05000263/LER-2001-002
| Docket Number | |
| Event date: | 01-19-2001 |
|---|---|
| Report date: | 03-29-2001 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 2632001002R01 - NRC Website | |
Description documentation Edition been topworks.
review edition.
entered components entered 2001, shutdown within requirements.
shutdown Other entered other (If more space is required, use additional copies of NRC Form 366A) (17) January 19, 2001, while operating at 100% power, following a request by the NRC resident inspector for related to a snubber' replacement on the High Pressure Coolant Injection (HPCI)2 system, the Monticello became aware that the requisite NIS-2 form had not been generated as required by paragraph IWA-7520 of the of the ASME Code Section Xl. Further investigation revealed repair and replacement plans and NIS-2 forms had generated for replacement activities involving other ASME Code Section XI snubbers and safety-relief valve (SRV)3 It was also determined that the Authorized Nuclear Inservice Inspector (ANII) was not given the opportunity and approval of the repair and replacement plans or NIS-2 forms in accordance with the ASME Code Section XI January 24, 2001, the Limiting Condition for Operation (LCO) described in Technical Specification Section 3.15.A for the snubbers not meeting code requirements. Monticello's Technical Specifications include an LCO that indicates failure to comply with the requirements of ASME Code Section XI for quality groups A, B, and C (Class 1, 2, and renders the components inoperable. Since the snubbers were inoperable, Technical Specification 3.6.H which allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to complete an engineering evaluation to demonstrate snubber acceptability. On January it was determined that the LCO should have been entered on January 19, 2001 when the problem was first identified.
a result, supported systems were promptly declared inoperable as required by Technical Specification 3.6.H and a was initiated to conform to the limiting Technical Specification requirement (i.e., be in a hot shutdown condition 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). Since the engineering evaluation which demonstrated snubber acceptability was completed prior to reaching shutdown, the shutdown was halted.
January 29, 2001, SRV topwork replacements were found to have been performed without complying with code Therefore the SRVs were declared inoperable and Technical Specification LCO 3.6.E.2 was entered. A was initiated. A Notice of Enforcement Discretion was requested and verbal approval granted on January 30, 2001.
shutdown was halted after the plant was granted the request for enforcement discretion.
components have been found that do not meet code requirements for similar reasons. These components have been into the corrective action process and evaluated for operability in accordance with Generic Letter (GL) 91-18 Notice of Enforcement Discretion. This LER supplement addresses items found after the original reported items. These components are shown in the table below.
work plant 1986 not for 1986 was 3) was 25, plant plant and Discovery Date ASME Section XI Component Non-Conformance Summary 1/31/01 Control Rod Drive System' accumulators5, piping' and vent and drain valves' ANII involvement was not obtained for repair and replacements 1/31/01 One snubber on Combustible Gas Control System', three snubbers on Residual Heat Removal System' and two snubbers on Core Spray System' Six safety related snubbers were found not to be included in Section XI ISI Program 1/31/01 Reactor � Pressure � Vessel � Recirculation System12 inlet nozzles' NIS-2 Forms were not completed for repair 1/31/01 Standby Liquid � Control � System" � squib valves, relief valves and accumulator NIS-2 Forms were not completed for repair and replacements 1 EIIS Code = SNB � 4 EIIS Code = AA � 7 EIIS Code = VTV � 10 EIIS Code = BM � 13 EIIS Code = NZL 2 EIIS Code = BJ � 5 EIIS Code = ACC � 8 EIIS Code = BB � 11 EMS Code = BR � 14 EIIS Code = V 3 EIIS Code = RV � 6 EIIS Code = PSP � 9 EIIS Code = BO � 12 EIIS Code = AD Discovery Date ASME Section XI Component Non-Conformance Summary 1/31/01 Various components repaired or replaced during 1993 refueling outage (See Note 1) NIS-2 Forms were not completed 2/1/01 Emergency Service Water System' piping, pipe supports2, pump3, valves, cooling units coils' and condenser heads' NIS-2 Forms were not completed for repair and replacements 2/2/01 Core Spray and Standby Liquid Control System relief valves NIS-2 Forms were not completed for bolting replacement 2/6/01 RHR Service Water System air vents, control valves', check valves, 1 pumps, biocide 1 injection 1 valves' and 1 corrosion coupon holders NIS-2 Forms were not completed for repair and replacements 2/6/01 Standby Liquid Control System relief valves Section 1 XI 1 coordinator 1 signature 1 was 1 not obtained for modification to install the valves 2/6/01 Emergency Service Water System biocide injection 1 valves 1 and 1 corrosion 1 coupon holders NIS-2 Forms were not completed for repair and replacements 2/7/01 Low Pressure Coolant Injection System valves and five associated supports ANII involvement was not obtained for repair and replacements 2/7/01 Emergency Service Water and Residual Heat Removal Service Water System piping and supports The NIS-2 Form did 1 not have the correct modification number on the form 2/8/01 Main Steam System" snubber NIS-2 Form was not completed for repair to the snubber clamp 2/12/01 Residual 1 Heat 1 Removal 1 System 1 heat exchangers'2 and valves.
Repair/Replacements plans and NIS-2 Forms were not prepared for repair and replacements 2/12/01 Combustible Gas Control System solenoid valves' Repair/Replacements plans and NIS-2 Forms were not prepared for replacements 2/13/01 Emergency Service Water System pumps, valves and piping The scope of work was not complete on NIS-2 Forms. ANII involvement could not be verified.
