|
|
Line 13: |
Line 13: |
| | document type = Meeting Briefing Package/Handouts | | | document type = Meeting Briefing Package/Handouts |
| | page count = 9 | | | page count = 9 |
| | | project = CAC:MF2400 |
| | }} |
| | |
| | =Text= |
| | {{#Wiki_filter:EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 1 of 9 Quantity of Coatings 1,570 lbs of epoxy coatings on the concrete walls and ceilings |
| | |
| | Unit 1 has a larger quantity since the Unit 2 walls are not epoxy coated all the way up to the ceiling and also since the ceilings in Unit 2 are not epoxy coated. |
| | |
| | The sump performance evaluation is based on the higher quantity for Unit 1. |
| | |
| | The Reactor Cavity is designated as Reactor Containment Building Room 001. |
| | |
| | EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 2 of 9 Type of Coatings Brown and Root Technical Reference Document 5A810WQ005-A "Painting Schedule for Service Level I, II, and III Areas" from 1979 calls for the Reactor Cavity walls to be coated using the following Level I concrete coating system defined in HL&P Specification #7A810XS002-A-HL: Nutec 10 Primer/Sealer Nutec #11S Concrete Masonry/Surfacer Nutec #11 Concrete Masonry/Surfacer Reactic (Nutec) #1201 Epoxy Topcoat |
| | |
| | Service History Just prior to the initial startup of Unit 1 in 1987 during hot functional testing, repairs were made to the concrete wall coating in the Reactor Cavity. Cracks in the originally installed coating were chased and the adjacent coating was removed. These areas were not top-coated due to the unavailability of that coating at the time. Instead the areas were left with only a base sealer. See Unit 1 photographs. |
| | |
| | No other evidence has been found of any repairs since then in Unit 1. Also Unit 2 had no evidence found of any repairs since the startup of the Unit in 1989. A search of work order history and a search of condition reports did not yield any repair requests or coating conditions in either Unit. The maintenance work planner who has been dealing with coatings since startup was not aware of Similar discussion with Engineering personnel who perform Reactor Vessel inspections in Room 001 also did not yield any information concerning coating problems. As seen in the photographs, the current condition of the coatings in both Units appears good. |
| | EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 3 of 9 Current Condition Based on Photographs of the Area Unit 1 Photos EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 4 of 9 Unit 1 Photos EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 5 of 9 Unit 2 Photos from 2013 EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 6 of 9 EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 7 of 9 |
| | |
| | Manufacturer Qualification Data The test document from Southern Imperial Coatings (filed as STI 525724) states that the Nutec 11/ 11S/ 1201 was qualified to 2E08 rads which is consistent with IEEE STD 323-1974 for normal plus LOCA dose. |
| | However, the epoxy coating for the reactor cavity was deemed by Bechtel during the unqualified coating calculation rack up to be unqualified due to the excessive calculated cumulative radiation dose. |
| | |
| | EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 8 of 9 Cumulative Dose Received The epoxy coating for the reactor cavity was deemed to be unqualified due to excessive radiation. |
| | From design criteria document 4E019NQ1009: |
| | The calculated cumulative design dose for the Reactor Cavity for 40 years is: Neutron 2.5E10 Rads Gamma 3.5E9 Rads The accident dose is given as 1.5E8 Rads. |
| | The calculated cumulative dose was determined considering the reactor operated at 100% power every day for 40 years. |
| | |
| | The actual capacity factor for each Unit to date is: Unit 1 83.7% from initial operation in 1988 to March 2016 (28 years) |
| | Unit 2 83.2% from initial operation in 1989 to March 2016 (27 years) |
| | Thus the actual cumulative dose to date is: = capacity factor x (operating years / design life years) x calculated lifetime dose = 0.8 x (28/40) x calculated lifetime dose |
