Information Notice 2012-21, Reactor Vessel Closure Head Studs Remain Detensioned During Plant Startup: Difference between revisions

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{{#Wiki_filter:ML12264A518 December 10, 2012 NRC INFORMATION NOTICE 2012-21: REACTOR VESSEL CLOSURE HEAD STUDS REMAIN DETENSIONED DURING PLANT STARTUP
{{#Wiki_filter: ML12264A518 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION OFFICE OF NEW REACTORS WASHINGTON, DC  20555-0001  December 10, 2012   NRC INFORMATION NOTICE 2012-21: REACTOR VESSEL CLOSURE HEAD STUDS REMAIN DETENSIONED DURING PLANT STARTUP


==ADDRESSEES==
==ADDRESSEES==
All holders of an operating license or construction permit for a nuclear power reactor under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vesse All holders of or applicants for an early site permit, standard design certification, standard design approval, manufacturing license, or combined license under 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."  
All holders of an operating license or construction permit for a nuclear power reactor under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
 
All holders of or applicants for an early site permit, standard design certification, standard design approval, manufacturing license, or combined license under 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."


==PURPOSE==
==PURPOSE==
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees of an event involving detensioned reactor vessel closure head studs at a boiling-water reactor that resulted in leakage from the reactor vessel during startup operations and a manual scra The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problem Suggestions contained in this IN are not NRC requirements; therefore, no specific action or written response is require
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees of an event involving detensioned reactor vessel closure head studs at a boiling-water reactor that resulted in leakage from the reactor vessel during startup operations and a manual scram.  The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.  Suggestions contained in this IN are not NRC requirements; therefore, no specific action or written response is required.


