IR 05000445/2014003

IR 05000445-14-003, 05000446-14-003; 03/28/2014 - 06/26/2014; Comanche Peak Nuclear Power Plant, Units 1 & 2; Integrated Inspection Report, Plant Modifications, Refueling and Other Outage Activities, Radiological Hazard Assessment and Expos

ML14218A072
Person / Time
Site:	Comanche Peak
Issue date:	08/06/2014
From:	Walker W C NRC/RGN-IV/DRP/RPB-A
To:	Flores R Luminant Generation Co
Walker W
References
EA-14-123 IR-14-003
Download: ML14218A072 (76)

Text

August 6, 2014

EA-14-123 Rafael Flores, Senior Vice President and Chief Nuclear Officer Luminant Generation Company, LLC Comanche Peak Nuclear Power Plan P.O. Box 1002 Glen Rose, TX 76043 COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000445/2014003 AND 05000446/2014003 AND NOTICE OF VIOLATION Mr. Flores Comanche Peak Nuclear Power Plant, Units 1 and 2On July 8, 2014, the NRC inspectors discussed the results of this inspection with , and other members of your staf Inspectors documented the results of this inspection in the enclosed inspection report. NRC inspectors documented seven findings of very low safety significance (Green) in this repor Five of these findings involved a violation of NRC requirement One of these violations was determined to be Severity Level IV under the traditional enforcement proces The NRC is treating four of the violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy, which can be found http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.htm The enclosed inspection report discusses a violation associated with a finding of very low safety significance (Green). The NRC evaluated this violation in accordance with Section 2.3.2.a of the NRC Enforcement Polic We determined that this violation did not meet the criteria to be treated as a non-cited violation because you- The NRC has determined that the reason, corrective actions taken and planned to address recurrence, and the date when full compliance will be achieved for this violation is adequately addressed and captured on the docket in a letter from Luminant Generation Company, LLC, dated June 10, 2014 (ML14188C054). Therefore, you are not required to respond to this letter unless the description herein does not accurately reflect your corrective actions or your positio In that case, or if you choose to provide additional information, you should follow the instructions specified in the enclosed Notic R. Flores - 2 - Further, inspectors documented licensee-identified violations which were determined to be of very low safety significance in this repor The NRC is treating these violations as NCVs consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the If you disagree with a cross-cutting aspect assignment in this report, then you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant In accordance with Title 10 of the Code of Federal Regulations Rules of Practice and Procedure, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). -- NPF-87, NPF-89 05000445/2014003 and 05000446/2014003 R. Flores -2- Further, inspectors documented licensee-identified violations which were determined to be of very low safety significance in this repor The NRC is treating these violations as NCVs consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the If you disagree with a cross-cutting aspect assignment in this report, then you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant In accordance with Title 10 of the Code of Federal Regulations Rules of Practice and Procedure, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). -- NPF-87, NPF-89 05000445/2014003 and 05000446/2014003 ADAMS ACCESSION NUMBER: SUNSI Review By: WWalker ADAMS Yes No Publicly Available Non-Publicly Available Non-Sensitive Sensitive OFFICE SRI:DRP/A SPE:DRP/A C:DRS/EB1 C:DRS/EB2 C:DRS/OB NAME JKramer/dll TBuchanan TFarnholtz JDixon VGaddy SIGNATURE /RA/E-Walker /RA/ /RA/ /RA/ /RA/ DATE 7/31/14 8/5/14 7/29/14 7/30/14 7/30/14 OFFICE D:DRS/PSB1 C:DRS/PSB2 C:DRS/TSB R4/ACES BC:DRP/A NAME MHaire HGepford GMiller VCampbell WWalker SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ DATE 7/29/14 7/29/14 8/4/14 8/4/14 8/5/14 OFFICIAL RECORD COPY E-EMail COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000445/2014003 AND 05000446/2014003 AND NOTICE OF VIOLATION DISTRIBUTION: Regional Administrator (Marc.Dapas@nrc.gov) Deputy Regional Administrator (Kriss.Kennedy@nrc.gov) Acting DRP Director (Troy.Pruett@nrc.gov) Acting DRP Deputy Director (Michael.Hay@nrc.gov) DRS Director (Anton.Vegal@nrc.gov) DRS Deputy Director (Jeff.Clark@nrc.gov) Senior Resident Inspector (John.Kramer@nrc.gov) Resident Inspector (Rayomand.Kumana@nrc.gov) Branch Chief, DRP/A (Wayne.Walker@nrc.gov) Senior Project Engineer, DRP/A (Ryan.Alexander@nrc.gov) Acting Senior Project Engineer (Theresa.Buchanan@nrc.gov) Project Engineer, DRP/A (Brian.Cummings@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Balwant.Singal@nrc.gov) Branch Chief, DRS/TSB (Geoffrey.Miller@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) ACES (R4Enforcement.Resource@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Technical Support Assistant (Loretta.Williams@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) RIV/ETA: OEDO (Anthony.Bowers@nrc.gov) ROPreports Electronic Distribution via Listserv for Comanche Peak Nuclear Power Plant Enclosure NOTICE OF VIOLATION Docket Nos. 50-445, 50-446 Comanche Peak Nuclear Power Plant, Units 1 and 2 License Nos. NPF-87, NPF-89 EA-14-123 During an NRC inspection conducted between March 28, 2014 and June 26, 2014, a violation of NRC requirements was identifie In accordance with the NRC Enforcement Policy, the violation is listed below: License Condition 2.G for Unit 1 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 78 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 License Condition 2.G for Unit 2 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 describes the fire protection quality assurance progra The fire protection quality assurance program states that measures shall be established to ensure that conditions adverse to fire protection such as failures, malfunctions, deficiencies or deviations, defective components, uncontrolled combustible material and non-conformances are promptly identified, reported, and correcte Contrary to the above, as of June 19, 2014, the licensee failed to ensure that two conditions adverse to fire protection were promptly correcte Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for two violations associated with the fire protection progra These violations were non-cited Violation 05000445/2009004-04; 05000446/2009004-non-cited Violation 05000445/2011007-02; 05000446/2011007-ns were untimely since the conditions adverse to fire protection have existed for more than five years and three years, respectivel This violation is associated with a Green significance determination process findin The NRC has concluded that information regarding the reason for the violation, the corrective actions taken and planned to correct the violation, and the date when full compliance will be achieved is already adequately addressed on the docket in a letter