| | = 0.56 x calculated lifetime dose Coatings Assessment The epoxy coatings in the Reactor Cavity (RCB Room 001) in both Unit 1 and in Unit 2 were applied as Service Level I coatings during the construction phase of the plant. However, these epoxy coatings for the reactor cavity were subsequently deemed to be unqualified due to the excessive calculated cumulative radiation dose. They were not part of the qualified coatings inspection program and there was no continuity of inspections. |
| | The epoxy coatings in the Reactor Cavity in both Unit 1 and in Unit 2 are now deemed as Qualified But Degraded based upon consideration of the following: |
| | * Initial application was safety related (Service Level I) |
| | * No historical evidence of coating problem conditions requiring repairs since installation |
| | * Recent visual observation (photos and personnel observations) shows no signs of degradation |
| | * Estimated cumulative dose to date is over half of the calculated cumulative dose but no current signs of degradation |
| | * Estimated cumulative dose to date exceeds the test qualification value but no current signs of degradation EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 9 of 9 EPRI Report 1014884 "Plant Support Engineering: Degradation Research for Nuclear Service Level I Coatings" concludes that the majority of coating failures and signs of degradation can be attributed to undetected deficiencies that occurred at the time of coating application. These deficiencies are the major cause of coating deterioration during the coating systems' service life. The report also states that high radiation is not believed to be a significant cause of coating degradation in containment. |
| | |
| | Since the epoxy coatings in the Reactor Cavity in both Units have been in service for over half of their design life with no signs of degradation, any degradation in the future is not expected. |
| | Thus these coatings are considered Qualified But Degraded. |
| | |
| | Debris Characteristics The Reactor Cavity epoxy coatings are considered as Qualified But Degraded and are assumed to fail as chips with the following debris characteristics: |
| | |
| | Range (inches) Weight Distribution (%) 1.0 - 2.0 32.0 0.5 - 1.0 9.04 0.25 - 0.50 4.41 0.125 - 0.25 5.02 < 0.125 37.1 as 15.6 mils chips; 12.3 as 6 mils particulate |
| | |
| | Transport Evaluation No transport is expected for epoxy coating debris inside the reactor cavity for breaks that do not occur within the reactor cavity because |
| | * for breaks that occur outside the reactor cavity, there is no flow into the reactor cavity, |
| | * the path out of the reactor cavity is sufficiently tortuous that this area will be essentially stagnant, and |
| | * any negligible flow from the reactor cavity that could occur will be to a region on the opposite side of the steam generator compartment from where the sumps are located. |
| | For breaks inside the Reactor Cavity (i.e. at the reactor Vessel nozzle), most of the epoxy coating debris will fail as chips. A portion of the debris that is fine chips and the portion that is particulate is subject to transport. The transport path out of Room 001 is through two 4 inch floor drains in the Reactor Cavity floor that go to the Normal Sump. This sump has a cover plate with an opening for water to go out onto the floor when the sump is full. Thus there will be some transport from the Reactor Cavity to the containment pool. |
| | |
| }} | | }} |
Revision as of 11:10, 21 March 2018
Letter Sequence Meeting |
---|
CAC:MF2400, Control Room Habitability (Approved, Closed) |
Administration
- Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting
|