==DESCRIPTION OF CIRCUMSTANCES==
==DESCRIPTION OF CIRCUMSTANCES==
Brunswick Steam Electric Plant, Unit 2 On November 16, 2011, the Brunswick Steam Electric Plant, Unit 2 (Brunswick 2, a General Electric boiling-water reactor) was in power ascension following a mid-cycle maintenance outage, which required reactor vessel disassembl Following the outage, the reactor vessel was reassembled and operators commenced startup operation With the reactor in Startup Mode (Mode 2) and at normal operating pressure, operators noted increasing drywell floor drain leakag At 3:01 a.m. eastern standard time (EST), an Unusual Event was declared as a result of unidentified drywell leakage exceeding 10 gallons per minute (gpm). At 3:09 a.m. EST, a manual reactor scram was initiated from approximately 7 percent of rated thermal power due to the continued increase in unidentified drywell leakag Following the scram, the reactor was depressurized and the unidentified leak rate decreased to less than 10 gpm within 1 hou At 1:45 p.m. EST on November 17, 2011, with the reactor in Cold Shutdown (Mode 4), leak investigation activities determined that the reactor vessel head studs were not fully tensioned during startup operations; therefore, an unanalyzed condition existed at Brunswick Subsequently, it was determined that none of the 64 reactor vessel head studs were adequately tensione Reactor vessel head stud tensioning is accomplished by attaching a tensioning device to the stud's uppermost thread Hydraulic pressure is applied to the tensioning device, which stretches the stu With the stud elongated by the tensioning device, personnel rotate the stud nut until it makes firm contact with the washer on the head flang When the hydraulic pressure is released, the nut maintains the tension and elongation in the stud, applying closure pressure to the flanges of the reactor vessel and hea The licensee's investigation determined that this event was the result of errors made while operating the reactor vessel head stud tensioning equipment and during the validation process to ensure the head was properly tensione Following the event, the licensee assessed the stud tensioning process through equipment troubleshooting, review of the reactor vessel reassembly procedure (Procedure 0SMP-RPV502), and interviews with refuel floor personne The equipment was found to be fully functiona However, the licensee determined that personnel operating the stud tensioning equipment misinterpreted the digital display of the hydraulic pressure being applied to elongate the stud Specifically, the licensee found that personnel incorrectly believed that the actual hydraulic pressure being applied to the tensioning device was a factor of ten greater than the pressure indicated on the devic As a result, none of the 64 studs were properly tensioned during the reactor vessel assembly proces The Stud Elongation Measurement System (SEMS III) is used at Brunswick 2 to validate proper stud elongatio Based on interviews with personnel, the licensee determined that the refuel floor crew incorrectly concluded that the target stud elongation value of 0.045 inches was achieved when the elongation values indicated on the SEMS III device were only between +/-0.004 inche The licensee attributed this error to the crew incorrectly assuming that the elongation value of 0.045 inches was automatically deducted from the post-tensioned elongation indication on the SEMS III devic Furthermore, the elongation values of +/-0.004 inches, as indicated on the SEMS III device, correspond to the stud elongation tolerance specified in Procedure 0SMP-RPV50 Accordingly, the crew compared the low reading on the SEMS III device to the stud elongation tolerance in the procedure and erroneously determined that acceptable stud elongation had been achieve The quality control inspector concurred with the consensus opinion of the cre As a result of these errors, the reactor vessel head studs were tensioned to only approximately 10 percent of the required amoun Therefore, Brunswick 2 reached Mode 2 with the head not properly tensione The increase in leakage and subsequent reactor scram were a direct result of this conditio The licensee performed a post-event evaluation of the integrity of the reactor vessel closure component The licensee concluded that no reactor coolant pressure boundary components were damaged or overstressed as result of the even After completing the integrity evaluation, the reactor vessel was reassemble Prior to plant restart, a hydrostatic test was completed to verify that proper head stud tensioning had been achieve The licensee attributed the root cause of this event to the failure to provide proper training and lack of procedure guidance to correctly interpret critical data used to validate that the reactor vessel head studs are properly tensione Specifically, the licensee concluded that the operator errors that occurred during the reactor vessel reassembly evolution were due to an inadequate understanding of the digital readings displayed on the hydraulic stud tensioning equipment and the SEMS III stud elongation measurement devic For both cases, the licensee determined that the crews relied on erroneous assumptions that led to incorrect conclusion Licensee Corrective Actions The licensee revised the reactor vessel reassembly procedure (Procedure 0SMP-RPV502) to include detailed guidance on the proper use of the SEMS III stud elongation measurement equipment and the interpretation of hydraulic pressure indications on the stud tensioning devic The licensee also provided training to refuel floor crew personnel on the proper operation of the SEMS III and hydraulic stud tensioning equipment during reactor vessel reassembl The licensee has revised its refuel floor training and qualification documents to include specific discussion on the correct operation of the SEMS III equipment and how to properly interpret hydraulic pressure indications on the stud tensioning devic The licensee has also revised Procedure 0SMP-RPV502 to require the necessary level of training regarding these activities, as provided in the revised training document In addition, the licensee has modified corporate-wide qualification programs for nuclear fleet refuel floor crew personnel to ensure that all refuel floor personnel at Brunswick maintain the necessary qualifications for performing their assigned activities and receive the necessary level of training on the SEMS III and stud tensioning equipment, as provided in the revised training document The licensee noted that, prior to the event, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outage Therefore, as a corrective action, the licensee revised plant procedures to require a pressure test of the reactor vessel following mid-cycle maintenance outages that require reactor vessel reassembl NRC Special Inspection Team Findings An NRC special inspection team reviewed the circumstances surrounding this even The inspection team reviewed the licensee's actions prior to the event and identified examples of improper procedure adherence that contributed to the inadequate reactor vessel head stud tensionin Specifically, the team determined that licensee personnel failed to properly pressurize the reactor vessel head stud tensioning equipment to the value specified in Procedure 0SMP-RPV502 because the tensioning equipment operators did not know how to correctly interpret the hydraulic pressure reading on the tensioning equipment displa The inspection team also determined that quality control personnel failed to verify proper reactor vessel stud elongation in accordance with stud elongation values specified in Procedure 0SMP-RPV50 Further, the inspection team determined that nine of the twelve refuel floor personnel performing reactor vessel reassembly did not have the necessary refuel floor support training, as required by Procedure TRN-NGCC-1000, "Conduct of Training." Finally, based on its review of Procedure 0PLP-20, "Post Maintenance Testing Program," which specifies "plant equipment shall be tested consistent with their safety functions following maintenance activities that may have impaired proper functioning of the components," the inspection team determined that the licensee failed to specify an adequate post maintenance test to verify the pressure retaining capability of the reactor vessel following a mid-cycle maintenance outag The Brunswick Steam Electric Plant, Unit 2, Licensee Event Report (LER) 50-324/2-2011-002, dated January 16, 2012, contains further discussion of this even The LER is available on the NRC's public Web site under Agencywide Documents Access and Management System (ADAMS) Accession No. ML12031A16 Additional information is available in NRC Special Inspection Report 05000324/2011013, dated January 25, 2012, under ADAMS Accession No. ML120250556; and NRC Inspection Report 05000324/2012007, dated April 20, 2012, under ADAMS Accession No. ML12114A03
Brunswick Steam Electric Plant, Unit 2 On November 16, 2011, the Brunswick Steam Electric Plant, Unit 2 (Brunswick 2, a General Electric boiling-water reactor) was in power ascension following a mid-cycle maintenance outage, which required reactor vessel disassembly.  Following the outage, the reactor vessel was reassembled and operators commenced startup operations.  With the reactor in Startup Mode (Mode 2) and at normal operating pressure, operators noted increasing drywell floor drain leakage.  At 3:01 a.m. eastern standard time (EST), an Unusual Event was declared as a result of unidentified drywell leakage exceeding 10 gallons per minute (gpm). At 3:09 a.m. EST, a manual reactor scram was initiated from approximately 7 percent of rated thermal power due to the continued increase in unidentified drywell leakage.  Following the scram, the reactor was depressurized and the unidentified leak rate decreased to less than 10 gpm within 1 hour.  At 1:45 p.m. EST on November 17, 2011, with the reactor in Cold Shutdown (Mode 4), leak investigation activities determined that the reactor vessel head studs were not fully tensioned during startup operations; therefore, an unanalyzed condition existed at Brunswick 2.  Subsequently, it was determined that none of the 64 reactor vessel head studs were adequately tensioned.  Reactor vessel head stud tensioning is accomplished by attaching a tensioning device to the stud's uppermost threads.  Hydraulic pressure is applied to the tensioning device, which
 