from Luminant Generation Company, LLC, dated June 10, 2014 (ML14188C054). However, you are required to submit a written statement or explanation pursuant to 10 CFR 2.201 if the description therein does not accurately reflect your corrective actions or your positio In that case, or if you choose to -14-123, and send it to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, Region IV, and a copy to the NRC Resident Inspector at the Comanche Peak Nuclear Power Plant facility, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-000 If you choose to respond, your response will be made available electronically for public Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.htm Therefore, to the extent possible, the response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redactio Dated this 6th day of August 201 , 05000446 NPF-87, NPF-89 05000445/2014003 and 05000446/2014003 Luminant Generation Company LLC Comanche Peak Nuclear Power Plant, Units 1 and 2 6322 N. FM-56, Glen Rose, Texas 05000446/2014003-Comanche Peak Nuclear Power Plant, Units 1 and 2; Integrated Inspection Report- Comanche Peak Nuclear Power Plant findings of very low safety significance (Green) are documented in this repor Five of these findings involved violations of NRC requirementsthese violations was determined to be Severity Level IV under the traditional enforcement proces T-The personnel who conducted the walk-downs ensure that anyone actually verified the physical details of the cable rout The licensee suspended the modification activities, repaired the damaged offsite power cable, and restored offsite power to the -The licensee entered the finding into the corrective action program as Condition Report CR-2013-012287. -licensee failed to ensure proper oversight of contractors to ensure deviations from standards and expectations were promptly corrected [H.2]. (Section 4OA3.2) Green. T-Specifically, the licensee failed to ensure the work instructions met the requirements of Procedure STA-of Maintenance and Work Activities, Revision 3 The licensee suspended the modification activities, repaired the damaged offsite power cable, and restored offsite power to the -The licensee entered the finding into the corrective action program as Condition Report CR-2013-012287. -licensee failed to ensure that work planning personnel planned for the possibility of mistakes and latent issues and did not implement appropriate error reduction tools [H.12]. (Section 4OA3.2) T- uninterruptible power supplyuninterruptible power supplyThe licensee entered the finding into the corrective action program as Condition Report CR-2013-00229 The inspectors identified a non-cited violation of 10 CFR 50.59, for failure to conduct an adequate written evaluation and submit a license amendment for a change to the facility that required a technical specification amendmen Specifically, the licensee changed Procedure NUC-211, , to allow for storage of uprated fuel in Region II (high density racks) of the spent fuel pool using a methodology for fuel burnup penalties that had not been previously approved by the NRC and therefore, required an amendment to Technical Specification implementatio Subsequently, the licensee stopped all fuel movement in Region II of the spent fuel pool unless notifying the NRC prior to the movemen CR-2014-00469 The failure to perform an adequate 10 CFR 50.59 evaluation and obtain prior NRC approval for a change to the facility that involved a change to the technical specifications was a performance deficienc The inspectors concluded that this issue involved traditional regulatory functio This performance deficiency is more than minor because it was associated with the reactivity control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or event Because the significance determination process does not directly address spent fuel pool criticality, a senior reactor analyst evaluated this issue using Appendix MBased on calculations provided by the licensee, the analyst determined that even with all uncertainties included in the calculations, the spent fuel pools would remain subcritical under all conditions, including a complete dilution of the borated wate The analyst qualitatively considered a completed dilution of the spent fuel pools to be a very low probability even Therefore, the analyst concluded that this issue was of very low safety significance (Green). Because this issue was considered to be Green, it is treated as a Severity Level IV violation in traditional enforcemen (Section 1R18) T-- - The operators immediately stopped the filling activity and restored cooling water to the . The licensee entered the finding into the corrective action procedure as Condition Report CR-2014-00411 licensee failed to ensure that the work process identified and managed the risk commensurate with the work [H.5]. (Section 1R20) -- the licensee failed to adequately inform the worker of current radiological dose rates prior to entry and the worker entered a posted high radiation area without proper knowledge of the radiological conditions (dose rates). Consequently, the workers received unanticipated high dose rate alarms on their electronic alarming dosimeters at 563 millirem per hour, 274 millirem per hour, and at 750 millirem per hour, respectivel As immediate corrective actions, the licensee performed follow-up surveys, coached the involved individuals, and reviewed the radiologically controlled area entry card requirement The licensee entered the three issues into the corrective action program as Condition Reports CR-2013-004154, CR-2014-003464, and CR-2014-00399 The failure to provide workers with proper knowledge of high radiation area radiological conditions prior to entry is a performance deficienc The performance deficiency is more than minor because it impacted the program and process attribute (exposure control) of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiatio Specifically, worker entry into high radiation areas without knowledge of the radiological conditions placed them at increased risk for unnecessary radiation exposur (Section 2RS1) - The inspectors directly observed the following nondestructive examinations: ---- --- --- - --- - ---- ---- --- --- - --- --- --- --- ---- ---- ---- --- --- The inspectors reviewed records for the following welding activities: reactor vessel upper and lower head penetrations to determine whether the licensee identified any evidence of boric acid challenging the structural integrity of the reactor head components and attachment The inspectors also verified that the required inspection coverage was achieved and limitations were properly recorde The inspectors reviewed whether the personnel performing the inspection were certified examiners to their respective nondestructive examination metho STA-737, - -- -- - - - - --he inspectors also reviewed the results of the steam generator secondary side sludge lancing and foreign object search and retrieval inspections - .1 a. Inspection Scope On May 19, 2014, These activities constitute completion of one quarterly licensed operator requalification program sample as