) system (BR), it was determined that the surveillance test for the SDLC pumps did not meet TS 4.4.A.1 requirements. This condition is reportable pursuant to 10CFR50.73(a)(2)(i) as a condition prohibited by the TS.
The TS requires the operabikty of the SBLC system to be verified at least once per month. Operability of the SBLC system is accomplished by recirculating domineralized water to the test tank. The recirculation stop valve I is throttled to obtain a pressure of 1225 psig on the discharge pressure gauge for the positive displacement pump. Pump minimum flow rate of 40 gpm should be verified against a system head corresponding to a reactor v:ssel pressure of 1225 psig.
The General Electric system design description includes a margin of 25 psi differential pressure drop due to system head loss from the pump discharge to the reactor vessel. The procedule acceptance critoria value for th3 purnp discharge pressure does not consider the system line losses to the vessel, any differences in elevation b tvveen the SBLC pumps discharge pressure gauge and the vessel nonie, or instrument inaccuracy,
- 11. CrutadJvent The cause of this condition was development of the surveillance procedure with incorrect acceptance criteria due to an inadequate procedure review process. Contributing causta to this event are the f ailure to identify this discrepancy in previous TS compliance reviews, and the inadequacy of the TS to capture plant operating r:quirements. The procedure acceptance criteria did nr.t include the system pressure losses identified in the system design description.
Ill. Anolysis of Event The SDLC pumps are positive displacement pumps rated for a discharge pressure of 1500 psig. The pre-op; rational test for the SBLC pumps verified operation at a discharge pressure of 1300 psig and a flow rate of 44.94 ppm. The SBLC pumps would be used during an Anticipated Transient Without Scram condition.
The safety significance of this condition is that the surveil!ance procedure did not verify pump operation at the discharge pressure required by TS 4.4.A.1. However, there is reasonable asseance that the SBLC pumps could have performed their design function, based on the 1500 psig design discharge pressure of the SBLC system piping and the purups, and the pre operational test results.
NAC roMM 3r4A to 9tl m
NRc f oRJ 366A is ou U.s. NUctiAR EEOULA109V COMMisslON LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION f ACillTY NAML til DOCKET NUMBER (21 LER NUMBER 16) PAof (3)
)(AR 6['Qu[rvilAL RL VISION Millstone Nuclear Power Station Unit 1 05000245 NUMar H NMf A 3 of 3 97 --
031 00 TlKT ttimwr space is requard, use add:toonaltotnes of MrC form 366Al 1111 l IV. Correctivo Actiqu No immedinto corrective actions were needed since the SDLC is not required to be in service.
A review is being performed for Technical Specification surveillance requirements banod on previous commitments for the Licensee Event Reports hated as similar events. Millstone has revised the administrativo procedure controls for the procedure review process. The revised administrative process requires additional tzview stops during the development of now procedures or proceduto revisions to identify tho source document which is the basis for the proceduto steps and the acceptance critoria.
Tha system dosion basis will be reviewed to datormine the correct pump discharge pressure acceptance critoria und
- 10. the Standby Liquid Control system pumps will be tested using the revised proceduro prior to operating cycle V. Additional Informotion None.
Similar Events97-020, Liquid Radwesto Monitor Functional Test Surveillance Not in Accordance with Technical Specificatiot LER 90 044, High Range Stack Noble Gas Monitor inoperable Due to inadequate Cakbration LER 90 043, inadequato Instrument Calibrations Due to Failure to Verify Response Time LER 90-019, Sensor Calibration of SRM/lRM Nuclear instrumentation not Performed as Required LEh 90-015, Rocirculation Pump Flow Mismatch Surveillance not Performed in Accordance with Technical Specihcations LER 97 020, Liquid R$dwasta Effluent Monitor Functional Test Surveillance Not in Accordance with Technical Specifications LER 97-025, Single Loop Jet Pump Baselino Data Collection Missed Surveillance LER ObO29, GL 96-01 Review Discovers inadequate Surveillance Testin0 in Accordance with Technical Specifications Manufacturer Data None.
Energy industry Idontification System (Ells) codos are identified in the text as IXXI.
