05000331/LER-2006-004, Regarding Containment Atmosphere Dilution System Condition Prohibited by Technical Specifications
| ML063120176 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 11/06/2006 |
| From: | Vanmiddlesworth G Duane Arnold |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NG-06-0770 LER 06-004-00 | |
| Download: ML063120176 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3312006004R00 - NRC Website | |
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Duane Arnold Energy Center FPL Energy Duane Arnold, LLC 3277 DAEC Road Palo, Iowa 52324 November 6, 2006 NG-06-0770 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Licensee Event Report #2006-004-00 Please find attached the subject Licensee Event Report (LER) submitted in accordance with 10 CFR 50.73. This letter contains no new NRC commitments.
'G'ýr Van Middlesworth Site Vice President, Duane Arnold Energy Center FPL Energy Duane Arnold, LLC cc:
Administrator, Region IIl, USNRC Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC Z,
BLIND CARBON COPY LIST FOR NG-06-0770 November 6, 2006 J. Bjorseth CNRB D. Lowens D. Curtland G. Hawkins S. Hailer R. Anderson S. Catron G. Rushworth PORC (L. Johnson)
INPO Central Iowa Power Cooperative Corn Belt Power Cooperative DAEC-CTS Project Licensing-LER Binder IRMS
SUBJECT:
Licensee Event Report No. 2006-004-00 File:
A-120
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/3012007 (6-2004)
, the NRC (See reverse for required number of may not conduct or sponsor, and a person is not required to respond to, the diqits/characters for each block) information collection.
- 3. PAGE Duane Arnold Energy Center o5000 331 1
OF 7
- 4. TITLE Containment Atmosphere Dilution System Condition Prohibited by Technical Specifications
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV M
FACILITY NAME DOCKET NUMBER NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 09 07 2006 2006 4
0 11 06 2006 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all. that apply) 1 E] 20.2201(b)
E] 20.2203(a)(3)(i)
[
50.73(a)(2)(i)(C)
[
50.73(a)(2)(vii)
[]
20.2201 (d)
[]20.2203(a)(3)(ii)
[]50.73(a)(2)(ii)(A)
[]50.73(a)(2)(viii)(A)
Ej 20.2203(a)(1)
[
20.2203(a)(4)
[
50.73(a)(2)(ii)(B)
[
50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
[
50.36(c)(1)(i)(A)
Li 50.73(a)(2)(iii)
[
50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[
20.2203(a)(2)(ii)
[
50.36(c)(1)(ii)(A)
[
50.73(a)(2)(iv)(A)
[
50.73(a)(2)(x) 100%
[
20.2203(a)(2)(iii)
[
50.36(c)(2)
E] 50.73(a)(2)(v)(A)
E] 73.71 (a)(4) 10%
]
20.2203(a)(2)(iv)
[]50.46(a)(3)(ii)
[j 50.73(a)(2)(v)(g)
E] 73.71 (a)(5)
[l 20.2203(a)(2)(v)
[] 50.73(a)(2)(i)(A)
[] 50.73(a)(2)(v)(C)
[
OTHER L] 20.2203(a)(2)(vi)
[
50.73(a)(2)(i)(B)
F1 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Robert J. Murrell Regulatory Affairs Engineering Analyst (319) 851-7900CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 16. EXPECTED MONTH DAY YEAR SUBMISSION C) YES (If yes, Complete 15. F_.XPECTED SUBMISSION DATE)
[]
NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On August 19, 2006, while operating at 100% power, Containment Atmosphere Dilution (CAD) North Torus Inboard Isolation valve (SV-4334A) was declared inoperable as the result of not being able to fully open during testing. As a result, valve V-43-112, 'A' CAD Torus Spray Isolation, was closed and tagged in order to comply with TS 3.6.1.3, Required Action A.1 (isolate affected penetration in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). On August 31, 2006, V-43-112 was inadvertently opened after performance of a post maintenance Local Leak Rate Test. This condition existed until September 7, 2006. During this time, the unisolated inoperable containment penetration resulted in the plant being in a Condition Prohibited by Technical Specifications. This event was reportable to the NRC under 10CFR 50.73(a)(2)(i)(B).
The cause of this event was inadequate execution of existing DAEC processes and expectations, as well as weaknesses in existing processes related to the tagging of primary containment isolation valves, which resulted in inappropriate actions being taken by Operations personnel associated with this event.
There were no actual safety consequences and no effect on public health and safety as a result of this event.
NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2006
-" 004 00 2of7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
I. Description of Event
On August 19, 2006, Containment Atmosphere Dilution North Torus Inboard Isolation valve (SV-4334A) was declared inoperable due to its failure to fully open during performance of Surveillance Testing Procedure (STP) 3.6.1.3-02. To comply with Technical Specification (TS) 3.6.1.3, Required Action A.1 (isolate affected penetration in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />), valve V-43-112, 'A' CAD Torus Spray Isolation, was closed. V-43-112 was closed on August 19, 2006 and controlled via a configuration control clearance (7303-SV4334A-RO) by use of a caution tag in the closed position.
On August 25, 2006, an as-found local leak-rate test (LLRT) was completed on SV-4334A per STP 3.6.1.1-12. This test requires V-43-112 to be closed during the test but restored to Locked Open (LOP) following the test per the Valve Lineup List. The Post-Test Position was modified by the Work Control Center Supervisor (WCCS) to require V-43-112 to be closed (CL). This change was annotated with a note stating "V-43-112 to remain closed to maintain compliance with TS required action to keep flow path isolated while SV4334A INOP. Configuration will be maintained via clearance 7303-SV4334A-RO." The valve was signed as closed and independently verified as closed.
On August 27, 2006, a clearance to support valve repairs was hung. The configuration control clearance was cleared as the maintenance clearance was hung. At this point, there was only a maintenance clearance on the CAD system controlling the primary containment isolation boundary.
The closed caution tag was lifted from V-43-112 and a closed danger tag was placed on the valve.
The valve was not repositioned. The maintenance personnel signed off their clearance on August 28, 2006.
On August 29, 2006, the maintenance clearance was removed while a new configuration control clearance was hung. The position required for configuration control clearance 7303-SV4334A-R2 was "Information" and did not directly specify a required position. The tag included the following: "Maintain valve CLOSED for PCIS isolation unless directed to operate valve per CRS/OSM under Administrative Controls. SV4334A INOP."
During the morning of August 31, 2006, a post-maintenance as-left LLRT was conducted per STP 3.6.1.1-12. The WCCS assigned an operator to perform the Post-Test Position lineup. The operator assigned to perform the Post-Test lineup opened V-43-112 and locked it open per the lineup instruction sheet. This lineup sheet had not been annotated regarding the position of V-43-112 as was previously done on the earlier LLRT performed on August 25, 2006. The operator opened V 112 believing to be in compliance with the information tag on the valve since he had been directed by the WCCS to perform the Post-Test Position lineup. The operator did not question the apparent conflict between the information tag wording and the action he was assigned to perform.
During the afternoon of August 31, 2006, another operator was directed to perform an Independent Verification on the lineup. This operator verified V-43-112 was locked open per the lineup sheet and observed the information tag on the valve. He noted the information tag number on the lineup sheet and brought the existence of the tag to the attention of the WCCS. The operator observed the WCCS annotate some information on the lineup sheet and considered the issue resolved. The WCCSU.S. NUCLEAR REGULATORY COMMISSION (1-200 1)
LICENSEE EVENT REPORT (LER)
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YEAR SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 1NUMBER NUMBER 2006
-- 004 00 3of7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) annotated on the lineup sheet "This valve is tagged closed until SV4334A is declared operable. This is configuration control tag 7303-SV4334A-R2. Configuration will be maintained via this clearance."
He also lined out the required position for V-43-112 from LOP to CL. The WCCS had the understanding that V-43-112 was still in the closed position. He failed to realize that the operators had just repositioned and verified V-43-112 to the Locked Open (LOP) position.
I On September 7, 2006, operators were performing the final post-maintenance testing on SV4334A when they identified that V-43-112 was Locked Open. CAD work was stopped and V-43-112 was closed to comply with Tech Specs. CAP 44066 was generated and appropriate notifications were made.
Since the line integrity was maintained (SV-4334A was not disassembled during the period that V43-112 was open), there were no actual safety consequences and no effect on public health and safety as a result of this event.
II. Assessment of Safety Consequences
Valve V43-0112 is located on a 2 inch line connecting CAD system nitrogen cylinders to the torus spray ring header. The isolation valves for this line are SV-4334A (inboard) and SV-4334B (outboard). The inboard isolation valve, SV-4334A, was declared inoperable on August 19, 2006. To comply with TS LCO 3.6.1.3, V43-0112 was closed on that date. However, it was found to be in the open position on September 7. Investigation revealed that V43-0112 had been in this state from the morning of August 31 until September 7, a time span of approximately 7 days.
The nuclear safety significance of manual isolation valve V43-0112 being left in the open position for seven days while inboard isolation valve SV-4334A was inoperable is very low. The line on which these valves are located was still capable of being isolated by outboard isolation valve SV-4334B unless SV-4334A was disassembled (SV-4334A was not disassembled during the period that V43-1.12 was open).