defined in Inspection Procedure 71111.11. b. Findings No findings were identified. .2 a. Inspection Scope The inspectors observed the performance of on-main control roo At the time of the observations, the plant was in a period of heightened activity or ris Tce of the following activities: These activities constitute completion of one quarterly licensed operator performance sample as defined in Inspection Procedure 71111.11. b. Findings No findings were identifie No findings were identifie uninterruptible power supply air conditioning unit X-01 cycling Condition Report CR-2014-001919, Unit 1, diesel generator 1-01 lubricating oil pipe crack --- -- ----failure to open --- No findings were identifie The inspectors reviewed Unresolved Item 05000445/2012004-03; 05000446/2012004-03, Potential Failure to Follow 10 CFR 50.59 for a Change to the Spent Fuel Pool Configuration to determine if a violation existed for revising the procedure that controlled off-loading spent fuel bundles to the spent fuel poo The inspectors identified a non-cited violation of NRC requirement This unresolved item is close Introductio The inspectors identified a non-cited violation of 10 CFR 50.59 for failure to conduct a written evaluation and submit a license amendment for a change to the facility that required a technical specification amendmen Specifically, the licensee changed Procedure NUC-, to allow for storage of uprated fuel in Region II (high density racks) of the spent fuel pool using a methodology for fuel burnup penalties that had not been previously approved by the NRC and therefore, required an amendment to Technical Specification Descriptio On August 28, 2007, the licensee submitted an application for a license amendment for a stretch power uprate of approximately six percent reactor powe Included in this application was a proposed change to Technical Specification 3.7.17 to support the eventual loading of the power uprate fuel to the spent fuel pool and associated criticality analyse For ease of review, the spent fuel pool criticality analysis portion of the amendment was separated from the stretch power uprate amendmen The license amendment approving the licensed reactor power uprate was issued on June 27, 200 However, based on an NRC technical staff request, the licensee submitted additional supplemental information for the spent fuel pool criticality analysis amendment on June 30, 200 A formal request for additional information was issued by the NRC for the spent fuel pool criticality analysis amendment on November 19, 200 The licensee submitted several separate responses to the request for additional information questions during 2008 and 200 The NRC technical staff issued a draft denial of the spent fuel pool criticality analysis amendment by a letter dated July 10, 200 In response to this draft denial of the license amendment, the licensee formally withdrew the license amendment by letter dated August 20, 2009, which was acknowledged by the NRC the next da Comanche Peak Units 1 and 2 both operated for two 18-month refueling intervals from 2009 to 2012 at the uprated reactor power condition However, the licensee had not resolved the issues associated with receiving a license amendment to allow for storage of the uprated fuel in the spent fuel poo In February 2009, the licensee performed a 10 CFR 50.59 screening of Procedure NUC-211 to address the potential storage of uprated fuel in the spent fuel pools if a license amendment was not approved by the end of the operating cycl The licensee added a precaution to Procedure NUC-and beyond, and assemblies from Unit 2 Cycle 12 and beyond, should NOT be stored in Region II until the Technical Specifications are revised to consider the effects of stretch because the Region II high density storage racks technical specification implemented a number of limitations on storage configurations based on fuel enrichment and fuel burnu Procedure NUC-211 did allow for storage of uprated fuel in the low density fuel racks, which was analyzed for any storage configuration regardless of fuel enrichmen The precautions of Procedure NUC-211, prohibiting the storage of uprated fuel in Region II of the spent fuel pool remained in effect for the duration of the first operating cycles following the approval of the power uprat The first storage of Unit 1 uprated fuel to Region I (the low density-unrestricted racks) of the spent fuel pool occurred on April 3, 2010, following Unit 1 cycle 1 The licensee determined that they did not have sufficient capacity in the spent fuel pools to discharge all fuel in Region I of the spent fuel racks and accommodate other fuel management consideration Thus, the licensee contacted a vendor to determine if they could provide an analysis to justify movement of the uprated fuel to Region II of the spent fuel rack On September 29, 2010, the vendor provided the licensee with the results of this analysi This analysis stated that the burnup versus enrichment curves for Technical Specification 3.7.17 for storage of fuel in Region II of the spent fuel racks were nonconservative when applied to fuel depleted at uprated condition The uprated fuel remained in Region I (unrestricted low density racks) because of this informatio for uprate conditions should the licensee desire to move the uprated spent fuel to Region II of the spent fuel pool In December 2010, the licensee relocated uprated spent fuel from Region I to Region II of the spent fuel pool Since the fuel discharge curves for spent fuel subject to uprated conditions of Technical Specification 3.7.17 were nonconservative, the licensee incorrectly invoked the direction of NRC Administrative Letter 98-noted that the NRC letter states that if a technical specification is found to be nonconservative, administrative controls to ensure nuclear safety is adequately protected is an acceptable short-term solutio The licensee believed that these compromise solutions of the administrative letter applied equally to current/past plant design as well as a desired future plant configuratio Thus, on December 15, 2010, the licensee revised Procedure NUC-211 to allow storage of uprated fuel in Region II of the spent fuel pool The 10 CFR 50.59 screening incorrectly stated that a technical specification amendment was unnecessar This method of applying burnup penalties was not reviewed and approved by the NRC and Technical Specification 3.7.17 was not Procedure NUC-211, that allowed the use of administrative controls to discharge uprated fuel to Region II of the spent fuel pools without prior NRC approval, was a violation of 10 CFR 50.5 obtain prior NRC approval for a change to the facility that involved a change to the technical specifications was a performance deficienc The inspectors concluded that this issue involved traditional enforcement because it had the potential for impacting the This performance deficiency is more than minor because it was associated with the reactivity control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or event Because the significance determination process does not directly address spent fuel pool