MONTHYEARML15274A5992015-10-0101 October 2015 10-01-15 Public Phone Call Project stage: Request ML16011A0612016-02-0202 February 2016 Summary of 10/1/15 Meeting with STP Nuclear Operating Company to Discuss Revised Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 at South Texas Project, Units 1 and 2 (CAC MF Project stage: Meeting ML16028A1522016-02-18018 February 2016 1/14/2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Modifications to the Licensee'S Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 Project stage: Meeting ML16088A2432016-04-0101 April 2016 2/18/2016, Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409) Project stage: Meeting ML16082A5072016-04-11011 April 2016 Request for Additional Information, Phased Response Requested, Exemption and License Amendment Request for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC Nos. MF2400-MF2409) Project stage: RAI ML16092A0442016-04-11011 April 2016 Summary of 3/3/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409) Project stage: Meeting ML16092A0852016-04-13013 April 2016 3/17/2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2401-MF2409) Project stage: Meeting ML16096A0652016-04-13013 April 2016 Summary of February 4 and 16, 2016, Regulatory Audit at Westinghouse in Rockville, MD, Boric Acid Precipitation, Exemption and License Amendment Request, Risk-Informed Approach to Resolve Generic Safety Issue 191 Project stage: Other ML16095A0102016-04-13013 April 2016 Summary of November 17-19, 2015 Thermal-Hydraulic Review at Texas A&M University; Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC MF2400-MF2409) Project stage: Other ML16111A0272016-04-18018 April 2016 STP Nuclear Operating Company Slideshow for Public Meeting on April 21, 2016 for GSI-191 Resolution (CAC No. MF2400-01) Project stage: Meeting ML16103A3442016-04-26026 April 2016 Summary of February 24-26, 2016 Audit, Debris Transport Review at Alion Science and Technology Corporation, Albuquerque, Nm; Pilot Submittal and Exemption Request, Risk-Informed Approach to Resolve GSI 191 Project stage: Other ML16032A3872016-04-26026 April 2016 Draft Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400-MF2409) Project stage: Draft Other ML16032A4032016-04-26026 April 2016 FRN, Draft Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400-MF2409) Project stage: Draft Other ML16127A4002016-05-11011 May 2016 Audit Summary, Thermal-Hydraulic Review on February 23-25, 2016 at Texas A&M University; Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue 191 (CAC MF2400-MF2409) Project stage: Other ML16125A2902016-05-26026 May 2016 Request for Additional Information, Risk Review, Exemption and License Amendment Request for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC Nos. MF2400-MF2409) Project stage: RAI ML16141B0812016-05-31031 May 2016 Audit Summary, Risk Audit on April 12-13, 2016, at Alumni Center at the University of Texas, Austin, Tx; Pilot Generic Safety Issue 191 Submittal and Exemption Request, and Draft Request for Additional Information Project stage: Draft RAI ML16175A1082016-06-24024 June 2016 Summary of 4/21/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 Project stage: Meeting ML16190A0082016-07-13013 July 2016 Summary of 6/22/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409) Project stage: Meeting ML16194A2342016-07-21021 July 2016 Request for Withholding Information from Public Disclosure - 5/6/16 Affidavit Executed by D. Munoz and M. Rozboril, Alion Science and Technology, Response to Follow-up RAIs 18, 38, and 44 Project stage: RAI ML16242A0102016-08-23023 August 2016 NRR E-mail Capture - (External_Sender) FW: South Texas Project GSI-191 Draft Risk Responses to Questions for Public Meeting on August 29, 2016 Project stage: Request ML16242A0092016-08-25025 August 2016 NRR E-mail Capture - (External_Sender) South Texas Project GSI-191 Draft thermal-hydraulic Responses to Questions for Public Meeting on August 29, 2016 Project stage: Request ML16238A5272016-09-0101 September 2016 Summary of 7/28/16, Public Meeting with STP Nuclear Operating Company to Discuss the License Amendment and Exemption Requests to Use a Risk-Informed Approach to the Resolution of GSI-191 (CAC Nos. MF2400 - MF2409) Project stage: Meeting ML16258A3672016-09-0606 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution SNPB-3-2, Bullet 4 16 -inch Bounding Break Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A1002016-09-12012 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI APLA-3-2 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A1042016-09-12012 