In addition, the line is small and therefore, as described below, the radiological consequences of potential releases through the line are small.
The probability and potential consequences of the primary containment not being isolated following an accident is assessed in the Level II PRA model. Treatment of isolation failures is described in the PRA documentation [Ref: Level 2 PSA Analysis dated April 1995, Appendix C.2] for the containment isolation event tree node (IS). Isolation failures are defined in this analysis to include failures in the containment boundary that allow leakage into the reactor building or directly to the environment. The two principal categories of leakage sources are 1) isolation valve failures and 2) pre-existing failures related to hatches, electrical penetration assemblies, or other non-valve penetrations. The success branch of the IS node is satisfied if the containment penetrations that communicate between the drywell or wetwell atmosphere and the reactor building (or environment) are "closed and isolated." The functional criteria used to satisfy the requirement of "closed and isolated" is that no line, hatch, or penetration has an opening greater than 2 inches equivalent diameter. As such, all isolation valves on lines 2 inches in diameter or less are screened from consideration as potential sources of containment failure. Inability to isolate any single pipe of this size does not influence the course of an accident or radionuclide release to a degree that requires its explicit consideration. The basis for considering small diameter piping to be of low safety significance with regard to containment failure is as follows:U.S. NUCLEAR REGULATORY COMMISSION (1-200 1)
LICENSEE EVENT REPORT (LER)
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YEAR SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2006 004 00 4of 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
They are not large enough to reduce containment pressure fast enough to prevent further containment over pressurization. That is, the containment is still subject to rupture, with resultant high radiological consequences, if containment pressure is not controlled via venting. As such, risk is dominated by failure to control containment pressure rather than by release from the small opening.
The radionuclide release rate is low and therefore of minor consequence compared to that of large containment ruptures or unisolated pipe.
Since the concern of interest (inadequate control of valve configuration) involved a line of only 2 inches in diameter, its safety significance is very low based solely on the relatively mild consequences of such a line failing to successfully isolate. However, due to the presence of closed isolation valve SV-4334B and the fact that SV-4334A was not disassembled while V43-112 was open, the likelihood of such an occurrence is very small. Although SV-4334A and SV-4334B are not included in the Level II PRA model, valves that isolate larger lines connected to the containment are included. The impact on plant risk of having only SV-4334B available to isolate the subject 2 inch CAD line can be assessed in a bounding manner by assuming one of two isolation valves in a larger line connected to the containment fails to close, or fails to remain closed, when demanded in response to an accident. Such a valve is CV-4313.
CV-4313 is the inlet isolation valve on a 6 inch Nitrogen Makeup line leading to the torus. Increase in Large/Early Release Probability (LERP) can be calculated for the condition where both the inboard (CV-4313) and outboard (CV-431 1) valves are initially placed in the open position, and CV-4313 fails to close with complete certainty when demanded. The adverse condition of both valves left in the open position is assumed to exist for a seven day period.
The base Large/Early Release Frequency (LERF) for the Rev. 5C Level II PRA model is 1.393E-06/yr.
(This value is derived using a truncation limit of 1 E-1 2 /yr.) When the basic event representing probability of both isolation valves initially being left in the open position (ISPWWEXH90L) and the basic event representing probability of CV-4313 failing to close in response to an accident (ISAV4313-24N) are both set to 1.0, LERF increases to 1.405E-06/yr. Therefore, over a 7 day period, increase in LERP is:
1.405E -06 C.9E0)x7asx l yr
=2.30E-10 C y -
yr33EO6(Cda1I 365 days)
Risk associated with any temporary configuration is regarded to be insignificant if the increase in Core Damage Probability (CDP) associated with the configuration is less than 1.OE-06 and the increase in LERP is less than 1.OE-07.
The condition of an open isolation valve on a process line penetrating the torus shell does not impact the probability of core damage. Therefore, increase in CDP is less than 1.OE-06. Increase in LERP for the given condition over a seven day period is estimated to be 2.30E-1 0. This value is less than 1.OE-07.
Therefore, the nuclear safety significance associated with manual valve V43-0112 left in the open position for approximately seven days while inboard isolation valve SV-4334A was inoperable is very low.U.S. NUCLEAR REGULATORY COMMISSION (1-200 1)
LICENSEE EVENT REPORT (LER)
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Duane Arnold Energy Center 05000331 YERINEUETAuMBER REIN~UMBER_
2006 004 00 5 of 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Ill. Cause of Event:
As a result of this event a Root Cause Evaluation was conducted (RCE 1055). As a result of data analysis, interviews, process reviews, and an extent of condition review, the RCE Team identified the following Inappropriate Actions associated with the data analysis conducted:
- 1. Inadequate controls established to maintain isolation of an inoperable primary containment penetration.