criticality, a senior reactor analyst evaluated this issue using Appendix MBased on calculations provided by the licensee, the analyst determined that even with all uncertainties included in the calculations, the spent fuel pools would remain subcritical under all conditions, including a complete dilution of the borated wate The analyst qualitatively considered a completed dilution of the spent fuel pools to be a very low probability even Therefore, the analyst concluded that this issue was of very low safety significance (Green). Because this issue was considered to be of Green safety significance, it is treated as a Severity Level IV violation in traditional enforcemen Title 10 CFR Part 50.59 states, in part, that licensees may make changes to the facility as described in the safety analysis report, without prior NRC approval, provided the change does not involve a change to the technical specification Contrary to the above, on December 15, 2010, the licensee made a change to the facility that involved a change to the technical specifications without prior NRC approva Specifically, the licensee revised Procedure NUC-211 to allow storage of uprated fuel in Region II of the spent fuel pool and failed to obtain a license amendment to Technical Subsequently, the licensee stopped all fuel movement in Region II of the spent fuel pool unless notifying the NRC prior to the movemen CR-2014-004693, -ection 2.3.2.a of the NCV 05000445/2014003-01; 05000446/2014003-01Pool Configuratio May 12, 2014, Unit 1, service water pump 1-01 testing following maintenance -- - No findings were identifie Introductio T-- Descriptio On April 8, 2014, operators attempted to restore the Unit 2 train B component cooling water system to service following maintenanc With the component cooling water trains cross-connected, an operator began to fill train B by throttling open train B heat exchanger inlet valve 2CC-0055, a 24-inch butterfly valv Almost immediately, the control room received a component cooling water surge tank empty alar The operating train A component cooling water system automatically isolated cooling to the non-safeguards loop of component cooling water, resulting in loss of cooling flow to spent fuel pool heat exchanger X-0 The control room operators immediately stopped the filling of train B and restored train A to servic The licensee maintained spent fuel pool cooling using spent fuel pool heat exchanger X-01, supplied by Unit The non-safeguards loop of Unit 2 component cooling water was restored after approximately two minute During that time, the shutdown risk profile as a result of having only one available train of spent fuel pool cooling available. The licensee allowed system restoration to be performed using the clearance release process to control the activities for filling the syste The operators had discussed the sequence of repositioning valves as they removed clearance tags and were aware that opening the isolation valve would result in lowering surge tank leve However, the surge tank level was low when the work activity was starte The clearance release -0055 slowly to ensure CCW surge tank flow. The inspectors determined that the instructions provided were not appropriate to the circumstance The inspectors identified the following deficiencies in the instruction The instructions did not include precautions to ensure there was enough water in the surge tan The instructions should have specified a different valve to control the fill rat In addition, the licensee operations and maintenance personnel did not fully consider and mitigate the risk of performing the system restoration. The licensee entered the event into their corrective action procedure as Condition Report CR-2014-00411 licensee failed to ensure that the work process identified and managed the risk commensurate with the work [H.5]. April 8, 2014the licensee used the clearance release process to control the activities for filling a component cooling water system and caused a loss of cooling water flow to a spent fuel pool heat exchanger---ection 2.3.2.a of the - - - ----- May 2, 2014, Unit 1, water inventory balance in accordance with Procedure OPT-eactor Coolant Syst -- - - - - Introductio The inspectors identified an unresolved item related to maintaining the effectiveness of the listandard 50.47(b)(4), which requires, in part, that a standard emergency classification and action level scheme is in use by the license Specifically, several main steam line monitors were out of service for extended periods of time without apparent contingency actions in place to ensure the correct emergency action level would be implemented. Descriptio On November 20, 2013, the licensee initiated Condition Report CR-2013-011914 identifying that the main steam line radiation monitors had a trend of being out-of-service for significant time period Monitor 1-RE-2328 was out of service for 110 and 210 days on two separate occasion Monitor 1-RE-2326 was out of service for 77 and 157 days on two separate occasion Monitor 1-RE-2325 was out of service for 61 day Four other monitors from the two units had been out of service, some more than once, for periods of five days or les There are four online main steam line monitors for each uni The licensee addressed the trend by trouble-shooting, repairing, and replacing detectors. The main steam line radiation monitors are important to emergency preparedness because they are inputs into the emergency action levels and define the initiating conditions related to abnormal radiation releases/radiation effluent emergency declarations. The inspectors determined that the licensee had taken appropriate action to initiate corrective action and repai The licensee also tracked the out of service time of the monitors as operational focus items and in the station tactical equipment issues lis All eight main steam line monitors are currently in service with zero out of service days in 201 However, there was no evidence that contingency actions were implemented to maintain the approved emergency action level scheme when the monitors were out of service. Title 10 CFR 50.54(q) requires licensees to maintain the effectiveness of an emergency plan that meets the requirements in the planning standards of 50.47(b). Title 10 CFR 50.47(b)(4) requires a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee. This issue was identified as an unresolved item because the NRC has not determined whether the licensee has adequately implemented planning standard 10 CFR 50.47(b)(4). Specifically, the NRC has not determined whether the emergency action level initiating condition was rendered ineffective, such that, any general emergency would not be declared for a particular off-normal event in an accurate and timely manner or in a degraded manner. The licensee has entered this issue into the corrective action program as Condition Report CR-2014-00587 This issue is identified as unresolved item URI 05000445/2014003-03; 05000446/2014003-03 - - -- on three separate occasionsinform workers of the high radiation area radiological conditions prior to entry. April 14, 2013, March 