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI SNPB-3-13 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A0962016-09-13013 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI SSIB-3-4 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A0982016-09-13013 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 33 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A2072016-09-14014 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 37 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16258A2082016-09-14014 September 2016 NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 34 Clarification for September 14, 2016 Public Meeting Project stage: Request ML16246A0222016-10-0505 October 2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items Related to Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400 - MF2409) Project stage: Meeting ML16279A3312016-10-27027 October 2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409) Project stage: Meeting ML16302A4532016-12-12012 December 2016 Closeout of Request for Additional Information Questions That Are No Longer Applicable, Resolution of Generic Safety Issue (GSI) 191 (CAC Nos. MF2400-MF2409) Project stage: RAI ML16351A1502016-12-16016 December 2016 Correction Letter Closeout of Request for Information Questions That Are No Longer Applicable Associated with the Resolution of Generic Safety Issue 191 (CAC Nos. MF2400 - MF2409) Project stage: RAI ML16278A5982017-05-0202 May 2017 Letter, Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400 to MF2409) Project stage: Other ML16278A5992017-05-0202 May 2017 Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400 to MF2409) Project stage: Other ML17137A3252017-05-17017 May 2017 Safety Evaluation of LAR by South Texas Project Nuclear Operating Company to Adopt a Risk-Informed Resolution of Generic Safety Issue-191 Project stage: Approval ML17151A8432017-06-16016 June 2017 OEDO-17-00326 - EDO Response to ACRS Chairman, Safety Evaluation of License Amendment Request by STP Nuclear Operating Company to Adopt a Risk-Informed Resolution of Generic Safety Issue-191 Project stage: Approval ML17055A5002017-07-11011 July 2017 Enclosure 4 to Amendment Nos. 212 and 198, Resolution of Licensee Comments on Safety Evaluation - Risk-Informed Approach to Resolve Generic Safety Issue 191 Project stage: Approval ML17019A0032017-07-11011 July 2017 Attachment 2 to Safety Evaluation - In-Vessel Thermal-Hydraulic Analysis, Issuance of Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191 Project stage: Approval ML17038A2232017-07-11011 July 2017 Issuance of Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191 Project stage: Approval ML17019A0022017-07-11011 July 2017 Safety Evaluation, Enclosure 3 to Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191 Project stage: Approval 2016-05-31
[Table View] |
|
---|
Category:Meeting Briefing Package/Handouts
MONTHYEARML24080A1292024-03-25025 March 2024 March 25, 2024, Pre-Submittal Meeting for STP QA Plan Change ML22243A1312022-08-31031 August 2022 STP Nuclear Operating Company August 31, 2022 Presentation - Pre-Submittal Meeting for Updated Main Steam Line Break and Locked Rotor Dose Consequence Analysis to Address Extended Cooldown Timelines ML22214A0972022-08-0202 August 2022 Presentation by South Texas Project ML22153A2942022-06-17017 June 2022 richardsd-hv-th28 ML21203A2702021-07-26026 July 2021 Presentation - 7-26-21 Meeting with STPNOC Proposed South Texas Project Moderator Temperature Coefficient TS Change ML20125A3362020-05-0404 May 2020 STP Presubmittal Meeting Exigent Accumulator LAR 2020-05-05 Presentation L-2020-LRM-0039 ML19326A0072019-11-22022 November 2019 Pre-Application Public Meeting for Proposed License Amendment Request to Revise the STPEGS Emergency Plan - December 3, 2019 ML16356A0612016-12-21021 December 2016 12/21/2016 South Texas Project Aluminum Bronze Aging Management Request for Additional Information ML16264A3632016-09-20020 September 2016 Pre-Application Public Meeting for Proposed License Amendment Request to Revise the STPEGS Emergency Plan Staff Augmentation Times