stretches the stud.  With the stud elongated by the tensioning device, personnel rotate the stud nut until it makes firm contact with the washer on the head flange.  When the hydraulic pressure is released, the nut maintains the tension and elongation in the stud, applying closure pressure to the flanges of the reactor vessel and head.
 
The licensee's investigation determined that this event was the result of errors made while operating the reactor vessel head stud tensioning equipment and during the validation process to ensure the head was properly tensioned.  Following the event, the licensee assessed the stud tensioning process through equipment troubleshooting, review of the reactor vessel reassembly procedure (Procedure 0SMP-RPV502), and interviews with refuel floor personnel.  The equipment was found to be fully functional.  However, the licensee determined that personnel operating the stud tensioning equipment misinterpreted the digital display of the
 
hydraulic pressure being applied to elongate the studs.  Specifically, the licensee found that personnel incorrectly believed that the actual hydraulic pressure being applied to the tensioning device was a factor of ten greater than the pressure indicated on the device.  As a result, none of the 64 studs were properly tensioned during the reactor vessel assembly process.  The Stud Elongation Measurement System (SEMS III) is used at Brunswick 2 to validate proper stud elongation.  Based on interviews with personnel, the licensee determined that the refuel floor crew incorrectly concluded that the target stud elongation value of 0.045 inches was achieved when the elongation values indicated on the SEMS III device were only between +/-0.004 inches.  The licensee attributed this error to the crew incorrectly assuming that the elongation value of 0.045 inches was automatically deducted from the post-tensioned elongation indication on the SEMS III device.  Furthermore, the elongation values of +/-0.004 inches, as indicated on the SEMS III device, correspond to the stud elongation tolerance specified in Procedure 0SMP-RPV502.  Accordingly, the crew compared the low reading on the SEMS III device to the stud elongation tolerance in the procedure and erroneously determined that acceptable stud elongation had been achieved.  The quality control inspector concurred with the consensus opinion of the crew.  As a result of these errors, the reactor vessel head studs were tensioned to only approximately 10 percent of the required amount.  Therefore, Brunswick 2 reached Mode 2 with the head not properly tensioned.  The increase in leakage and subsequent reactor scram were a direct result of this condition.
 
