05000400/LER-1997-021, :on 980210,discovered That SFP Water Level Had Not Been Verified Greater than 23 Feet Above BWR Fuel Assemblies.Caused by Misinterpretation of TS Requirements. Will Submit TS Change Request to Revise TS 3.9.1.11
| ML18016A470 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/30/1998 |
| From: | Verrilli M CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18016A469 | List:
|
| References | |
| LER-97-021, LER-97-21, NUDOCS 9807070218 | |
| Download: ML18016A470 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4001997021R00 - NRC Website | |
text
NRC FORM 366 l4 95)
U.S.
AR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
A VED BY OMB NO. 3150-0104 EXPIRES 04/30I98 ESTIMATED BURDEN PER
RESPONSE
TO COMPLY WITH THIS MANDATORY INFORMATION COHECTION REDDEST: 50.0 HRS.
REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE UCENSUIG PROCESS ANO FEO BACK TO INDUSTRY.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION ANO RECORDS MANAGEMENTBRANCH IT 6 F33), U.S. NUCLEAR REGULATORY COMMISSION, VIASHINGTON, OC Z05554001, AND TO THE PAPERWORK REOUCRON PROJECT (3150 0)04L OFFICE OF MANAGEIJIENT ANO BUDGET. WASHINGTON, DC 20503.
FACILITYNAME (1)
Harris Nuclear Plant Unit-1 DOCKET NUMBER (2) 50-400 PAGE (3) 1OF5 TITLE(4)
Technical Specification Surveillance Procedure Review Project identified Deficiencies.
MONTH OAY 2
10 YEAR 98 EVENT DATE (5)
LER NUMBER (6)
SEQUENTIAL REVISiON NUN'IBER NUMBER 97
021
03 MONTH OAY 6
30 YEAR 98 REPORT DATE (7)
FACILITYNAME FACIUTYNAME OTHER FACILITIES INVOLVED(8)
OOCKETNUMBER DOCKET NUMBER 05000 OPERATING MODE (9)
POWER LEVEL (10) 100or THIS REPORT IS SUBMITTED PUR 20.2201(b) 20.2203(9)(1) 20.2203(e) (2) (i) 20.2203(e) (2) (ii) 20.2203(al(2)(iii) r more)
(11)
SUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one o 50.73(a) (2) (I) 50.73(B) (2)(ii) 20.2203(B) (2) (v) 20.2203(B)(3)(i) 20.2203(a) (3) (ii) 20.2203(e) (4) 50.36(c)(1) 50.73(a) (2) (iii) 50.73(a) (2) (iv) 50.73(a)(2)(v) 50.73(a)(2)(viii) 5O.73(B)(21(x) 73.71 OTHER Specify in Abstract below or In NRC Form 366A
- 20. 2203(e) (2) (Iv) 50.36(c) (2)
LICENSEE CONTACT FOR THIS LER (12) 50.73(a) (2) (vii)
NAME Michael Verrilli Sr. Analyst - Licensing COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES TELEPHONE NUMBER (Include Ares Code)
'I (919) 362-2303 CRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MAIJUFACTURER REPORTABLE TO )JPROS
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABI.E TO NPROS SUPPLEMENTAL REPORT EXPECTED (14)
YES (If yes, complete EXPECTED SUBMISSION DATE).
NO EXPECTED SUBMISSION DATE (15)
MONTH OAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., BpProximately 15 single-spaced typewritten lines)
(16)
On August 14, 1997, with the plant at approximately 100% power in mode 1, a condition was identified during the on-going Technical Specification (TS) surveillance procedure review project related to inadequate maintenance of Spent Fuel Pool water level.
Specifically, Technical Specification (TS) 3/4.9.11 requires that "at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks."
This depth of water will provide sufficient "scrubbing" to remove 99%
of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
Contrary to this requirement, water level has not been venfied greater than 23 feet above the boiling water reactor (BWR) fuel assemblies received from CP&L's Brunswick Plant, which are currently stored in the Harris Plant fuel pools.
These BWR assemblies have a bail handle that extends approximately 6 inches above the top nozzle base plate.
When the BWR storage racks were installed in 1991, the 23 foot water level reference mark was established from the top nozzle base plate of the BWR fuel seated in the storage racks, not from the top of the bail handles.
This approach was determined at the time to be conservative since the base plate elevation exceeds that of the fuel rods which would be the source of any released fission gasses.
However, verbatim compliance with the TS requirements would require 23 feet of water over the BWR fuel assembly structure, including the top bail handle.
This condition was caused by a misinterpretation of TS requirements and design mputs during the establishment of the 23 foot water level reference mark and the subsequent setup of water level indicators, when the BWR fuel storage racks were initially installed at the Harris Plant.
