05000289/LER-2009-001, Regarding Multiple Main Steam Safety Valve Test Failures
| ML093630040 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/22/2009 |
| From: | Dougherty T Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TMI-09-153 LER 09-001-00 | |
| Download: ML093630040 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(b)(2)(ii) |
| 2892009001R00 - NRC Website | |
text
Exelkn Three Mile Island Unit 1 Route 441 South, P.O. Box 480 Middletown, PA 17057 Telephone 717-948-8ooo Nuclear December 22, 2009 TMI-09-153 10 CFR 50.73 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI-1)
OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 SUBJECT: LICENSEE EVENT REPORT (LER) NO. 2009-001-00
" Multiple Main Steam Safety Valve Test Failures This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(i)(B). For additional information regarding this LER contact Adam Miller of TMI Unit 1 Regulatory Assurance at (717) 948-8128.
There are no regulatory commitments contained in this LER.
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Thomasd.
Qtou Plant Manager TJD/awm ATTACHMENT: List of Regulatory Commitments cc:
TMI Senior Resident Inspector Administrator, Region I TMI-1 Senior Project Manager
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the
- 3. PAGE Three Mile Island, Unit 1 05000289 1 OF 5
- 4. TITLE:
Multiple Main Steam Safety Valve Test Failures
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE J
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER SEQUENTIAL REVI 50 MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR N/A 05000 2FACILITY NAME DOCKET NUMBER 10 23 2009 2009 -
001 00 12 22 2009 N/A 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
[I 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[I 50.73(a)(2)(vii)
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- 10. POWER LEVEL [l 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
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Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER F-ACILITY NAME TELEPHONE NUMBER (Include Area Code)
Adam W. Miller of TMI-1 Regulatory Assurance (717) 948-8128MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FACURE R
EPO X
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX B
SB RV D245 Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION [E YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
During testing prior to refueling outage IR18 (October 2009), six TMI-I Main Steam Safety Valves (MS-V-19A, 17C, 18C, 19C, 21B and 19D) failed "as-found" setpoint testing. These valves are Dresser 3700 series steam safety valves.
Valve lift pressures were more than 3% above setpoint during performance of surveillance test 1303-11.3. The valves were declared inoperable, and reactor power was reduced to comply with Technical Specification 3.4.1 until operability was restored. Many industry event reports, broader Exelon experience, TMI experience, and EPRI studies associated with this problem with Dresser 3700 series steam safety valves were reviewed to develop the cause and corrective actions. The primary root cause is a design deficiency, in that the design does not prevent the formation of an oxide bond between the valve disc and seat, which results in unacceptable valve performance. Material changes and process control improvements will be established, which have been shown to be the most effective means for preventing unacceptable Main Steam Safety Valve (MSSV) lift pressures. This problem applies to Dresser series 3700 relief valves. At TMI-I, these valves are only used in Main Steam Safety Valve applications. These valves are required to limit system pressures to within design limits.
The condition of the affected MSSVs during cycle 17 would not have prevented the performance of MSSV safety functions credited in the safety analysis. The submittal of this LER constitutes reporting to the NRC in accordance with 10 CFR 50.73 (a)(2)(i)(B), "a condition prohibited by Technical Specifications."
NRC FORM 366 (9-2007)(9-2007)
LICENSEE EVENT REPORT (LEI U.S. NUCLEAR REGULATORY COMMISSION I
- 1. FACILITY NAME Three Mile Island, Unit 1 A.
EVENT DESCRIPTION
Plant Conditions before the event:
Babcock & Wilcox - Pressurized Water Reactor - 2568 MWth Core Power Date/Time: October 21, 2009/1429 hours Power Level: 85% steady state power Mode: Power Operations There were no structures, systems, or components out of service that contributed to this event.
Event:
Between October 21 and 25 in 2009, lift setpoint verification IN SITU testing was performed by mechanical maintenance in accordance with 1303-11.3 "Surveillance Test and Set Main Steam Safety Valves." Six Main Steam Safety Valves [MS-V-19A, 170, 180, 190, 21B and 19D] *[SB/RV] were identified with lift pressures above the IST acceptance criteria (i.e., more the 3% above setpoint). Each valve is tested to determine lift setpoint at least twice before any adjustments are performed. Each valve was declared inoperable until the setpoint had been adjusted and repeated tests demonstrated lift pressures within +/- 1% of setpoint. An additional valve [MS-V-21A] was quarantined to preserve evidence of the problem, declared inoperable, and not tested at that time.