NHC roRM 3MA (4 99
NXC IORM 366 U.S. NUCLE AR REGUL A10HY COMMISSION APPaoveb sv oma iIo albo eto4 14 EM i*Peits 04/3 Wee l
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- digits / characters for each block) i l t Actuty esAus til Ducast stuutta ut Pact tai Millstono Nuclear Power Station Unit 1 05000246 1 of 6 l l
) trtLt 14i Certain Core Spray Components Not included in the IST Program EVENT DATE (61 LER NUMBER (b) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
MONTH DAY ' 'IAR YEAR SiOUENTIAL REVISION MONT H DAY YEAR 8 AciuTv NAug ooott Nuveen NUMBER 02 17 97 97 014 " "' " " " * " "
02 08 29 97 OPERATING THIS REPORT IS SUDMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR le (Check one or more) (til DE W 20.220l tbl 20.2203taH2Hv) X td.. 3(al(2Hil 60.73(aH2Hvust
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20.2203(aH2Hus) 60.36(cH1) 60.73taH2Hv) soecify in Abstrett below or in NHc Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME 1titPHONE NUMDIn unessede A<me codel Robert W. Walpolo, MP1 Nuclear Licensing Manager (860)440 2191 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRISED IN THIS REPORT (13)
CAubt sv611M COMPONENT MANur ACTUHtH RE PoM1 ABit CAU$t sv$rtM CoMPONt NT MANUr ACTUHtH HEPOH r ABL E Yo NPMos io NPHOs l m.,
SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONT H oAv vtAR YES .
Uf yea, complete EXPECTED SUBMISSION (JATE).
X NO SUBMISSION ABSTRACT (Limit to 1400 spaces. I e., approximately 16 s6noie spaced typewritten hnes) (16)
On February 17,1997, with the plant in COLD SHUTDOWN and reactor defueled, a review performed on the Inservice Testing UST)
Program, es part of the 10 Cf R 50.54(f) effort identified a historical IST Program deficiency, Core Spray valves 1 CS 13A/B were not included in the IST Program prior to 1992. Additionally, on February 18,1997, the same review identified that four Core Spray system check valves located on the keep fill lines were not leak tested in the IST Program to support verification of the system's intigrity as a closed system as credited in the 10CFR50 Appendia J program. On February 25,1997, additional IST program deficiencies were identified. On Ju!y 30.- 1997, several IST program deficiencies, which were identified during the 50.54(f) testsament cf the IST program, were discovered as being omitted from the previous condition reporting process. l Tto cause for this event is a lack of management commitment to support the IST Program by not allocating the resources to implement a successful IST Program ' No immediate action is required since the plant is in COLD SHUTDOWN and the reactor d3 fueled.
.c The IST Program manual already has been revised to include the core spray minimum flow line check valves. The IST surveillance dificiencies and other IST Prograrn deficiencies will be corrected. NNECO will provide sufficient trained staffing to implement and mtnage the IST Program. NNECO will revise the lST Program Administrative Control Procedure based on the 10CFR50.54tfl findings. Any applicable corrective actions as a result of this review will be completed prior to startup for operating Cycle 10.
- NHC FoHM 36e 14 9H .
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leRC ioRM 36sA U.s. NUCLEAR RioVLATo3Y Commission LICENSEE EVENT REPORT (LER)
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TEXT CONTINUATION F ACILITY NAME 11) DoCEti NUM&iR (2) LER NuMsLR tal PAoE (3)
'iTAR SEQUENTIAL REViS60N Millstone Nuclear _ Power Station Unit 1 05000245 NUMBIR NV,Mj(R. 2 of 8 97 - 014 -
02 TEK1 lit enore space is required, use addorionniecques of NRC form 366A) (11) i
- 1. Description of Event On February 17,1997, with the plant in COLD SHUTDOWN and reactor defueled, a review being performed on tha inservice Testing (IST) Program, as part of the 10CFR50.54(f) effort identified a historical IST Program dsficiency. Core Spray IBM) valves 1-CS 13A/B were not included in the IST Program prior to 1992. The fact that -these valves were not included in the IST Program is reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B), "any operation or condition prohibited by the plant's Technical Specifications," because thsse valves were not tested as required by Technical Specification 4.13 prior to the 1992 revision of the IST Program.