Corrective actions included directions to Operations to maintain and monitor fuel pool level at or above 23 feet 7 inches to ensure required water level oyer the BWR bail handles.
This'was completed by issuing an Operations night order and revising the daily surveillance procedures.
Additional actions will include reviewing this event with appropriate Engmeering personnel to emphasize the importance of verbatim TS compliance and an evaluation of the fuel pool level alarm setpoints.
This revision is submitted to revise the due date for planned corrective action No. 1.
alt'807070~F8 qaoSSO PDR ADQC1(e 05000 8(4.96)
~
LlCEMSEE EVENT REPORT (LER)
TEXT CONTINUATION US. NUCLEAR REGUUITORY COMMISSION FACILITY NAIAE (I)
Shearon Harris Nuciear Plant
~ Unit N1 TEXT Pi more speceis reevved, vse edCh'tdmel copies of IIRC Farm 3664) (IT)
OOCKET 50 400 LER NUMBER (6)
YEAR SLOUEN'FIAL BEY)SION NUNIBER NU)ABER 97 021 03 PAGE (3) 2 OF 5
EVENT DESCRIPTION
On August 14, 1997, with the plant at approximately 100% power in mode 1, a condition was identified during the on-going Technical Specification (TS) Surveillance Procedure Review Project related to inadequate maintenance of Spent Fuel Pool water level.
Specifically, Technical Specification (TS) 3i4.9.11 requires that "at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks."
As described in the TS Bases section for this TS, this depth of water willprovide sufficient "scrubbing" to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
Contrary to this, water level has not been verified greater than 23 feet above the boiling water reactor (BWR) fuel assemblies received from CP&L's Brunswick Plant, which are currently stored in the Harris Plant (HNP) fuel pools.
These BWR assemblies haves bail handle that extends approximately 6 inches above the top nozzle base plate.
When the'BWR storage racks were installed in 1988, the 23 foot water level reference mark was established from the top nozzle base plate of the BWR fuel seated in the storage racks, not from the top of the bail handles.
This approach was determined at the time to be conservative since the top nozzle base plate elevation exceeds that of the fuel rods which would be the source of any released fission gasses.
However,,verbatim compliance with the TS requirements would require 23 feet of water over the BWR fuel assembly structure, including the top bail handle.
The method of verifying adequate water level in the Spent Fuel Pools at HNP involved confirming that the low-level alarm was not present.
The low level alarm setpoint was established at 23 feet 2.5 inches and was consistent with the 23 foot reference mark from the top nozzle base plate.
Therefore water levels could have dropped below 23 feet above the top of the BWR assembly bail handles and not result in a low level alarm.
The following additional Technical Specification Surveillance related deficiencies have been identified by the on-going comprehensive TS Surveillance Procedure Review Project.
The TS Surveillance Review project was originally committed to in LER 95-07, dated September 28, 1995.
On September 22, 1997, a design deficiency in'the Fuel Handling Building Emergency Exhaust System (FHBEES) was determined to be reportable.
Specifically, the FHBEES contains two units (E-12 86 E-13) which each consist of a fan, charcoal adsorber beds and HEPA filters. To prevent degradation of the charcoal bed removal efficiency, the units contain heaters to control the humidity of the air passing through the charcoal.
To prevent potential auto-ignition of the charcoal in the idle unit due to heat from the decay of radionuclides captured by the charcoal, the system is designed to provide. cooling flow or "bleed flow" (approximately 5% of total flow) through the idle unit. NRC Reg. Guide 1.52 and the HNP Final Safety Analysis Report (FSAR) section 6.5.1 indicate that the bleed flow passes from the discharge of the idle unit to the suction of the on-line unit. Any releases would thus be filtered through a charcoal filter with the appropriate design efficiency.
Contrary to this, the bleed flow on the FHBEES units passes from the discharge of the idle unit to the discharge of the on-line unit.
Since the heaters that control relative humidity do not run when the unit is not in service, the humidity in the idle unit.is not controlled.
Therefore, efficiency of the charcoal in the idle unit could potentially be degraded and air flowing through this idle unit could be filtered with an efficiency lower than the assumed 95% contained in the FSAR chapter 15 fuel handling accident analysis.
This would have resulted in higher calculated off-site doses had fuel movement occurred as assumed in the FSAR analysis and allowed by plant procedures.
Therefore, this condition is being voluntarily reported as a condition that could have resulted in operation outside the design basis of the plant per 10CFR50.73.a.2.ii.
2.
Also on September 22, 1997, a condition involving inadequate surveillance testing of the FHB and RAB Emergency Exhaust system charcoal adsorber beds was determined to be reportable.