At 2:29 PM on 10/21/09, MS-V-19D did not lift at 6% above setpoint. TMI-1 Technical Specification (TS) 3.4.1.2.3 requires all eighteen MSSVs to be operable above 5% reactor power. With one MSSV inoperable, Tech Spec 3.4.1.2.3 requires that reactor power be reduced below 92.4%. When the MS-V-19D was declared inoperable, reactor power was - 85%. The reactor was coasting down toward refueling outage 1 R18. On 10/23/09, MS-V-19C failed to lift at 6% above setpoint as well. With two MSSVs inoperable on one Once Through Steam Generator (OTSG), TS 3.4.1.2.3 requires that reactor power be reduced below 79.4%. At 1:52 PM, reactor power had been reduced from approximately 83%
to 74% to ensure compliance with TSs. MS-V-1 9C setpoint was adjusted; the valve was tested satisfactorily and declared operable. At 5:41 PM, when testing of MS-V-1 8C identified that a third valve had failed, and each valve that failed had been refurbished in the previous outage, the additional valves refurbished in the previous outage were declared inoperable. In the previous outage (1 R1 7 November 2007), six MSSVs [MS-V-19A, 21A, 18C, 19C, 21B and 19D] were refurbished. With more than three MSSVs inoperable on one OTSG, TS 3.4.1.2.3 requires the MSSVs be restored to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the reactor be placed in Hot Shutdown within an additional six hours. By 7:12 PM, following setpoint adjustment and satisfactory testing activities, only three valves remained inoperable [MS-V-19A, MS-V-21 A, and MS-V-21 B] and reactor power had been lowered to 65% (below the TS maximum power level for the three inoperable MSSVs - 79.4% reactor power). By 10/25/09, adjustments and testing were completed, and MS-V-1 9A and MS-V-21 B were restored to operable status.
The extent of condition is limited to the eighteen MSSVs. Seventeen of eighteen MSSVs were tested IN SITU. MS-V-21A was quarantined and sent to NWS Technologies LLC (NWS) for testing. Eleven valves performed satisfactorily during testing. There are no other Dresser 3700 series steam safety valves used at TMI-I. Thus, it is assumed that seven of the eighteen MSSVs were inoperable at some time during the last operating cycle. This condition did not meet the requirements of TS 3.4.1.2.3, and therefore this event is reported as a 60 day LER in accordance with 10 CFR 50.73 (a)(2)(i)(B).U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
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B.
CAUSE OF EVENT
The primary root cause is a design deficiency, in that the design does not prevent the formation of an oxide bond between the valve disc and seat, which results in unacceptable valve performance.
Additional causes for specific valve problems will be established based on inspection and failure analysis that is being performed at a certified off-site test facility.
C.
ANALYSIS / SAFETY SIGNIFICANCE The safety significance of seven inoperable MSSVs (MS-V-21A, MS-V-21B, MS-V-19A, MS-V-17C, MS-V-18C, MS-V-19C and MS-V-19D) was reviewed. First, the affect of MSSV capacity on each of the Updated Final Safety Analysis Report (UFSAR) Chapter 14 analyses was reviewed. Each of the analyses is bounded by the analysis of three events (1) Loss of electrical load, (2) Anticipated Transient Without Scram (ATWS), and (3) OTSG heat removal via the MSSV following a Reactor and Turbine Trip.
(1) The limiting design event for MSSV capacity is a loss of electrical load. In the analysis of this event the turbine is tripped and an Anticipatory Reactor Trip due to a Turbine Trip (ARTTT) does not occur. The analysis models the real event where the turbine control valves close to prevent turbine overspeed but the condition does not initiate an ARTTT. The MSSV and RCS code safety valve capacity must be sufficient to ensure RCS pressure remains less than 110% of design (2750 psig) and OTSG pressure remains below ASME code maximum of 1169.7 psia. Analysis was performed with six large (six inch valve) MSSVs unavailable. In cycle 17, two of the affected valves were small MSSVs (three inch valve).
The capacity of a small valve is approximately 25% of a large valve. The total capacity of the seven inoperable valves during cycle 17 was less than that of six large valves. Therefore the analysis result bounds the TMI cycle 17 condition.
The analysis showed a peak RCS pressure of 2583 psia and peak OTSG pressure of 1169.1 psia. In addition, the analysis included further conservative assumptions.
Reactor power was 102% of 2568 MWt No reactor power reduction due to core physical response to rising RCS temperature was credited No credit was taken for Integrated Control System (ICS) action to insert control rods No credit was taken for turbine bypass valve or atmospheric dump valve steam relief capacity Additional analysis of the loss of electrical load event has been completed to support a power uprate.