The licensee performed a post-event evaluation of the integrity of the reactor vessel closure components.  The licensee concluded that no reactor coolant pressure boundary components were damaged or overstressed as result of the event.  After completing the integrity evaluation, the reactor vessel was reassembled.  Prior to plant restart, a hydrostatic test was completed to verify that proper head stud tensioning had been achieved. The licensee attributed the root cause of this event to the failure to provide proper training and lack of procedure guidance to correctly interpret critical data used to validate that the reactor vessel head studs are properly tensioned.  Specifically, the licensee concluded that the operator errors that occurred during the reactor vessel reassembly evolution were due to an inadequate understanding of the digital readings displayed on the hydraulic stud tensioning equipment and the SEMS III stud elongation measurement device.  For both cases, the licensee determined
 
that the crews relied on erroneous assumptions that led to incorrect conclusions.  Licensee Corrective Actions The licensee revised the reactor vessel reassembly procedure (Procedure 0SMP-RPV502) to include detailed guidance on the proper use of the SEMS III stud elongation measurement equipment and the interpretation of hydraulic pressure indications on the stud tensioning device.  The licensee also provided training to refuel floor crew personnel on the proper operation of the SEMS III and hydraulic stud tensioning equipment during reactor vessel reassembly.  The licensee has revised its refuel floor training and qualification documents to include specific discussion on the correct operation of the SEMS III equipment and how to properly interpret hydraulic pressure indications on the stud tensioning device.  The licensee has also revised Procedure 0SMP-RPV502 to require the necessary level of training regarding these activities, as provided in the revised training documents.  In addition, the licensee has modified corporate-wide qualification programs for nuclear fleet refuel floor crew personnel to ensure that all refuel floor personnel at Brunswick maintain the necessary qualifications for performing their assigned activities and receive the necessary level of training on the SEMS III and stud tensioning equipment, as provided in the revised training documents.  The licensee noted that, prior to the event, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outages.  Therefore, as a corrective action, the licensee revised plant procedures to require a pressure test of the reactor vessel following mid-cycle maintenance outages that require reactor vessel reassembly.  NRC Special Inspection Team Findings An NRC special inspection team reviewed the circumstances surrounding this event.  The inspection team reviewed the licensee's actions prior to the event and identified examples of improper procedure adherence that contributed to the inadequate reactor vessel head stud tensioning.  Specifically, the team determined that licensee personnel failed to properly pressurize the reactor vessel head stud tensioning equipment to the value specified in Procedure 0SMP-RPV502 because the tensioning equipment operators did not know how to correctly interpret the hydraulic pressure reading on the tensioning equipment display.  The inspection team also determined that quality control personnel failed to verify proper reactor vessel stud elongation in accordance with stud elongation values specified in Procedure 0SMP-RPV502.  Further, the inspection team determined that nine of the twelve refuel floor personnel performing reactor vessel reassembly did not have the necessary refuel floor support training, as required by Procedure TRN-NGCC-1000, "Conduct of Training." Finally, based on its review of Procedure 0PLP-20, "Post Maintenance Testing Program," which specifies "plant equipment shall be tested consistent with their safety functions following maintenance activities that may have impaired proper functioning of the components," the inspection team determined that the licensee failed to specify an adequate post maintenance test to verify the pressure retaining capability of the reactor vessel following a mid-cycle maintenance outage.  The Brunswick Steam Electric Plant, Unit 2, Licensee Event Report (LER) 50-324/2-2011-002, dated January 16, 2012, contains further discussion of this event.  The LER is available on the NRC's public Web site under Agencywide Documents Access and Management System (ADAMS) Accession No. ML12031A167.  Additional information is available in NRC Special Inspection Report 05000324/2011013, dated January 25, 2012, under ADAMS Accession No. ML120250556; and NRC Inspection Report 05000324/2012007, dated April 20, 2012, under ADAMS Accession No. ML12114A036.