Note 1: Components involved include Residual Heat Removal System valves, pipe welds and pipe supports, High Pressure Coolant Injection valves, pipe supports and rupture discs', Reactor Coolant Injection System9 valves and rupture discs, Standby Liquid Control System pump, Reactor Pressure Vessel head studs, Reactor Recirculation System piping thermowelll° and the Primary Containment Hard Pipe Vent System.
1 EIIS Code = BI 1 4 EIIS Code = CCL 1 7 EIIS Code = INV 1 10 EIIS Code = TW 1 13 EIIS Code = FSV 2 EIIS Code = SPT 1 5 EIIS Code = CDU 1 8 EIIS Code = RPD 1 11 EIIS Code = SB 3 EIIS Code = P 1 6 EIIS Code = FCV 9 EllS Code = BN 1 12 EIIS Code = HX
Event Analysis
Analysis of Reportability This event is being reported as required by 10CFR50.73(a)(2)(i)(B) as a condition prohibited by Technical Specification 3.15.A and 3.6.E.2 for SRV topworks and 3.6.H for snubbers. These components were replaced, in violation of the ASME Code Section XI, without preparation of repair and replacement plans, NIS-2 forms and ANII involvement.
This event is being reported as required by 10 CFR50.73(a)(2)(vii) since the snubbers were declared inoperable, as required by Technical Specifications 3.15.A, and affected redundant systems were also considered inoperable. Subsequent analysis demonstrated the snubbers to be acceptable and the affected systems to be operable.
Additionally, this event is being reported as required by 10CFR50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications since the LCO for Section 3.6.H (snubbers) was entered on January 24, 2001 instead of on January 19, 2001 when the problem was first identified. Therefore, that LCO was not satisfied.
Other components have been identified that have similar non-conformances with the ASME Code Section Xl. This supplement is being reported as required by 10CFR50.73(a)(2)(i)(B) as a condition prohibited by Technical Specification 3.15.A. These components were replaced or repaired, in violation of the ASME Code Section XI, without preparation of repair and replacement plans, NIS-2 forms and ANII involvement.
Safety Significance
The ANII involvement provides and documents third party review of technical and quality requirements of the code. The Monticello quality assurance, quality control and work control processes compensate for (but do not substitute for) the lack of ANII involvement in the repairs or replacements. Therefore, we believe that there is very low safety significance to this event.
The recently identified ASME Code Section XI non-conformances have been evaluated in accordance with NRC GL 91-18 to ensure that component operability is not adversely affected. Evaluation of operability per NRC GL 91-18 is consistent with practices for other instances of degraded or non-conforming conditions and is incorporated as an integral part of the Monticello Corrective Action Program.
A bounding quantitative probabilistic risk assessment (PRA) has been performed as a sensitivity study to show that the potential increase in risk associated with failure to involve the ANII as required by the ASME Code is small. The PRA analysts believe that the additional likelihood for failure of the SRVs to perform their functions is less than 1% over their currently assumed failure rate. A result of 1.46 E-05/yr core damage frequency (CDF) is obtained by assuming a 10% increase in the failure rates. This can be compared to a baseline CDF of 1.44 E-05/yr. This amounts to less than a 1.5% increase in CDF due to the exaggerated degradation in reliability of SRVs to perform their function. In conclusion, there is less than minimal increase in risk due to lack of the ANII involvement since the SRVs are able to perform their intended function. The sensitivity study shows that any potential increase in risk is very small.
Similar assessments for other non-conformances have been found to show similar results.
Cause
Interpreting ASME Code Section XI requirements often requires expert knowledge. Although the plant staff responsible for work planning has a good general knowledge of ASME Code Section XI requirements for repair and replacement activities for reactor pressure boundary components (e.g. pumps and valves), they lacked expertise to assure compliance for replacement of component supports and other components.
The work control documents did not provide sufficient direction for completing repair and replacement plans, NIS-2 forms, and obtaining ANII involvement.
The Technical Specifications were not literally followed when, on January 19, 2001, it was first realized, that ASME Code Section XI requirements for ISI were not fully met. Technical Specification LCO 3.15.A and 3.6.H should have been entered at that time.
An investigation team is currently in the process of determining the root cause of this event.
Corrective Actions
Plans have been developed to determine the full extent of condition for ASME Code Section XI non-compliance and to restore compliance. Independent ASME Code experts have completed a Self-Assessment. A root cause investigation team has been formed to identify the root cause. Findings and actions are being entered into the Corrective Action Program for disposition.
Management expectations have been reinforced regarding literal compliance with Technical Specifications.
Evaluations have been conducted which demonstrated that the affected components are operable.
Process changes, training and procedural improvements are being formulated.
A License Amendment Request, "Relocation of ASME Inservice Inspection Requirements to a Licensee Program" has been approved on March 1, 2001.
Failed Component Identification none Similar Events Interval Due to Inaccurate Drawings and Failure to Report This Event in a Timely Manner Due to Personnel Error" The corrective actions for these LERs did not prevent this event because they did not focus on the lack of organizational knowledge regarding the AMSE Code Section XI non-conformances and the extent of condition. These two similar events were reported as a condition prohibited by Technical Specification 3.15.A. However, the affected systems were not declared inoperable as required by Technical Specification LCO 3.15.A. Since these events were discovered after January 23, 2001 and occurred over 3 years ago, additional reporting is not required.