Additionally, on February 18, 1997, with the plant in COLD SHUTDOWN and teactor defueled, review being pztlormed on the IST, as part of the 10CFR50.54(f) effort identified that four Core Spray system check valves 1 CS 10A/B and 1 CS 20 A/B, located on the keep fill lines, were not leak tested in the IST Program to support varification of the systum's integrity as a closed system as credited in it e 10CFR50 Appendix J program. They were only exercise tested. The program review detoimined that there were several *.nstances where the Core Spray system is documented as a ' closed loop
- system for containment isolation purposes. This is not entirely-correct. The Core Spray system's keep fill connection check valves are tested (disassembled and inspected) only for an Indication of gross diversion of flow. Tim isolation valves for this branch line are required to be leakage tested to ensure system integrity. They are not presently leakage tested. This issue is reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(BL *any operation or condition prohibited by the plant's Technical Spscifications," because these valves are inadequately tested according to their safety function.
On February 25,1997, witn the plant in COLD SHUTDOWN and reactor defueled, a review performed on the IST Program,6 part of the 10CFR50.54(f) ef fort, identified additional deficiencies in the IST Program:
Inadeounte IST Surveillances:
- 1. For valvon 1 CS-4A/B and 1 LP 9A/B, the surveillance procedures do not contain stroke time acceptance criteria as required by the IST Program.
2._ For standby liquid control (SBLC) (BR] pumps M8 55A/B, the surveillonce directs acquisition of vibration data prior to the Code required hold time to allow pump conditions to stabilize,
- 3. For Core Spray pumps M8 74A/B, the Code required hold time is not specified in the quarterly surveillance procedures to allow pump conditions to stabilize prior to recording data.
'4. Surveillance procedure for Condensate Booster ISD] pumps M2 7A/B does not throttle flow to a reference value as required by relief request R 1.
- 5. Quarterly testing of the Low Pressure Coolant injection 1801 check valves 1 LP 3A/B/C/D is required. Based on GL 89 04 criteria.-maximum accident' flow must be verified. From the LOCA Analysis 5405 gpm is required. Current testing only requires a minimum of 5000 gpm be verified.
~ 6 SBLC- storage tank level is ' not measured to the accuracy required to meet the instrument accuracy requirements of the Code (+ 2%).
- NRC FORM 366A (4 99
l NEC FOR] 3:6A 14 st) U.S. NUCLEAR r.tOUL ATORY Commission UCENSEE EVENT REPORT (LER)
. TEXT CONTINUATION f ACill1Y NAME til ' Docket NUMsER (2) LtR NUMBER (4) PAGE13)
YEAR SEQUENilAL HLVis!cTa Millstone Nuclear Power Station Unit 1 05000245 NUMBf R NUMBf R 3 of 6 97 -
014 -
02 TEXT Ilf more spete is requoted, use additionalcopws of NRC form 366Al (11)
- 7. Testing of TIP shear valves does not comply with OM 10 paragraph 4.4.1 in that 20% of the charges are not fired und replaced within the 2 year required time period.
B. Surveillance procedure for Low Pressure Coolant injection pumps does not throttle pump differential pressure to a specified reference value.
- 9. Testin0 of service water (Bil valve 1 SW 99 does not comply with OM 10 requirements. Stroke timing is not obtained from initiation of actuating signal. Valve stem movement is timod locally.
- 10. Instrumentation used in refuel performance testing of the condensato and condensate booster pumps does not meet the Code full scale range requirements (OM.0 paragraph 4.0.1.2)
AW1tLone.Lhstinti Not includ2d in the IST Prggrea_rD
- 1. Low Pressure Coolant Injection keep fill check valves 1 LP-46/47 nood to be Category A valves and leak tested to support verification of systern integrity as a closed system as credited in the Appendix J Program.
Compong,iltLNot included in thojS.,T Prontam
- 3. Turbine Building Secondary Closed Cooling Water (TBSCCW) system IKB) control valves 1 SC-7A/B,1-SC-29A/B/C/D and 1 SC 47.