==DISCUSSION==
==DISCUSSION==
Section 50.120, "Training and qualification of nuclear power plant personnel," of 10 CFR states, in part, that the training program must incorporate the instructional requirements necessary to provide qualified personnel to operate and maintain the facility in a safe manner in all modes of operatio Criterion V, "Instructions, Procedures, and Drawings," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 states, in part, that instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplishe The root cause of this event was the failure to provide the necessary training and procedure guidance to correctly interpret critical indications on the stud tensioning and stud elongation measurement equipment for verifying that proper stud tensioning had been achieve The failure to adequately tension the reactor vessel closure head studs during reactor vessel reassembly undermined the integrity of the reactor coolant pressure boundary, one of the primary barriers to fission product release, during startup operation In addition, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outage However, the reactor vessel head was removed and reinstalled during this outage in the same fashion as during a refueling outag Therefore, this event highlights the importance of conducting mid-cycle maintenance outage activities, particularly those that require reactor vessel disassembly and reassembly, with the same level of rigor as scheduled refueling outage activitie This event also highlights the importance of human performance and oversight of maintenance activitie For example, operators of the stud tensioning equipment were not familiar with the pressure display, yet they proceeded with tensioning based on an incorrect interpretation of indicated tensioner pressur In addition, a licensee lead mechanic and a quality control inspector signed a procedure checklist for stud elongation measurements using flawed data, based on incorrect explanations by other members of the maintenance cre Other findings related to human performance can be found in the April 20, 2012, inspection repor
Section 50.120, "Training and qualification of nuclear power plant personnel," of 10 CFR states, in part, that the training program must incorporate the instructional requirements necessary to provide qualified personnel to operate and maintain the facility in a safe manner in all modes of operation.  Criterion V, "Instructions, Procedures, and Drawings," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 states, in part, that instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
 
The root cause of this event was the failure to provide the necessary training and procedure guidance to correctly interpret critical indications on the stud tensioning and stud elongation measurement equipment for verifying that proper stud tensioning had been achieved.  The failure to adequately tension the reactor vessel closure head studs during reactor vessel
 
reassembly undermined the integrity of the reactor coolant pressure boundary, one of the primary barriers to fission product release, during startup operations.  In addition, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outages.  However, the reactor vessel head was removed and reinstalled during this outage in the same fashion as during a refueling outage.  Therefore, this event highlights the importance of conducting mid-cycle maintenance outage activities, particularly those that require reactor vessel disassembly and reassembly, with the same level of rigor as scheduled refueling outage activities.  This event also highlights the importance of human performance and oversight of maintenance activities.  For example, operators of the stud tensioning equipment were not familiar with the pressure display, yet they proceeded with tensioning based on an incorrect interpretation of indicated tensioner pressure.  In addition, a licensee lead mechanic and a quality control inspector signed a procedure checklist for stud elongation measurements using flawed data, based on incorrect explanations by other members of the maintenance crew.  Other findings related to human performance can be found in the April 20, 2012, inspection report.


==CONTACT==
==CONTACT==
This IN requires no specific action or written respons Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manage /RA/ /RA by JLuehman for/ Timothy J. McGinty, Director Laura A. Dudes, Director Division of Policy and Rulemaking Division of Construction Inspection Office of Nuclear Reactor Regulation and Operational Programs Office of New Reactors Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Librar
This IN requires no specific action or written response.  Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manager.  /RA/       /RA by JLuehman for/ Timothy J. McGinty, Director   Laura A. Dudes, Director Division of Policy and Rulemaking   Division of Construction Inspection Office of Nuclear Reactor Regulation     and Operational Programs       Office of New Reactors   Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov
 
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library.


==CONTACT==
==CONTACT==
This IN requires no specific action or written respons Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manage /RA/ /RA by JLuehman for/
This IN requires no specific action or written response.  Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manager.
Timothy J. McGinty, Director Laura A. Dudes, Director Division of Policy and Rulemaking Division of Construction Inspection Office of Nuclear Reactor Regulation and Operational Programs Office of New Reactors Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail: Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Librar ADAMS Accession No.: ML12264A518 *via e-mail TAC No. ME8863 OFFICE EVIB:NRR AHPB:NRR Tech Editor BC:EVIB:NRR BC:AHPB:NRR NAME CSydnor* MKeefe* CHsu* SRosenberg* UShoop* DATE 11/27/12 11/27/12 10/11/12 11/27/12 11/27/12 OFFICE D:DE PM:PGCB:NRR LA:PGCB:NRR BC:PGCB:NRR D:DCIP:NRO NAME PHiland (MCheok for) TAlexion CHawes DPelton (EBowman for) LDudes (JLuehman for) DATE 11/27/12* 11/28/12 11/29/12 11/27/12* 12/03/12 OFFICE D:DPR:NRR NAME TMcGinty OFFICE 12/10 /12 OFFICIAL RECORD COPY}}
 
/RA/       /RA by JLuehman for/
Timothy J. McGinty, Director   Laura A. Dudes, Director Division of Policy and Rulemaking   Division of Construction Inspection Office of Nuclear Reactor Regulation     and Operational Programs       Office of New Reactors
 
Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail: Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov
 
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library.
 