30, 2014, and April 7, 2014, workers entered posted high radiation areas without proper knowledge of the radiological conditions (dose rates) in the areas accesse As a result, the electronic alarming dosimeters worn by the workers alarmed on high dose rate. - --- --- --- - - -- --- - --- --- ------ As immediate corrective actions for each of these events, the licensee performed follow-up surveys, coached the involved individuals, and reviewed the requirement As noted above, these three issues were entered into the dition Reports CR-2013-004154, CR-2014-003464, and CR-2014-003997, respectively. The failure to provide workers with proper knowledge of high radiation area radiological conditions prior to entry is a performance deficienc The performance deficiency is more than minor and a violation of Technical Specification 5.7.1 because it impacted the program and process attribute (exposure control) of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiatio Specifically, worker entry into high radiation areas without knowledge of the radiological conditions placed them at increased risk for unnecessary radiation exposur Example 6(h) of IMC 0612, Appendix E, describes a similar example to this performance issue where the workers were authorized to work in a high radiation area, but were not made aware of the radiological conditions as authorized by their radiation work permi Therefore, as provided in Example 6(h), the inspectors determined that the performance deficiency was more than mino Technical Specification 5.7.1, require, in part, that (b) access to, and activities in, high radiation areas shall be controlled by means of a radiation work permit that includes specification of radiation dose rates in the immediate work areas and (e) entry into high radiation areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of the Contrary to the above, on April 14, 2013, March 30, 2014, and April 7, 2014, workers entered high radiation areas with radiation dose rates in excess of the established controls in the radiation work permit and without being made knowledgeable of these elevated dose rates in the immediate work area Specifically, on these three occasions, workers entered high radiation areas with dose rates of 100 millirem per hour or greater (563 millirem per hour, 274 millirem per hour, and 750 millirem per hour) without being informed of these dose rates and in excess of electronic alarming dosimeter setpoints established in the s for controlling worker dos Since this violation is of very low safety significance and was entered into the CR-2013-004154, CR-2014-003464, and CR-2014-003997, - --Failure to Adequately Brief Workers on Radiological Conditions Prior to Entry into High Radiation Areas The inspectors assessed licensee performance with respect to maintaining occupational individual and collective radiation exposures as low as is reasonably achievable (ALARA) Site-specific ALARA procedures and ollective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements ALARA work activity evaluations/postjob reviews, exposure estimates, and exposure mitigation requirements The methodology for estimating work activity exposures, the intended dose outcome, the accuracy of dose rate and man-hour estimates, and intended versus actual work activity doses and the reasons for any inconsistencies Records detailing the historical trends and current status of tracked plant source terms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry - -- - -- - - uninterruptible power supply air conditioning unit X-01 -- -- -- - - Introductio T- uninterruptible power supply Descriptio On March 5, 2013, the inspectors observed that the component cooling water supply valve for the uninterruptible power supply air conditioning unit X-01 was cyclin The inspectors informed the licensee of the observatio The licensee found a -in the condense The licensee declared the air conditioning unit inoperable and repaired the leak. The licensee performed a past operability review of the uninterruptible power supply heating, ventilation, and air conditioning syste The system consists of dedicated uninterruptible power supply room emergency fan coil units in each room and two electrical independent and redundant air conditioning trains that, either of which, can support all four safety-related uninterruptible power supply room This gives the system a unique 300 percent capabilit Even with air conditioning unit X-01 inoperable, the licensee concluded that the required actions were always met for Technical Specification 3.7.20, two uninterruptible power supply heating, ventilation, and air conditioning train The inspectors reviewed the past operability and did not identify any di cause of the component failure was excessive heating of the copper fittings during the brazing process when installing a replacement heat exchange Procedure MSG-1001, Revision 0, provides the requirements and criteria for repairs and replacement of piping system Step 8.21.1 requires, in part, that when brazing 3/4-inch diameter and larger copper tubing, to exercise care to assure the tubing is not overheate The inspectors concluded that personnel overheated the copper tubing and caused the uninterruptible power supply - Procedure MSG-1001, Revision 0, provides the requirements and criteria for repairs and replacement of piping system Step 8.21.1 requires, in part, that when brazing 3/4 inch diameter and larger copper tubing, to exercise care to assure the tubing is not overheate September 25, 2009Procedure MSG-1001Step 8.21.1 exercise care to assure the copper tubing was not overheated---ection 2.3.2.a of the -- 2) Failure to Correct Fire Protection Violations in a Timely Manner - - -- ------- -- -Potential for Loss of Remote Shutdown Capability During a Control Room Fire - 05000445/2013-003-0005000445/2013-003-00 -- ---- - - --- -------- - -- The analyst used the fire ignition frequency for the control room (FIFCR) and the cable spreading room (FIFCSR) listed in the Comanche Peak Steam Electric Station Individual Plant Examination of External Events for Severe Accident Vulnerabilities, Revision 0, as the best available informatio The analyst multiplied each fire ignition frequency by a severity factor (SF) and a nonsuppression probabilit For the control room, the nonsuppression probability (NPCR) indicated the probability that operators failed to extinguish the fire within 20 minutes (assuming 2 minutes for detection), which required control room evacuatio For the cable spreading room, the nonsuppression probability indicated the probability that the automatic Halon system failed (NPCSR-A) and the probability that the fire brigade failed to manually suppress the fire prior to damage that required control room evacuation (NPCSR-M). The resulting control room (FCR-EVAC) and cable spreading room (FCSR-EVAC) evacuation frequencies were: FCR-EVAC = FIFCR * SF * NPCR = 1.9E-2/yr * 0.1 * 0.013 = 2.5E-5/yr FCSR-EVAC = FIFCSR * SF * NPCSR-A * NPCSR-M = 3.2E-3/yr * 0.1 * 0.05 * 0.24 = 3.8E-6/yr The licensee confirmed that the Unit 2 control room has 116 electrical panels for both