On July 30,1997, several IST program deficiencies, which were identified during the 50.54(f) assessment of tha IST program, were discovered as being omitted from the previous condition reporting process. This discovery occurred upon completion of the engineering self assessment report which included a review of the cpproximately 500 items identified as part of the 10CFR50.54(f) effort. Three additional categ',ry B passive valves in the Reactor Recirculation system (AD) 1 RR-4A,1 RR 48, and 1 RR 58, were identifbd as requiring position indication testing and need to be included in the IST Program. It was also noted that the surveillance procedures for the Core Spray pumps do not throttle pump flow to the specified reference value within the required plus or minus 2 percent tolerance.
- 11. C.ttqge of Event The cause for this event is a lack of management commitment to support the IST Program by not allocating the resources to implement a successful IST Program.
111. Analysis of Event These conditions are being reported pursuant to the requirements of 10 CFR 50.73(a)(2)(i)(B) which identifies any operation or condition prohibited by the plant's Technical Specifications. Technical Specification 4.13 requires I
NRC roRM 3f b A (4 95)
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NIC f oR'J 366A U.S. NUCLEAR K[oVLAf oRY Commission 14 961 !
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, , TEXT CONTINUATION FACILiiY NAME (1) DOCKET NUMsER (2) LER NUMsER 16) PAoE13)
YEAR . S[QUE NTIAL REVI5loN ,
Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 4 of 0 ;
97 -
014 -
02 l
TEXT Illmore spose is required. Use additional copies of MIC form 366A) (111 that inservice testing of ASME Code Class 1,2 and 3 equivalent components shall be performed in accordance !
- with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by {
10CFR50.55algl, '
During the 10CFR 50.54(f) review of the IST Program, the following was identified: i r
1 CS.13A/S The IST Program is intended to test the operational readiness of certain ASME Code class 1, 2, and 3 t components required to shutdown the reactor to saie shutdown, maintain safe shutdown, or mitigate the consequences of an accident. The components in question are minimum flow check valves for the Core Spray l system. These minimum flow check valves have a safety function to open, providing a minimum flow path f preventing pump damage.; The safety function of the minimum flow check valves is from initial design, which meant that they met the scope statement of the original IST Program. Technical Specification 4.13 requires .
Millstone Unit No.1 to comply with the requirements of IST. The f act that these components were not included in the IST Program is reportable pursuant to the requirements of 10CFRLO 73(a)(2Hl)(B) because the components in' question were not tested as required by Technical Specification 4.13 prior to the 1992 revision of the IST Program. ;
1 CS 19A/B and 1-CS 20A/B !
The program review determined that there were several instances where the Core Spray system is documented as a ' closed loop' system for containment isolation purposes. This is not entirely correct. A ' closed loop'
. containment isolation system is defined in 10CFR50 Appendix A, General Design Criteria 57. The Core Spray system does not meet this definition, nor is this definition being imposed on the system. NUREG 0800 Section ;
6.2,4.ll.0e provides an option for a single containment isolation valve where the system is closed outside of containment (not a ' closed loop'),' This requires that this closed system be tested to ensure system integrity. The Core Spray system's keep fill connection is an untested branch line off the otherwise closed system. The isolation valves for this branch line are required to be leakage tested to ensure system integrity. They are not presently leakage tested.
This condition is reportable because the valves in question were incorrectly categorized when the IST Program was scopedi Category 'C' test verifies gross diversion of process flow and a category 'A' or 'A/C' test imposes .
a leakage requirement. These valves are presently categorized as 'C' only and should be 'A/C'. This test has not !
been performed . and constitutes a missed surveillance per - Technical- Specification 4.13. Also, Technical Specification 0.13 requires a leakage program be in place to monitor certa:n systems for leakage. The Core Spray system meets the criteria to be included in this program, however, this leakage path is not monitered.
There were no safety consequent'es as a result of this event. The safety significance of this event is negligible based on the following:
- i
-e For the core spray keep fill valves, dual barricts are provided where only one is required and current testing ensured leakage la minimal compared to a failed valve. 4 o Current testing confirmed the valve's capability to function (disassembly and inspection confirmed the condition of the valve and freedom of movement of the valve disc). All the subject valves have been !
- inspected. y l
NRC FORM 366A 14 961 l'
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- ~ - , ::,r,, .v/ 25, 1997 LER 90&d 'I
" ' u 4 te pus w ,mr4 if any of the findi, and on July 30,review 1997,
"*** a thorough of T
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wee r cf echnical LER 90 03" N
+ s w ww .unt time none of the associated systems tr,e systems is not affected at thi ae are opn erable or are required oes edentihad on February s time.
