05000461/LER-2010-001, For Clinton, Unit 1, Regarding Unanalyzed Leakage Pathway Affection Residual Heat Removal a Pump Room Flooding Analysis
| ML100900264 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/25/2010 |
| From: | Kearney F Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| U-603949 LER 10-001-00 | |
| Download: ML100900264 (5) | |
| 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)(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(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4612010001R00 - NRC Website | |
text
Exelon.
Nuclear Clinton Power Station 8401 Power Road Clinton, IL 61727 U-603949 March 25, 2010 SRRS 5A.108 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461
Subject:
Licensee Event Report 2010-001-00 Enclosed is Licensee Event Report (LER) No. 2010-001-00: Unanalyzed Leakage Pathway Affecting Residual Heat Removal A Pump Room Flooding Analysis. This report is being submitted as a voluntary LER as provided in NUREG 1022.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this report, please contact D. J. Kemper, at (217)-937-2800.
Respecffully, A.Ke iey Site Vice Presiden Clinton Power Station JLP/blf
Enclosures:
Licensee Event Report 2010-001-00 cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - Clinton Power Station Office of Nuclear Facility Safety-IEMA Division of Nuclear Safety
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 information collection.
- 3. PAGE Clinton Power Station, Unit 1 05000461 1 OF 4
- 4. TITLE Unanalyzed Leakage Pathway Affecting Residual Heat Removal A Pump Room Flooding Analysis
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEUNILRVFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REVNO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 10 07 2009 2010 -
001 00 03 25 2010 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)
[I 50.73(a)(2)(i)(C)
[E 50.73(a)(2)(vii)
[1 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
[I 50.73(a)(2)(viii)(B)
_ 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71(a)(4) 97 El 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C) x OTHER El 20.2203(a)(2)(vi)
[E 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in The elevation of the RHR A Pump Room floor is 707"-6" and the room is designed to be watertight up to elevation 731'-5". The elevation of the line that connects to the RHR A pump room is 720'-6", or 13'-0" off the floor. Based on the leakage rate into the room, it will take approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to reach the elevation of 720'-6". The room will continue to fill.
The suppression pool design parameters, such as level and volumes, are shown in USAR Table 6.2-1. The suppression pool water volume at low suppression pool water level (730"-11") is 135,220 ft3 (containment) plus 10,707 cuft3 (drywell). The minimum suppression pool water level is 727'-1" (minimum vent coverage) per USAR 6.2.4.3.3 and mechanical design drawing M05-1069. This volume of water is approximately 224,000 gallons.
Upper pool dump, which adds 14,748 ft3 (110,000 gallons), provides an additional 2'-0" (approx.) to the suppression pool, if needed. If suppression pool level were to approach the minimum suppression pool water level, Emergency Operating Procedures would direct the Control Room operators to dump the upper pool prior to reaching 727'-1". At a rate of 206 gpm, without upper pool dump, the suppression pool could lower to the minimum vent coverage level in approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. If the upper pool dump volume is considered, this would add 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, for a total of 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />, to reach the minimum design level irrespective of any flooding issues in the RHR A pump room. Additionally, operators can add water to the suppression pool from the Cycled Condensate Storage Tank (CST) [KA] using CPS Procedure 3208.01, "Cycled/Makeup Condensate". This procedure provides a means of gravity draining of the CST to the suppression pool. The CST has a usable volume of 312,000 gallons and provides additional several hours of water inventory.
Assuming that a loss of offsite power occurs at the time of the piping failure in the RHR A pump room, a reactor scram will occur accompanied with an isolation of the main steam isolation valves. The Reactor Core Isolation Cooling (RCIC) [BN] system will initiate on low water level and will inject automatically. RCIC system operation will continue to occur until reactor pressure is reduced to 150 psig. No credit is taken for water additions to the suppression pool from the RCIC tank due to RCIC operations. However, the volume of the RCIC tank contains several hours of RCIC operation (about 125,000 gallons) prior to switching to the suppression pool.
High Pressure Core Spray [BG], Low Pressure Core Spray, and RHR C system are available to continue the cooldown to cold shutdown and RHR B can be operated in the Shutdown Cooling mode to maintain the unit in cold shutdown. Cold shutdown can be achieved in about 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> with only one heat exchanger in operation (reference USAR Fig. 5.4-12). This is well within the time before the suppression pool would lower to the minimum suppression pool water level of concern.
CAUSE OF EVENT
The cause of this event was determined to be a historical design oversight during plant construction that allowed the RHR A pump room floor drains to be connected to the radwaste pipe tunnel.
SAFETY ANALYSIS
This event did not result in any safety system functional failure.
A flooding event due to a moderate energy break concurrent with a loss of offsite power is not likely to occur.
This event is considered to have no safety significance since there was no actual loss of a safety function.
Based on the above, significant margin exists to achieve cold shutdown well within the time required before minimum vent coverage would be reached, or well before any loss of safety function.
In summary, based on an analysis of a postulated piping break coincident with a single active failure of the pump suction valve does not result in loss of safety function, degradation of plant safety barriers, or unanalyzed condition.
CORRECTIVE ACTIONS
The original design issue was corrected with the installation of a welded plate in the floor drain line. An Engineering Change 377321 was initiated to capture the welded plate into plant design documents.
PREVIOUS OCCURRENCES
None
COMPONENT FAILURE DATA
NonePRINTED ON RECYCLED PAPER NRIC FORM 366A (9-2007)
PRINTED ON RECYCLED PAPER