ADAMS Accession No.: ML12264A518                 *via e-mail                   TAC No. ME8863 OFFICE EVIB:NRR AHPB:NRR Tech Editor BC:EVIB:NRR BC:AHPB:NRR NAME CSydnor* MKeefe* CHsu* SRosenberg* UShoop* DATE 11/27/12 11/27/12 10/11/12 11/27/12 11/27/12 OFFICE D:DE PM:PGCB:NRR LA:PGCB:NRR BC:PGCB:NRR D:DCIP:NRO NAME PHiland (MCheok for) TAlexion CHawes DPelton (EBowman for) LDudes (JLuehman for) DATE 11/27/12* 11/28/12 11/29/12 11/27/12* 12/03/12 OFFICE D:DPR:NRR     NAME TMcGinty     OFFICE 12/10 /12     OFFICIAL RECORD COPY
 
}}


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Revision as of 10:38, 28 March 2018

Reactor Vessel Closure Head Studs Remain Detensioned During Plant Startup
ML12264A518
Person / Time
Issue date: 12/10/2012
From: Dudes L A, McGinty T J
Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking
To:
Alexion, T
References
IN-12-20
Download: ML12264A518 (6)
v • d • e


ML12264A518 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION OFFICE OF NEW REACTORS WASHINGTON, DC 20555-0001 December 10, 2012 NRC INFORMATION NOTICE 2012-21: REACTOR VESSEL CLOSURE HEAD STUDS REMAIN DETENSIONED DURING PLANT STARTUP

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power reactor under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.

All holders of or applicants for an early site permit, standard design certification, standard design approval, manufacturing license, or combined license under 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees of an event involving detensioned reactor vessel closure head studs at a boiling-water reactor that resulted in leakage from the reactor vessel during startup operations and a manual scram. The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Suggestions contained in this IN are not NRC requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

Brunswick Steam Electric Plant, Unit 2 On November 16, 2011, the Brunswick Steam Electric Plant, Unit 2 (Brunswick 2, a General Electric boiling-water reactor) was in power ascension following a mid-cycle maintenance outage, which required reactor vessel disassembly. Following the outage, the reactor vessel was reassembled and operators commenced startup operations. With the reactor in Startup Mode (Mode 2) and at normal operating pressure, operators noted increasing drywell floor drain leakage. At 3:01 a.m. eastern standard time (EST), an Unusual Event was declared as a result of unidentified drywell leakage exceeding 10 gallons per minute (gpm). At 3:09 a.m. EST, a manual reactor scram was initiated from approximately 7 percent of rated thermal power due to the continued increase in unidentified drywell leakage. Following the scram, the reactor was depressurized and the unidentified leak rate decreased to less than 10 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. At 1:45 p.m. EST on November 17, 2011, with the reactor in Cold Shutdown (Mode 4), leak investigation activities determined that the reactor vessel head studs were not fully tensioned during startup operations; therefore, an unanalyzed condition existed at Brunswick 2. Subsequently, it was determined that none of the 64 reactor vessel head studs were adequately tensioned. Reactor vessel head stud tensioning is accomplished by attaching a tensioning device to the stud's uppermost threads. Hydraulic pressure is applied to the tensioning device, which

stretches the stud. With the stud elongated by the tensioning device, personnel rotate the stud nut until it makes firm contact with the washer on the head flange. When the hydraulic pressure is released, the nut maintains the tension and elongation in the stud, applying closure pressure to the flanges of the reactor vessel and head.