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,ruary 25,1997, 25,1997 and There were no safety
{ t(!!!
w on July 30,1997.
i ~**--$ m are that LPCI, SBLC and TBSCCW The safety may not have l ilone \
EnetM i %+" mem 't e piart is in cold i y
w %. " " m shutdown and the teactor defueled. I v" -
msed to include the coro spray minimum flo ,
, s ,py,, wf ?n ,r4cate that w line check valves.
- 3 g e J and a surveillance d, for the core spray keep -
1 f
eveloped to test the valves prior to startup
' ^
e
% netura systems e.operabl** ee wa-iMca orocedures as ne m ,.y .
- - % ,,,." wy,u. *m to correct the identified (ST ce survein emy,,.n e at ttoral rerts to existing surveillance procedure e
- ..w m ,, est requirements for components currently r or in the m% myy.y,I m u wstmg surve ll
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m ,, ,er me new components added ti ance procedures or write ne e ures if needed to
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- 1 ~ v g,w* w o the IST Program prior to declaring
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- y. ,wea*M will t:e endorsed bee ranng to implement and ma
, , , y the program management as well a m mq f s ,y%, w unit i
xc 7 ym m. ..a ::mgram control requirement 4ttwill be end,m~::tnre w4 orsed by the program) findings prior my we-e ver management w, ww ~wm ec as part of the current 10CFR50 54(f) .
n~m md be completed prior to startup for review effort. Any applicable operating Cycle 16.
,~
marm.:ceu n c t e insemce test program as earece usrer rupe benows within containment pa result of proDramm f A
- enetration test program qqt $pH5J M id i
_ _.__-__.- --J
l NRc FORM 366A 44 961 U.s. NUCLEAR REoVLAf oRY Commission l
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION F ACillTY NAME til DOCKET NuMsER (2) LER NUMsER (6) PAGE (3)
YEAR SEQUENTIAL REVIsloN Millstone Nuclear Power Station Unit 1 05000245 NUMBFR NUMBE R 4 og g j 97 --
014 -
02 l TEK1 ft!more space is required, use additional copies of NRC form .166A) (11) thtt inservice testing of ASME Code Class 1, 2 and 3 equivalent components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as requ' red by 10CFR50.55alg).
During the 10CFR 50.54(f) review of the IST Program, the following was identified:
1 CS 13A/B The IST Program is intended to test the operatinnal readiness of certain ASME Code class 1, 2, and 3
- components required to shutdown the reactor to safo shutdown, maintain safe shutdown, or mitigate the consequences of an accident. The components in question are minimum flow check valves for the Core Spray system. These minimum flow check _ valves have a safety function to open, providing a minimum flow path preventing pump damage. The safety function of the minimum flow check valves is from initial design, which msant that they met the scope statement of the original IST Program. Technical Specification 4.13 requ!res Millstone Unit No.1 to comply with the requirements of IST. The fact ' hat these components were not included in the IST Program is reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) because the components in quection were not tested as required by Technical Specification 4.13 prior to the 1992 revision of the IST Program, o
1-CS 19A/B and 1-CS 20A/B The program review deterinined that there were several instances where the Core Spray system is documented as a ' closed loop' system for containment isolation purposes. This is not entirely correct, A ' closed loop' containment isolation system is defined in 10CFR53 Appendix A, General Design Criteria 57, The Core Spray system does not meet this dafinition, nor is th!s definition being imposed on the system. NUREG 0800 Section 6.2.4.ll.6e provides an option for a single containtient isolation valve where the system is closed outside of containment (not a ' closed loop'). This requires that this closed system be tested to ensure system integrity. The Core Spray system's keep fill connection is an untested branch line off the otherwise closed system. The isolation valves for this branch line are required to be leakage tested to ensure system integrity. They are not presently leakage tested, l
This condition is reportable because the valves in question were incorrectly categorized when the IST Program ;
was scoped. Category 'C' test verifies gross divers.lon of process flow and a category 'A' or 'A/C' test imposes a laakage requirement. These valves are presently categorized as 'C' only and should be 'A/C'. This test has not
-)i bs2n performed and constitutes a missed surveillance per Technical Specification 4.13. Also, Technical Spscification 6.13 requires a leakage program be in place to monitor certain systems for leakage. The Core Spray l system meets the siteria to be included in this program, however, this leakage path is not monitored. '
Thsre were no safety consequences as a result of this event, The safety significance of this event is negligible bxsed on the following:
- For the core spray keep fill valves, dual barriers are provided where only one is required and current testing ensured leakage is minimal compared to a failed valve.