Specifically, HNP NCFRM3 A(4
)(4.96)
LiCENSEE EVENT REPORT (LERj TEXT CONTINUATION U.S. NUClEAR RECU(ATORY COMMISSION FACI(n'Y NAME (I)
Shearon Harris Nuclear Plant
~ Unit )I'1 DOCKET
'0-400 LER NUMBER (6)
YEAR SEOUENTIA(
BEY)SION NUMBER NUMBER PACE (3)
TEXT Pl more spsseis roooped, ose eddi(drool copies ol/NC Form N641 (1))
==EVENT DESCRIPTION==97 021 03 3
OF 5
Technical Specifications require that a laboratory analysis be performed on the charcoal "after every 720. hours of charcoal a(fsorber operation".
The hours of operation have only been accumulated based on the time period that the on-line unit was in service.
The accumulation should also have included the time that bleed flow was passing through the idle unit.
This condition was considered to be a violation of TS surveillance requirements 4.7.6.c and 4.9.12.c.
On February 10, 1998, an additional TS testing deficiency was identified during the on-going TS Surveillance Review Project.
This deficiency involved past testing of the non-safety related Pressurizer Power Operated Relief Valve (PORV, 1RC-116, EIIS Code:AB-RV), which was not performed at the correct plant conditions per the requirements of TS surveillance requirement 4.4.4.1.
Specifically, TS 4.4.4.1.b states that "each PORV shall be demonstrated OPERABLE at least once per 18 months by: (b) Operating the valve through one complete cycle of full travel during MODES 3 or 4, prior to going below 325 degrees F". Testing to satisfy this surveillance requirement has been performed by the "Pressurizer Safety Grade PORV Operability - Quarterly Test" Operations Surveillance Test procedure (OST-1117).
Past performance of this test was credited for satisfying the entire TS surveillance requirement.
However, testing of 1RC-116, the non-safety related PORV, was not included in OST-1117.
1RC-116 has been tested by the "Pressurizer PORV - Quarterly Operability Test" (OST-1503) to satisfy In-service Inspection requirements.
However, OST-1503 is performed in MODE 5 with Reactor Coolant System temperature less than 200 degrees F.
Therefore, 1RC-116 has not been tested in the mode and plant conditions specified by TS 4.4.4.1.b.
The failure to include 1RC-116 in the scope of test procedure OST-1117, occurred during implementation of amendment 27 to the Harris Plant Operating License (TS) in September 1991'.
This amendment was generated to address the concerns stated in NRC Generic. Letter 90-06 "Power Operated Relief Valves and Block Valves in PWR Plants".
CAUSE
The original condition was caused by a misinterpretation of TS requirements and design inputs during the establishment of the 23 foot water level reference mark (above the BWR top nozzle base plate) and the subsequent setup of water level indicators, when the BWR fuel storage racks were initially installed at the Harris Plant.
Cause for Additional Items identified:
Item 1:
The cause of the FHBEES design deficiency was engineering oversight during initial plant design and construction Item 2:
The cause of the TS surveillance violation related to charcoal testing after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation was incorrect interpretation of TS testing requirements.
Surveillance test procedures were initially set up to satisfy the 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> requirement based on accumulation of hours for the in-service unit only.
Item 3:
The cause of the pressurizer PORV (1RC-116) testing deficiency was inadequate surveillance test procedures.
The changes made to TS 4.4.4.1 in amendment 27 were not understood by the plant staff and the associated development of OST-1117 did not ensure proper testing of 1RC-116.
SAFETY SIGNII ICANCE:
There were no actual safety consequences associated with this event.
Adequate water depth (23 feet) has been maintained above the active fuel rods, which would be the source of any released fission gasses.
This ensures the iodine removal capability required by TS in the event of a ruptured irradiated fuel assembly:
This condition is being reported per 10CFR50.73.a.2.i as a condition prohibited by Technical Specifications.
NRC M
A (4-
))4$5)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. IIUCLEAR REGULATORY COMMISSION FACILITYNAME II)
Y I
Shearon Harris Nuclear Plant
~ Unit A'1 DOCKET 50 400 LER NUMBER I6)
YEAR SPGUENTIAL REYISIN NUMBER NUMBER 97 -
021 03 PAGE I3) 4 OF 5
TEXT frfmore speceis ferro<fed, ose eddirior<sf copies of NRC Foim 3664r I)T)
==SAFETY SIGNIFICANCE==Safet Conse uences for Adtlitionnl Items Identified:
Item 1:
There were no actual safety consequences associated with this additional item. Had a fuel handling accident occurred in the FHB with the improperly configured bleed flow between the filtration units, the resulting off-site dose rates would not have exceeded the maximum value analyzed in the HNP FSAR and would have remained within 10CFR100 limits. This result is based on the fact that fuel off-load during past HNP refueling outages has never occurred prior to 257 hours0.00297 days <br />0.0714 hours <br />4.249339e-4 weeks <br />9.77885e-5 months <br /> following reactor shutdown, which allowed for substantial iodine decay resulting in a lower fuel handling accident source term than that assumed in the FSAR. It is also based on past charcoal efficiency surveillance testing, which has indicated a worst case value of 98%, which is greater than the 95% assumed efficiency value in the FSAR accident analysis.