The licensee's investigation determined that this event was the result of errors made while operating the reactor vessel head stud tensioning equipment and during the validation process to ensure the head was properly tensioned. Following the event, the licensee assessed the stud tensioning process through equipment troubleshooting, review of the reactor vessel reassembly procedure (Procedure 0SMP-RPV502), and interviews with refuel floor personnel. The equipment was found to be fully functional. However, the licensee determined that personnel operating the stud tensioning equipment misinterpreted the digital display of the

hydraulic pressure being applied to elongate the studs. Specifically, the licensee found that personnel incorrectly believed that the actual hydraulic pressure being applied to the tensioning device was a factor of ten greater than the pressure indicated on the device. As a result, none of the 64 studs were properly tensioned during the reactor vessel assembly process. The Stud Elongation Measurement System (SEMS III) is used at Brunswick 2 to validate proper stud elongation. Based on interviews with personnel, the licensee determined that the refuel floor crew incorrectly concluded that the target stud elongation value of 0.045 inches was achieved when the elongation values indicated on the SEMS III device were only between +/-0.004 inches. The licensee attributed this error to the crew incorrectly assuming that the elongation value of 0.045 inches was automatically deducted from the post-tensioned elongation indication on the SEMS III device. Furthermore, the elongation values of +/-0.004 inches, as indicated on the SEMS III device, correspond to the stud elongation tolerance specified in Procedure 0SMP-RPV502. Accordingly, the crew compared the low reading on the SEMS III device to the stud elongation tolerance in the procedure and erroneously determined that acceptable stud elongation had been achieved. The quality control inspector concurred with the consensus opinion of the crew. As a result of these errors, the reactor vessel head studs were tensioned to only approximately 10 percent of the required amount. Therefore, Brunswick 2 reached Mode 2 with the head not properly tensioned. The increase in leakage and subsequent reactor scram were a direct result of this condition.

The licensee performed a post-event evaluation of the integrity of the reactor vessel closure components. The licensee concluded that no reactor coolant pressure boundary components were damaged or overstressed as result of the event. After completing the integrity evaluation, the reactor vessel was reassembled. Prior to plant restart, a hydrostatic test was completed to verify that proper head stud tensioning had been achieved. The licensee attributed the root cause of this event to the failure to provide proper training and lack of procedure guidance to correctly interpret critical data used to validate that the reactor vessel head studs are properly tensioned. Specifically, the licensee concluded that the operator errors that occurred during the reactor vessel reassembly evolution were due to an inadequate understanding of the digital readings displayed on the hydraulic stud tensioning equipment and the SEMS III stud elongation measurement device. For both cases, the licensee determined

that the crews relied on erroneous assumptions that led to incorrect conclusions. Licensee Corrective Actions The licensee revised the reactor vessel reassembly procedure (Procedure 0SMP-RPV502) to include detailed guidance on the proper use of the SEMS III stud elongation measurement equipment and the interpretation of hydraulic pressure indications on the stud tensioning device. The licensee also provided training to refuel floor crew personnel on the proper operation of the SEMS III and hydraulic stud tensioning equipment during reactor vessel reassembly. The licensee has revised its refuel floor training and qualification documents to include specific discussion on the correct operation of the SEMS III equipment and how to properly interpret hydraulic pressure indications on the stud tensioning device. The licensee has also revised Procedure 0SMP-RPV502 to require the necessary level of training regarding these activities, as provided in the revised training documents. In addition, the licensee has modified corporate-wide qualification programs for nuclear fleet refuel floor crew personnel to ensure that all refuel floor personnel at Brunswick maintain the necessary qualifications for performing their assigned activities and receive the necessary level of training on the SEMS III and stud tensioning equipment, as provided in the revised training documents. The licensee noted that, prior to the event, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outages. Therefore, as a corrective action, the licensee revised plant procedures to require a pressure test of the reactor vessel following mid-cycle maintenance outages that require reactor vessel reassembly. NRC Special Inspection Team Findings An NRC special inspection team reviewed the circumstances surrounding this event. The inspection team reviewed the licensee's actions prior to the event and identified examples of improper procedure adherence that contributed to the inadequate reactor vessel head stud tensioning. Specifically, the team determined that licensee personnel failed to properly pressurize the reactor vessel head stud tensioning equipment to the value specified in Procedure 0SMP-RPV502 because the tensioning equipment operators did not know how to correctly interpret the hydraulic pressure reading on the tensioning equipment display. The inspection team also determined that quality control personnel failed to verify proper reactor vessel stud elongation in accordance with stud elongation values specified in Procedure 0SMP-RPV502. Further, the inspection team determined that nine of the twelve refuel floor personnel performing reactor vessel reassembly did not have the necessary refuel floor support training, as required by Procedure TRN-NGCC-1000, "Conduct of Training." Finally, based on its review of Procedure 0PLP-20, "Post Maintenance Testing Program," which specifies "plant equipment shall be tested consistent with their safety functions following maintenance activities that may have impaired proper functioning of the components," the inspection team determined that the licensee failed to specify an adequate post maintenance test to verify the pressure retaining capability of the reactor vessel following a mid-cycle maintenance outage. The Brunswick Steam Electric Plant, Unit 2, Licensee Event Report (LER) 50-324/2-2011-002, dated January 16, 2012, contains further discussion of this event. The LER is available on the NRC's public Web site under Agencywide Documents Access and Management System (ADAMS) Accession No. ML12031A167. Additional information is available in NRC Special Inspection Report 05000324/2011013, dated January 25, 2012, under ADAMS Accession No. ML120250556; and NRC Inspection Report 05000324/2012007, dated April 20, 2012, under ADAMS Accession No. ML12114A036.