Unit 2 and common equipment, and the cable spreading room has 80 cabinets (54 termination racks and 26 electrical panels). The analyst used a value of 0.6 for the conditional probability of this hot shor The analyst obtained this value from Inspection Manual Chapter 0609, Appendix F, Table 2.SP Factors Dependent on Cable Type and The analyst noted that this alternative shutdown scenario would only impact risk if the credited centrifugal charging pump was running at the time of the postulated fir The analyst estimated that the probability the credited charging pump was running (PPUMP) to be The analyst calculated a bounding change in core damage frequency for the control room (CR) and cable spreading room CSR) as follows: CR = FCR-EVAC * PPUMP * ((5/116) * 0.6 + (1/116) * (0.6 + .62)) = 2.5E-5/yr * 0.5 * 0.033 = 4.1E-7/yr CSR = FCSR-EVAC * PPUMP * (2/80) * = 3.8E-6/yr * 0.5 * 0.015 = 2.9E-8/yr Because the postulated fire ignition frequencies for the control room and cable spreading room are independent of each other, the total change in core damage frequency can be determined by a simple addition of the change in core damage frequency from the two rooms calculated separatel The resulting overall change in core damage frequency for this example was calculated to have an upper bound of 4.4E-7/yr (Green). This frequency was considered to be bounding because it assumed: A fire induced hot short in the applicable cabinets would cause a volume control tank outlet valve to spuriously close The conditional core damage probability given either a control room or cable spreading room fire with evacuation and the spurious closure of a volume control tank outlet valve was equal to one The performance deficiency accounted for the entire change in core damage frequency (i.e., the base line core damage frequency for this event was zero) In accordance with the guidance in Inspection Manual Chapter 0609, Appendix H, the senior risk analyst screened the performance deficiency for its potential risk contribution to large early release frequency since the bounding change in core damage frequency provided a risk significance estimate greater than 1E-7/y Given that Comanche Peak has a large, dry containment and that control room abandonment sequences do not include steam generator tube ruptures or intersystem loss of coolant accidents, the analyst determined that this example was not significant with respect to large early release frequenc The analyst determined this example was of very low risk significance (Green). -- - ----- The analyst included the remaining nine fire damage scenarios included in the 2011 evaluatio The analyst determined that the cabling associated with the emergency core cooling system in the nonaffected train was not located within the affected zone of the fir The analyst made a bounding assumption that all equipment in either train A or train B was lost due to the fire scenario by failing the train A or train B 6.9-kV busse The analyst used the Comanche Peak SPAR model, Revision 8.28, downloaded July 1, 2014, to calculate the conditional core damage probability for each fire scenario that included the affected power-operated relief valve and its associated block valve failed open with no recover The analyst used a cutset truncation of 1E-1 The following table summarizes the Phase 3 evaluation results: Ignition Source Fire Ignition Frequency CCDP 2EA2 1.88E-6 1.13E-2 2.12E-8 2EA2 (HEAF) 1.62E-6 1.13E-2 1.83E-8 2EB3-2 1.08E-6 1.03E-2 1.11E-8 2EB4-2 1.08E-6 1.13E-2 1.22E-8 T2EB3 1.08E-7 1.03E-2 1.11E-9 Transients - 3.06E-6 1.03E-2 3.15E-8 Transients - 3.40E-7 1.03E-2 3.50E-9 Transients - 3.06E-6 1.13E-2 3.46E-8 Transients - 3.40E-7 1.13E-2 3.84E-9 Total 1.37E-7 The resulting overall change in core damage frequency for this example was calculated to have an upper bound of 1.37E-7/yr (Green). In accordance with the guidance in Inspection Manual Chapter 0609, Appendix H, the senior risk analyst screened the performance deficiency for its potential risk contribution to large early release frequency since the bounding change in core damage frequency provided a risk significance estimate greater than 1E-7/y Given that Comanche Peak has a large, dry containment and that these scenarios do not include steam generator tube ruptures or intersystem loss of coolant accidents, the analyst determined that this example was not significant with respect to large early release frequenc The analyst determined this example was of very low risk significance (Green). -failed to include the identification and management of risk commensurate to the work performed [H.5]. License Condition 2.G for Unit 1 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 78 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 License Condition 2.G for Unit 2 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 Section 13.3B.5 of the Final Safety Analysis Report describes the fire protection quality assurance progra The fire protection quality assurance program states that measures shall be established to ensure that conditions adverse to fire protection such as failures, malfunctions, deficiencies or deviations, defective components, uncontrolled combustible material and nonconformances are promptly identified, reported, and correcte Contrary to the above, prior to June 19, 2014, the licensee failed to ensure that two conditions adverse to fire protection were promptly correcte Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for two violations associated with the fire protection progra These violations were non-cited Violation 05000445/2009004-04; 05000446/2009004-Inadequate Postfire Safe Shutdown Procedurenon-cited Violation 05000445/2011007-02; 05000446/2011007-Failure to Mitigate or Correct Potential Single Spurious Failuresto fire protection have existed for more than 5 years and 3 years, respectivel The licensee entered this issue into its corrective action program as Condition Report CR-2014-00771 The licensee maintained the compensatory measures that were in place for these issue These compensatory measures include hourly fire watches, changes to the safe shutdown procedures, and administrative changes to the fire protection progra Because the licensee failed to restore compliance within a reasonable period of time after these violations were initially identified, this violation is being treated as a cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Polic This is a violation of License Condition 2.G for Units 1 and A Notice of Violation is included with this report: VIO 05000445/2014003-06; 05000446/2014003-06, Failure to Correct Fire Protection Violations in a Timely Manne (Closed) Licensee Event Report 05000445/2013-003-Auxiliary Feedwater Pumps and Emergency Diesel Generators Due to a Loss of Both The inspectors performed a review of a licensee event report documenting an activation of emergency systems that occurred on December 4, 2013, in which the diesel generators and auxiliary feedwater pumps started on a loss of both sources of offsite powe The inspectors examined associated procedures, work orders, condition reports, Introductio