- 2. Inadequate post LLRT valve lineup resulted in failure to ensure that an inoperable primary containment penetration was isolated.
- 3.
In conflict with site expectations, operator was given the post LLRT valve lineup without being informed of tagged valve.
- 4. The operator repositioned V43-112.
- 5. The WCCS failed to recognize that the valve was not in the position required by Technical Specifications.
Once the inappropriate actions were arrived at, the RCE Team analyzed the actions using the Why Staircase methodology to derive the root causes and contributing factors associated with this event. As a result, the following root causes were found:
0 Inadequate procedure guidance for maintaining inoperable primary containment penetrations isolated.
LLRT valve lineups failed to meet site standards for repositioning valves and Independent Verification of locked valves in accordance with Administrative Control Procedures (ACPs) 1410.9, "Locked Valve Program," 101.2, "Verification Process and Self/Peer Checking Practices",
and Integrated Plant Operating Instruction (IPOI) 7, "Special Operations."
In addition to the root causes identified, the following contributing factors were also found:
In conflict with site expectations, operator was given the post LLRT valve lineup without being informed of a tagged valve (No Pre Job Brief). This action is in conflict with expectations of senior management as indicated in interviews.
Operator took actions to open V43-112 without fully considering the consequences and without verifying what were the Administrative Controls required by ACP 1410.7, "Guidelines for Primary Containment Valves and Penetrations."
" WCCS took action to change the listed post-LLRT valve lineup position without questioning the actual position of the valve. This was undetected due to communication failures between the WCCS and the Verifier.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
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YEAR I SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2006 004 00 6of7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
IV. Corrective Actions
On September 7, 2006, TS compliance was restored when V-43-112 was closed.
Completed RCE 1055. As a result of this RCE, the following corrective actions were identified:
Interim Corrective Actions Prepare and conduct a case study of RCE 1055 with all of Operations and the Leadership Team.
Responsible Group: Operations Completion Due Date: 11/10/06 CA44201 Corrective Actions to Prevent Recurrence (CATPRs)
RCI - CATPR - ACP 1410.5, "Clearance Program" will be revised to require any inoperable primary containment penetration be isolated using a Caution Card with a required position of Closed.
Additionally, ACP 1410.7, "Guidelines for Primary Containment Valves and Penetrations," will be revised to properly refer to ACP 1410.5.
Responsible Group: Operations Completion Due Date: 12/10/06 CA44202 RC2 - CATPR - Revise LLRT valve lineups to require independent verifications of repositioned locked valves in accordance with existing site standards. This revision will include changing the format of LLRT valve lineups to allow for documentation of identified discrepancies (align with IPOI 7 guidance.)
Responsible Group: Operations Completion Due Date: 12/10/06 CA44203 Corrective Actions for Contributing Factors CFH - Communicate expectations for the conduct of prejob briefs related to valve lineups and other activities that might be considered "routine."
Responsible Group: Operations Completion Due Date: 12/10/06 CA44204U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
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SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 YEAR NUMBER NUMBER 2006
-- 004 00 7of7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
CF2 - Conduct training for all of Operations on the application of "Administrative Controls" per Technical Specifications.
Responsible Group: Operations Completion Due Date: 12/10/06 CA44205 CF3 - CATPR2 for RC2 will also address this item.
CF4 - Develop and conduct a Communications exercise for all of Operations that explores the pitfalls of imprecise communication and reinforces the techniques for avoiding communication problems.
Responsible Group: Operations Completion Due Date: 12/10/06 CA44206 Other Corrective Actions In regards to human performance issues associated with Operator Fundamentals, INPO conducted an assist visit at DAEC between July 31st and August 4th, 2006 to identify performance improvement opportunities in the areas of Operator Fundamentals. This assist visit is captured as a self-assessment (SA01 0827) and generated several action items to improve performance in the area of operator fundamentals. No other specific corrective actions are needed at this time. Operations will continue to monitor human performance through Department Roll Up Meetings (DRUMS), Observations, and Crew Management Review Meetings (MRMs).
V. Additional Information
Previous Similar Occurrences:
A review of LERs at the DAEC over the last 3 years identified no LERs with similar events.
EIIS System and Component Codes:
JM Containment Isolation Control System
Reporting Requirements
This report is being submitted under 10 CFR 50.73(a)(2)(i)(B).