ML16210A0442016-07-28028 July 2016 STP Nuclear Operating Company Handouts for July 28, 2016 Public Meeting on GSI-191 ML16172A0192016-07-25025 July 2016 STP Aluminum Bronze Selective Leaching Public Meeting - June 21, 2016 ML16202A4872016-07-21021 July 2016 07/21/2016 Talking Points, Supplemental Information for South Texas Project and NRC Public Meeting ML16111A0272016-04-18018 April 2016 STP Nuclear Operating Company Slideshow for Public Meeting on April 21, 2016 for GSI-191 Resolution (CAC No. MF2400-01) ML16062A3042016-03-0202 March 2016 TS 5.3.2 LAR Pre-Submittal Meeting Presentation - March 2, 2016 ML16033A0042016-01-19019 January 2016 Applicant's Slides for STP Meeting on Aluminum Bronze Aging Management ML15274A5992015-10-0101 October 2015 10-01-15 Public Phone Call ML15271A3042015-09-28028 September 2015 Public Phone Call Meeting Description of Supplement 2 to STPNOC Risk-Informed Licensing Application to Address GSI-191 and Respond to GL-2004-02 ML15111A0582015-04-15015 April 2015 Licensee Handouts (2 of 2) for 4/22/2015 Public Meeting Possible Roverd Section on Fixed Filtration (TAC Nos. MF2400-MF2409) ML15111A0562015-04-15015 April 2015 Licensee Handouts (1 of 2) for 4/22/2015 Public Meeting Revised Roverd Section on Large Early Release Frequency Change for Large Break Loss-of-Coolant Accident(Tac Nos. MF2400-MF2409) ML15056A3862015-02-25025 February 2015 NRC Slides for February 25, 2015 Public Meeting with STPNOC Concerning GSI-191 ML15049A4882015-02-18018 February 2015 Slides for February 25, 2015 Public Meeting ML15034A1512015-02-0505 February 2015 Integrated Leak Rate Test (ILRT) from 10 Years to 15 Years. NRC Public Meeting Slides - ILRT License Amendment ML15034A1142015-02-0404 February 2015 STP Risk- Informed Approach to GSI - 191 Roverd 2015 Meeting Blue 1 29 15 ML14192A9842014-08-0404 August 2014 Summary of Pre-Licensing Meeting with STP Nuclear Operating Company to Discuss Future License Amendment Request to Revise TS 6.9.1.6.b.9 to the Leading Edge Flow Meter Technology for Feedwater Flow Measurement ML14120A0112014-04-29029 April 2014 4/29/2014 - South Texas Project Regulatory Approach for Risk-Informed Pilot Submittal Presentation Relating to 10 CFR 50.46c ML13352A1422013-12-16016 December 2013 GSI-191 Licensing Submittal Comparison of Changes Between Rev 1 and Rev 2 ML13352A1242013-12-16016 December 2013 NRC Staff Slides South Texas Project Risk-Informed Generic Safety Issue - 191, December 16, 2013 ML13316B9052013-11-13013 November 2013 Albrz Testing Update for NRR Final ML13150A2162013-05-23023 May 2013 Licensee Slides for 5/23/13 Public Meeting Regarding GSI-191 ML13140A2562013-05-20020 May 2013 Licensee Slides for 5/23/13 Meeting - STP Pilot Submittal for Risk Informed Approach to Resolving GSI-191 ML13023A3342013-02-25025 February 2013 1/15/2013 Summary of Public Meetings Conducted to Discuss Draft Supplemental Environmental Impact Statement Related to Review of South Texas Project, Units 1 & 2, License Renewal Application ML13051A8552013-02-20020 February 2013 Licensee Slides for 3/5/13 Pre-Application Meeting for Proposed Licensing Action to Revise the Fire Protection Program at STP ML13029A4972013-01-15015 January 2013 Slides - Preliminary Site-Specific Results of the Environmental Review for South Texas Project License Renewal ML12297A3312012-11-27027 November 2012 Summary of Prelicensing Meeting with STP Nuclear Operating Company to Discuss Proposed Amendment for Approval of Revised Fire Protection Program Related to Alternate Shutdown Capability for South Texas, Units 1 and 2 ML12264A3202012-10-11011 October 2012 Meeting Handout for 10/11/12 Pre-licensing Meeting License Amendment Request to Revise the Fire Protection Program ML12145A4382012-05-18018 May 2012 Licensee Handout on Calibration and Benchmarking of Single and Two-Phase Jet Cfd Models ML1209004042012-03-30030 March 2012 3-27-2012 - STP Public Meeting Handout from Nuclear Innovation North America, LLC on Fukushima Lessons Learned for South Texas Project Units 3 and 4 ML1205405572012-02-22022 February 2012 Licensee Slides for 03/1/12 Meeting with STP Nuclear Operating Company