T-- Descriptio In 2012, the licensee initiated a permanent design modification of the offsite AC power system to add transformer XST1A which could be used in place of transformer XST The modification was implemented using Procedure ECE-5.01 Control Program,Revision 24, and Procedure ECE-5.01-Change Process, Revision 1 The modification was performed by contractor personnel. As part of the design process, the licensee conducted field walk-downs to identify locations of cables and switch boxe The contractor organization used personnel to conduct the walk-dow One was an engineer-in-training who had been in the organization for several months, and one was a superintendent who had worked at the site for several year The route for the transformer XST1 cables going to Unit 2 passed through an infrequently accessed overhead cable spac Approximately halfway along this path, the Unit 2 cables from startup transformer XST2 entered the spac The cables then exited the overhead space and into the Unit 2 normal switchgear roo The personnel who conducted the walk-down incorrectly identified the cables from transformer XST2 as being the cables from transformer XST1. The responsible engineer accepted the inputs from the engineer-in-training without independently verifying the accurac The responsible engineer did not perform a complete walk-down of the modificatio The inputs were then used to develop drawings and work orders that formed the bases for the modificatio Procedure ECE-5.01-08, step 3.1.3.2 required, in part, that the responsible engineer provides sufficient details to fully describe the chang The details provided were insufficient because they were incorrect. On July 18, 2012, the licensee approved the first version of the design modificatio All further design work in Unit 2, including work locations, was based on the incorrect informatio The design included installation of switch boxes where the cables from transformers XST1, transformer XST1A, and the safety busses would terminate to allow the busses to be supplied by either transforme These boxes were incorrectly installed in the path of the transformer XST This design error contributed to the workers cutting the cables for transformer XST2 instead of transformer XST1, resulting in a loss of offsite power to both units. After the loss of offsite power, the licensee immediately stopped all work associated with the modificatio The inspectors reviewed the design modification and drawings, performed walk-downs of the transformer cables, and interviewed the responsible personne The inspectors determined that the personnel who conducted the walk-downs did noorganization did not ensure that anyone actually verified the physical details of the cable rout Although the licensee had an expectation that a qualified engineer would verify the details of the change, a qualified engineer did not verify the details, and instead relied on an engineer-in-training engineer and a senior worker. -licensee failed to ensure proper oversight of contractors to ensure deviations from standards and expectations were promptly corrected [H.2]. The licensee entered the finding into the corrective action program as Condition Report CR-2013-01228 The issue is being characterized as FIN 05000445-- Introductio T-Specifically, the licensee failed to ensure the work instructions met the requirements of Procedure STA-, Revision 3 Descriptio On December 4, 2013, transformer XST2 was supplying offsite power to both units while transformer XST1 was out of servic Workers attempted to connect power supply cables from transformer XST1 to newly installed switch boxes 04Y and 05Y in the Unit 2 normal switchgear roo The workers were not aware that the switch boxes 04Y and 05Y had been installed in the wrong locatio The workers at switch box 05Y cut into an energized 6.9 kV cable from transformer XST2, resulting in the transformer isolating from the safety-related busses of both units and a complete loss of offsite power to safety-related busse The inspectors reviewed the work instructions provided to the worker The inspectors Procedure STA-,Revision 3 Step 6.7.5 of the procedure requires, in part, that: (1) equipment is made safe for work to proceed; the equipment was not made safe because the work instructions did not include appropriate steps to verify the cables were de-energized (2) adequate work instructions are prescribed that identify critical steps and industry operating experience is used; adequate work instructions were not prescribed and the work instruction failed to adequately incorporate recent operating experience involving workers cutting a wrong cable on October 30, 2013, and (3) proper inspection or verification requirements are designated; proper verification requirements were not designate licensee failed to ensure that work planning personnel planned for the possibility of mistakes and latent issues and did not implement appropriate error reduction tools [H.12]. The licensee entered the finding into the corrective action program as Condition Report CR-2013-01228 The issue is being characterized as FIN 05000445-- -- -The enforcement aspects of this finding are discussed in Section 4OA This licensee event report is close .4 Unusual Event declared an unusual event as a result of a Halon release into the Unit 2 cable spreading room and the lowering of oxygen levels below a safe habitable level, HU .1 (Closed) Unresolved Item 05000445/2013005-03; 05000446/2013005-03, Notice of Enforcement Discretion 13-4-004 for a Loss of Both Required Offsite Power Circuits The inspectors performed a review of the circumstances associated with the granting of Notice of Enforcement Discretion 13-4-004 for Luminant Generation Company, LLC telephonically on December 5, 201 The notice of enforcement discretion granted an additional 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to restore compliance with Technical Specification 3.8.1Sources - Operatin The inspectors verified the likely cause and compensatory measures, and verified the notice of enforcement discretion request wa The licensee documented the issue in Licensee Event Report 05000445/2013-003-0 The inspectors closed the licensee event report in Section 4OA3.1 of this repor No findings were identified. .2 Impact of Financial Conditions on Continued Safe Performance In that the licensees parent company, Energy Future Holdings, was under bankruptcy protection/reorganization during the inspection period, NRC Region IV conducted special reviews of processes at Comanche Peak. The inspectors evaluated several aspects of the licensees operations to determine whether the financial condition of the station impacted plant safety. The factors reviewed included: (1) impact on staffing, (2) corrective maintenance backlog, (3) changes to the planned maintenance schedule, (4) corrective action program implementation, and (5) reduction in outage scope, including risk-significant modification In particular, the inspectors verified that licensee personnel continued to identify problems at an appropriate threshold and enter these problems into the corrective action