- Current testing confirmed the valve's capability to function (disassembly and inspection confirmed the condition-of the valve and freedom of movement of the valve- disc). All the subject valves - have been inspected.
NRC roRM 366A to 95)
, e NRC FORM 364A U.s, NUCLEAR REoVLATORY Commission UCENSEE EVENT REPORT (LER)
. . TEXT CONTINUATION FACILITY NAME lt) OOcKET NuMsER (2) LER NuMsER (s) PAoE 131 YEAR SEQUENTIAL FtE VISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 5 of 6 97 --
014 - 02 TEKT ritmore space is required. Use addtional copes of NRC form 366Al (11)
For the deficiencies identified or' ec bruary 25, 1997, and on July 30,1997, a thorough review of Technical l Specifications was performed to determine if any of the findings would impact the operability of associated
) systems. It was concluded that at the present time none of the associated systems are operable or are required l to be operablei Therefore, operability of the systems is not affected at this time. There were no safety consequences as a result of the deficiencies identified on February 25,1997 end on July 30,1997. The safety implications of th6 deficiencies identified on February 25,1907, are that LPCI, SBLC and TBSCCW may not have
, bssn able to provide their safety functions, l t
- IV, Corrective Aqugn No immediate action la required since the plant is In cold shutdown and the reactor defueled.
l - The IST Program manual already has been revised to include the core spray ininimum flow line check valves.
The IST Program manual will be revised to indicate that, for the core spray keep fill line check valves (1 CS-19A/B & 1-CS 20A/B), the category is 'A/C,' and a surveillance developed to test the valves prior to startup for operating Cycle 16.
NNECO will revise the applicable surveillance procedures as needed to correct the identified IST surveillance inadequacies prior to declaring the applicable systems operable.
NNECO will add the additional test requirements to existing surveillance procedures or write new procedures if nstded to implement the identified additional test requirements for components currently in the IST Program prior to declaring the applicable systems operabie.
NNECO will add the test requirements to existing surveillance procodures or write new procedures if needed to implement the identified test requirements for the new components added to the IST Program prior to declaring the applicable systems operable, i
NNECO will provide sufficient trained staffing to implement and manage the IST Program prior to startup for optrating Cycle 16. This staffing requirement will be endorsed by the program management as well as the unit msnagement.
NNECO will revise the IST Program Administrative Control Procedure based on the 10CFR50.54(f) findings prior !
to startup for operating Cycle 16. This program control requirement will be endorsed by the program management j as well as tho unit management.
The IST Program review is being performed as part of the current 10CFR50.54(f) review effort. Any applicable corrective actions as a result of this review will be completed prior to startup for operating Cycle 16.-
V. Additional Information Similar Events LER 96-021 Components not included in the inservice test program as a result of programmatic deficiencies LER 96 023 Failure to include fuel transfer tube bellows within containment penetration test program N# C f oRM 366A (4-96)
NOC 70HM 366A U,s, NUCLEAR REGULATOQY COMMISSION 149M LICENSEE EVENT REPORT (LER)
, , TEXT CONTINUATION FACillTY NAME til DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
N(AH SE QU[ NTIAL RE VISION Millstone Nuclear Power Station Unit 1 05000245 NuMar R NuMm 0 of 6 ;
97 -
014 -
02 '
TEXT (11rnore space os required, use additionalcoturs of NRC form 366A) t1T)
LER 90 024 Missed Inservice testing of a stop check valve in the steam supply to the turbine driven auxiliary feed water pump due to programmatic weakness LER 90-039 in Service Testing of Volves Not Performed in Accordance with Program and ASME Code l Requirements Menufacturer Data None l
Energy industry identification System (Ells) codes are identified in the text as IXXI, l
NRC f oMM 366A 14 SM