DISCUSSION

Section 50.120, "Training and qualification of nuclear power plant personnel," of 10 CFR states, in part, that the training program must incorporate the instructional requirements necessary to provide qualified personnel to operate and maintain the facility in a safe manner in all modes of operation. Criterion V, "Instructions, Procedures, and Drawings," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 states, in part, that instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

The root cause of this event was the failure to provide the necessary training and procedure guidance to correctly interpret critical indications on the stud tensioning and stud elongation measurement equipment for verifying that proper stud tensioning had been achieved. The failure to adequately tension the reactor vessel closure head studs during reactor vessel

reassembly undermined the integrity of the reactor coolant pressure boundary, one of the primary barriers to fission product release, during startup operations. In addition, a decision was made that a post maintenance reactor vessel pressure test was not necessary because there are no regulatory requirements to conduct this test following mid-cycle maintenance outages. However, the reactor vessel head was removed and reinstalled during this outage in the same fashion as during a refueling outage. Therefore, this event highlights the importance of conducting mid-cycle maintenance outage activities, particularly those that require reactor vessel disassembly and reassembly, with the same level of rigor as scheduled refueling outage activities. This event also highlights the importance of human performance and oversight of maintenance activities. For example, operators of the stud tensioning equipment were not familiar with the pressure display, yet they proceeded with tensioning based on an incorrect interpretation of indicated tensioner pressure. In addition, a licensee lead mechanic and a quality control inspector signed a procedure checklist for stud elongation measurements using flawed data, based on incorrect explanations by other members of the maintenance crew. Other findings related to human performance can be found in the April 20, 2012, inspection report.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manager. /RA/ /RA by JLuehman for/ Timothy J. McGinty, Director Laura A. Dudes, Director Division of Policy and Rulemaking Division of Construction Inspection Office of Nuclear Reactor Regulation and Operational Programs Office of New Reactors Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors project manager.

/RA/ /RA by JLuehman for/

Timothy J. McGinty, Director Laura A. Dudes, Director Division of Policy and Rulemaking Division of Construction Inspection Office of Nuclear Reactor Regulation and Operational Programs Office of New Reactors

Technical Contacts: Christopher R. Sydnor, NRR Molly J. Keefe, NRR 301-415-6065 301-415-5717 E-mail: E-mail: Christopher.Sydnor@nrc.gov Molly.Keefe@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library.

ADAMS Accession No.: ML12264A518 *via e-mail TAC No. ME8863 OFFICE EVIB:NRR AHPB:NRR Tech Editor BC:EVIB:NRR BC:AHPB:NRR NAME CSydnor* MKeefe* CHsu* SRosenberg* UShoop* DATE 11/27/12 11/27/12 10/11/12 11/27/12 11/27/12 OFFICE D:DE PM:PGCB:NRR LA:PGCB:NRR BC:PGCB:NRR D:DCIP:NRO NAME PHiland (MCheok for) TAlexion CHawes DPelton (EBowman for) LDudes (JLuehman for) DATE 11/27/12* 11/28/12 11/29/12 11/27/12* 12/03/12 OFFICE D:DPR:NRR NAME TMcGinty OFFICE 12/10 /12 OFFICIAL RECORD COPY


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