Meeting to Discuss GSI-191 ML1205405702012-02-22022 February 2012 Uncertainty Modeling of LOCA Frequencies and Break Size Distributions for the STP GSI-191 Resolution ML1205407582012-02-22022 February 2012 STP Summary of January Meeting ML1204400652012-02-0909 February 2012 Licensee Slides for 2/9/12 Meeting Regarding GSI-191 ML1133505632011-12-0101 December 2011 Licensee Presentation from 12/1/11 Meeting Via Conference Call to Discuss Risk-Informed GSI-191,Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance ML1130506322011-10-26026 October 2011 Licensee Handouts from November 1, 2011 Meeting with STP Nuclear Operating Company on GSI-191 ML1127701762011-10-0303 October 2011 Licensee Handouts from 10/3/11 Meeting STP LOCA Frequency ML1127702032011-10-0303 October 2011 Licensee Slides from 10/3/11 Meeting LOCA Frequencies, Final Results ML1123508732011-08-22022 August 2011 Models and Methods Used for Casa Grande ML1123508832011-08-22022 August 2011 LOCA Initiating Event Frequencies and Uncertainties Status Report ML1120707072011-07-26026 July 2011 Meeting Notice with South Texas Project, Units 1 and 2 - Licensee Slide Models and Methods Used for Casa Grande ML1120707292011-07-26026 July 2011 Meeting Notice with South Texas Project, Units 1 and 2 - Licensee Slides Computational Fluid Dynamics Validation Plan ML1118903802011-07-0707 July 2011 Licensee Slides, LOCA Initiating Event Frequencies and Uncertainties(Draft) 2024-03-25
[Table view] |
Text
EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 1 of 9 Quantity of Coatings 1,570 lbs of epoxy coatings on the concrete walls and ceilings
Unit 1 has a larger quantity since the Unit 2 walls are not epoxy coated all the way up to the ceiling and also since the ceilings in Unit 2 are not epoxy coated.
The sump performance evaluation is based on the higher quantity for Unit 1.
The Reactor Cavity is designated as Reactor Containment Building Room 001.
EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 2 of 9 Type of Coatings Brown and Root Technical Reference Document 5A810WQ005-A "Painting Schedule for Service Level I, II, and III Areas" from 1979 calls for the Reactor Cavity walls to be coated using the following Level I concrete coating system defined in HL&P Specification #7A810XS002-A-HL: Nutec 10 Primer/Sealer Nutec #11S Concrete Masonry/Surfacer Nutec #11 Concrete Masonry/Surfacer Reactic (Nutec) #1201 Epoxy Topcoat
Service History Just prior to the initial startup of Unit 1 in 1987 during hot functional testing, repairs were made to the concrete wall coating in the Reactor Cavity. Cracks in the originally installed coating were chased and the adjacent coating was removed. These areas were not top-coated due to the unavailability of that coating at the time. Instead the areas were left with only a base sealer. See Unit 1 photographs.
No other evidence has been found of any repairs since then in Unit 1. Also Unit 2 had no evidence found of any repairs since the startup of the Unit in 1989. A search of work order history and a search of condition reports did not yield any repair requests or coating conditions in either Unit. The maintenance work planner who has been dealing with coatings since startup was not aware of Similar discussion with Engineering personnel who perform Reactor Vessel inspections in Room 001 also did not yield any information concerning coating problems. As seen in the photographs, the current condition of the coatings in both Units appears good.
EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 3 of 9 Current Condition Based on Photographs of the Area Unit 1 Photos EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 4 of 9 Unit 1 Photos EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 5 of 9 Unit 2 Photos from 2013 EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 6 of 9 EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 7 of 9
Manufacturer Qualification Data The test document from Southern Imperial Coatings (filed as STI 525724) states that the Nutec 11/ 11S/ 1201 was qualified to 2E08 rads which is consistent with IEEE STD 323-1974 for normal plus LOCA dose.