Administrative controls have also been in place to ensure that filter charcoal is replaced ifsurveillance testing indicates an efficiency less than 99%
Therefore, this condition is being voluntarily reported as a condition that could have resulted in operation outside the design basis of the plant per 10CFR50.73.a.2.ii.
Item 2:
There were no consequences as a result of the TS surveillance requirement violation based on the charcoal surveillance test results described above.
This condition is being reported as a condition prohibited by Technical Specifications per 10CFR50.73.a.2.i Item 3:
There were no consequences as a result of the pressurizer PORV (1RC-116) testing deficiency. Testing has been satisfactorily performed per the requirements of TS 4.4.4.1.b to verify the operability of the two safety related pressurizer PORVs (1RC-114 and 1RC-118).
This testing ensures that both safety related PORVs would have been available to operators in the event of a Steam Generator tube rupture accident and for use as low temperature over-pressure protection.
This condition is being reported as a condition prohibited by Technical Specifications per 10CFR50.73.a.2.i.
PREVIOUS SIMILAREVENTS
There have been no previous events related to inadequate verification of Spent Fuel Pool water level or improperly configured bleed flow between air handling units as a result of design oversight.
HNP submitted LER 96-002 to report numerous deficiencies caused by incorrectly interpreting TS testing requirements during initial surveillance test development.
These were a result of HNP's actions to address NRC Generic Letter 96-01.
CORRE<CTIVE ACTIONS COMPLETED:
1.
Directions were provided to Operations to maintain and monitor Spent Fuel Pool water level at or above 23 feet 7 inches to ensure required water level over the BWR bail handles.
This was completed by issuing an Operations Night Order on August 14, 1997, revising the Reactor Auxiliary Building Operator Logs on August 25, 1997 and requiring the actual Spent Fuel Pool water level to be entered in the daily surveillance requirement test, rather than the previous practice of confirming the absence of the low level alarm.
2.
This event was reviewed with appropriate Engineering personnel to emphasize the importance of verbatim TS compliance.
This was completed on October 6, 1997.
CORRECTIVE< ACTIONS PLANNED:
1 A Technical Specification (TS) Change Request will be submitted to the NRC to revise TS 3.9.1.11 to require that the Spent Fuel Pool water level be maintained at least 23 feet over the top of the irradiated "fuel rods" and not "fuel assemblies".
The TS Chang'e request will be submitted prior to September 1, 1998.
NRC ORM A
I4
)
IIRC FORM 366A (4 S61 LtCENSEE EVE6lT REPORT (LER)
TEXT CONTINUATION US. NUClEAR REGUlATORY COMAUSSION FACILITYIIAME((i Shaaron Harris Nuclear Plant
~ Unit 0'1 TEXT P! repro spore rs receded, ose eddiriooe! copies o!rriICForre JSSAI (17]
OOCKET 50 400 LER NUMRER (6)
YEAR SEOUENTIAL REVISION NUMSER NUMSER 97 021 03 PAGE (3) 5 OF 5
Corrective Actions for Additional Items Identified:
Item 1:
A Justification for Continued Operation (JCO) was approved and implemented for the FHBEES bleed flow issue on October 10, 1997. A long term resolution for the improperly configured FHBEES bleed flowissue will be provided via Engineering Set+ice Request 97-00737.
This willensure that the systems satisfy the appropriate FSAR and Reg. Guide 1.52 design requirements and will be completed prior to the next fuel off-load.
Item 2:
As an interim measure, Operations personnel will record run times on both the in-service filtration unit and the idle filtration unit in the FHB and RAB Emergency Exhaust Systems.
This was directed by an Operations Night Order on October 15, 1997.
Revisions were completed to operations procedures that require recording run times for FHB and RAB Emergency Exhaust System filtration units (OP-170 "Fuel Handling. Building HVAC" and OP-172 "Reactor Auxiliary Building HVAC"). These revisions were completed on October 30, 1997.
Item 3:
Operations Surveillance Test (OST-1117 "Pressurizer Safety Grade PORV Operability - Quarterly Test") was revised (revision 6) to include testing of 1RC-116 at the appropriate plant conditions.
This was completed on April 23, 1998.
NR FORM 66A (4
EI