This analysis was the same as discussed above except that reactor power was 102% of 2772 MWt. This analysis for six inoperable MSSVs shows that RCS pressure is only 1 psig higher, 2584 psia. This indicates that primary pressure is controlled by the primary safety valves and is not as sensitive to MSSV capacity.
(2) In the UFSAR Chapter 14 ATWS limiting event (Loss of Main Feedwater), the turbine trip follows the reactor trip. Because of this sequence, the secondary side peak pressure and required MSSV capacityU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
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of are bounded by the turbine trip event.
(3) A conservative computation performed in support of a planned MSSV TS change shows that one large MSSV on each OTSG (two valves in total) is sufficient for RCS heat removal after the reactor is shutdown. This assessment was performed for a 2772 MWt core design and included reactor coolant pump heat.
In summary, the seven inoperable MSSVs would not adversely affect RCS, OTSG or Main Steam system Integrity.
The impact of seven inoperable MSSVs (i.e. lifting above acceptable pressure set point) on the TMI Probabilistic Risk assessment (PRA) was reviewed. The MSSVs provide OTSG heat removal for ATWS and non-ATWS cases. MSSVs are credited when Atmospheric Dump Valves (ADV) and Turbine Bypass Valves (TBV) are not available. For non-ATWS cases, one MSSV on one OTSG is needed to remove adequate decay heat. A common cause failure of all MSSVs must occur to have any impact on risk. The failure of several valves will not have a measurable impact on risk. For ATWS cases, two MSSVs on each OTSG are required to open to remove decay heat. Again, common cause failures are needed to have any impact on risk, although fewer valves need to fail to have a risk impact compared to the non-ATWS case. The extent of cause for the cycle 17 MSSV failures is limited to the MSSVs refurbished in 1 R1 7. However, a majority of the valves would need to fail before there would be any measurable impact on risk. The condition during cycle 17 at TMI would have no noticeable impact on success criteria or timing for the TMI PRA. There is no measurable increase in risk; therefore, the increase in Core Damage Frequency (CDF) is less than 1 E.
D.
CORRECTIVE ACTIONS
MS-V-1 9D and MS-V-21 B were removed and shipped to NWS for inspection, disc analysis by Altran Labs, valve refurbishment by NWS including use of pre-oxidized discs and nozzle passivation, and post refurbishment certified lift pressure testing. (This action has been completed.)
MS-V-1 9A, MS-V-1 7C, MS-V-1 8C and MS-V-1 9C were removed and shipped to NWS for inspection, failure analysis, valve refurbishment including use of pre-oxidized discs and nozzle passivation, and post refurbishment certified lift pressure testing. (This action has been completed, except for failure analysis.)
MS-V-21A was removed and shipped to NWS for as found lift pressure testing, inspection, valve refurbishment including use of pre-oxidized discs and nozzle passivation, and post refurbishment certified lift pressure testing. (This action has been completed.)
Material changes and process control improvements will be established, which have been shown to be the most effective means for preventing unacceptable Main Steam Safety Valve (MSSV) lift pressures. A plant procedure will be revised to ensure passivation of nozzle/seat is required for all MSSV refurbishments.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
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E.
PREVIOUS OCCURENCES
The oxide bonding phenomenon is well known in the industry. Industry efforts to eliminate this problem have been evolving since the mid 1990s. Today, a definitive solution has not been promulgated by the original equipment manufacturer and industry peers have reached various conclusions on how to resolve the issue. Industry efforts include change of material for the discs, use of a pre-oxidation process for the disc surface, modification of the nozzle/seat finish, and treatment of the finished seat. In two recent tests (one in 2005 and one in 2007), TMI-1 had experience with valve lift tests indicative of oxide bonding. In each case, the valve had not been refurbished in the prior outage. This made the cause determination less certain. There was no prior TMI-1 history with high lifts due to oxide bonding. Although TMI-1 didn't have significant oxide bonding related test failures, process improvements (pre-oxidize the disc material) were implemented in 2005. After the 2007 event, TMI-1 conducted further analysis of TMI-1 and industry experience. Additional process improvements were initiated (require nozzle passivation after refurbishment). However, this analysis and these improvements were not implemented before 2007 refurbishments were completed.
Therefore the seats of the valves refurbished in 2007 were not passivated to mitigate oxide bonding.
- Energy Industry Identification System (EIIS), System Identification (SI) and Component Function Identification (CFI) Codes are included in brackets, [SI/CFI] where applicable, as required by 10 CFR 50.73 (b)(2)(ii)(F).