However, the epoxy coating for the reactor cavity was deemed by Bechtel during the unqualified coating calculation rack up to be unqualified due to the excessive calculated cumulative radiation dose.
EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 8 of 9 Cumulative Dose Received The epoxy coating for the reactor cavity was deemed to be unqualified due to excessive radiation.
From design criteria document 4E019NQ1009:
The calculated cumulative design dose for the Reactor Cavity for 40 years is: Neutron 2.5E10 Rads Gamma 3.5E9 Rads The accident dose is given as 1.5E8 Rads.
The calculated cumulative dose was determined considering the reactor operated at 100% power every day for 40 years.
The actual capacity factor for each Unit to date is: Unit 1 83.7% from initial operation in 1988 to March 2016 (28 years)
Unit 2 83.2% from initial operation in 1989 to March 2016 (27 years)
Thus the actual cumulative dose to date is: = capacity factor x (operating years / design life years) x calculated lifetime dose = 0.8 x (28/40) x calculated lifetime dose
= 0.56 x calculated lifetime dose Coatings Assessment The epoxy coatings in the Reactor Cavity (RCB Room 001) in both Unit 1 and in Unit 2 were applied as Service Level I coatings during the construction phase of the plant. However, these epoxy coatings for the reactor cavity were subsequently deemed to be unqualified due to the excessive calculated cumulative radiation dose. They were not part of the qualified coatings inspection program and there was no continuity of inspections.
The epoxy coatings in the Reactor Cavity in both Unit 1 and in Unit 2 are now deemed as Qualified But Degraded based upon consideration of the following:
- Initial application was safety related (Service Level I)
- No historical evidence of coating problem conditions requiring repairs since installation
- Recent visual observation (photos and personnel observations) shows no signs of degradation
- Estimated cumulative dose to date is over half of the calculated cumulative dose but no current signs of degradation
- Estimated cumulative dose to date exceeds the test qualification value but no current signs of degradation EPOXY COATINGS IN THE REACTOR CAVITY - STP 4-18-2016 Page 9 of 9 EPRI Report 1014884 "Plant Support Engineering: Degradation Research for Nuclear Service Level I Coatings" concludes that the majority of coating failures and signs of degradation can be attributed to undetected deficiencies that occurred at the time of coating application. These deficiencies are the major cause of coating deterioration during the coating systems' service life. The report also states that high radiation is not believed to be a significant cause of coating degradation in containment.
Since the epoxy coatings in the Reactor Cavity in both Units have been in service for over half of their design life with no signs of degradation, any degradation in the future is not expected.
Thus these coatings are considered Qualified But Degraded.
Debris Characteristics The Reactor Cavity epoxy coatings are considered as Qualified But Degraded and are assumed to fail as chips with the following debris characteristics:
Range (inches) Weight Distribution (%) 1.0 - 2.0 32.0 0.5 - 1.0 9.04 0.25 - 0.50 4.41 0.125 - 0.25 5.02 < 0.125 37.1 as 15.6 mils chips; 12.3 as 6 mils particulate
Transport Evaluation No transport is expected for epoxy coating debris inside the reactor cavity for breaks that do not occur within the reactor cavity because
- for breaks that occur outside the reactor cavity, there is no flow into the reactor cavity,
- the path out of the reactor cavity is sufficiently tortuous that this area will be essentially stagnant, and
- any negligible flow from the reactor cavity that could occur will be to a region on the opposite side of the steam generator compartment from where the sumps are located.
For breaks inside the Reactor Cavity (i.e. at the reactor Vessel nozzle), most of the epoxy coating debris will fail as chips. A portion of the debris that is fine chips and the portion that is particulate is subject to transport. The transport path out of Room 001 is through two 4 inch floor drains in the Reactor Cavity floor that go to the Normal Sump. This sump has a cover plate with an opening for water to go out onto the floor when the sump is full. Thus there will be some transport from the Reactor Cavity to the containment pool.