program for resolution. The inspectors also verified that the licensee continued to develop and implement corrective actions commensurate with the significance of the problems identifie The special review of processes at Comanche Peak included continuous reviews by the Resident Inspectors, as well as the specialist-led baseline inspections completed during the inspection period which are documented previously in this repor failure to correct fire protection violations in a timely mannerManager, Regulatory Affairs, No proprietary information was in the repor The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non-cited violation .1 CR-2014-000960 .2 License Condition 2.G for Unit 2 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 SectioComanche Peak Nuclear Power Plant Fire Protection Repor Section II of the Fire Protection Report is the Fire Hazards Analysis Repor Section otection of the Fire Hazards Analysis Report specifies the allowable methods of ensuring that one of the redundant sets of systems necessary to achieve and maintain hot standby conditions is free of fire damag Section 4.5.2.c, states, in part, that, enclosures of cables (if not one hour fire rated cables) and equipment and associated nonsafety circuits of components of redundant sets of systems in a fire barrier have a 1-hour ratin In addition, fire detectors and automatic fire suppressions systems adequate for hazards in the fire area are installe CPNPP Fire Protection Program the Fire Protection Report specifies that the separation method per Section 4.5 of the Fire Hazards Analysis Report utilized in fire area SB is method Contrary to the above, prior to June 19, 2014, the licensee failed to ensure that an associated nonsafety circuit in fire area SB was enclosed in a fire barrier having a 1-hour ratin Specifically, the licensee failed to ensure that cable EO223531, a cable in the control circuit for the train A containment recirculation sump isolation valve, was enclosed in a fire barrier having a 1-hour fire ratin The licensee documented this issue in Condition Report CR-2014-00472 This violation was determined to be of very low safety significance (Green) based on Inspection Manual Chapter 0609, Appendix F, Question 1.4. .3 Hot Short Prevention Cable Shield Conductor License Condition 2.G for Unit 2 requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 2 Section 13.3B.5, of the Final Safety Analysis Report describes the fire protection quality assurance progra The fire protection quality assurance program states that measures shall be established to ensure that conditions adverse to fire protection such as failures, malfunctions, deficiencies or deviations, defective components, uncontrolled combustible material and nonconformances are promptly identified, reported, and corrected. Contrary to the above, prior to June 19, 2014, the licensee failed to ensure that a condition adverse to fire protection was promptly correcte Specifically, the licensee failed to properly connect the shield conductor for the hot short prevention cable that was installed as part of a modification to resolve a fire protection nonconformanc The licensee identified that the modification design connected the hot short prevention cable shield conductor to the plant ground instead of the dc negative potentia The licensee identified this issue in Condition Report CR-2014-00519 The senior reactor analyst determined this violation was of very low safety significance (Green) based on a bounding Phase 3 evaluatio A1-1 R. Flores, Senior Vice President and Chief Nuclear Officer S. Bradley, Manager, Radiation Protection D. Goodwin, Director, Work Management T. Hope, Manager, Regulatory Affairs F. Madden, Director, External Affairs B. Mays, Assistant Chief Nuclear Officer T. McCool, Vice President, Engineering and Support D. McGaughey, Director, Performance Improvement B. Moore, Director, Nuclear Training K. Nickerson, Director, Engineering Support B. Patrick, Director, Maintenance K. Peters, Site Vice President B. Reppa, Director, Site Engineering , Plant Manager M. Smith, Director, Nuclear Operations - - - - - - Failure to Follow 10 CFR 50.59 for a Change to the Spent Fuel Pool Configuration (Section 1R18) - - - Failure to Adequately Brief Workers on Radiological Conditions Prior to Entry into High Radiation Areas (Section 2RS1) - - - - - -

A1-2 05000445/2012004-03 05000446/2012004-03 Potential Failure to Follow 10 CFR 50.59 for a Change to the Spent Fuel Pool Configuration (Section 1R18) - - -- 05000445/2013-003-00 Safeguards Electrical Power -- -- 2013-000093 2013-000359 2013-004894 2013-007282 2013-012642 2014-001856 2014-004668 2014-004911 - - - - - -

A1-3 --- --- - - - - - - - - - - - - - - - - - - STA-737 --- --- --- -- - ---- ---- -- --- - -- --

A1-4 -- -- -- Miscellaneous Documents -- -- -- - - - - - - -- --- ---

A1-5 Miscellaneous Documents - --- -- - --- - -- -- - - - -

A1-6 Section 1R15: Operability Determinations and Functionality Assessments Calculations --- - - - - - Procedures - -- Procedures NUC-211 Surveillance of Region II Storage Limitations 1 and 2 ODA-308 LCO Tracking Program 15 STA-422 Processing Condition Reports 28 STA-707 10 CFR 50.59 Evaluations 1 Miscellaneous Documents Comanche Peak FSAR Section NRC Administrative Letter 98- NRC Information Notice 2011- TXU Letter TXX--022, Spent Fuel NRC Letter Dated A1-7 - 2014-002383 2014-002913 2014-004127 2014-005042 2014-007171 Procedures - - - - 4494551 4508568 4636907 4756056 4794262 4806680 4810738 4813424 Calculations -- 2013-009406 2013-010799 2014-003437 2014-003707 2014-004111 2014-004133 2014-004290 2014-004299 2014-004482 2014-004501 2014-004604 2014-004626 2014-004721 2014-005063 2014-005571 2014-005723 2014-005964 2014-007904 - - Procedures -

A1-8 Procedures -- - -- -- -- - - Section 1R22: Surveillance Testing Calculation --- - - - - 2014-007171 - - Procedures - - - - - - -

A1-9 Procedures - - - ---- - - - - - Procedures - Emergency Response Organization Staffing and Augmentation System Miscellaneous Documents -

A1-10 Miscellaneous Documents - Maintenance of Emergency Preparedness - - - -000673 -002183 -009431 -009443 -010232 -011053 -012085 -013480 -000089 -000092 -000121 -000123 -000132 -000163 -000169 -000171 -000174 -000279 -000536 -000965 -001504 -001629 -002773 -002865 -002935 -003075 -003696 -004580 -004714 -005791 -006442 -006709 -006714 -007278 -007948 -008128 -008988 -009019 -009445 -009642 -010505 -010960 -011082 -011115 -011854 -011914 -011985 -011986 -012654 -000025 -001263 -001535 -001563 -002814 -003060 -004375 -004817 -004871 -004988 -005182 -005280 -005777 -005820 Procedures - - - - - - -

A1-11 Miscellaneous Documents -- - - - - CR-2013-008741 Mid-Cycle Strategic Self-Assessment CR-2013-005022 Tactical Self-Assessment of Radiological Surveys and Postings CR-2013-005021 Tactical Self-Assessment on Radwaste Transportation and Environmental Monitoring CR-2013-001057 Tactical Self-Assessment of the Radiation Protection Instrumentation Program 2013-004154 2013-004182 2013-004313 - 2013-004569 2013-004721 - 2013-007139 2013-007162 - 2014-001123 2014-001179 - - 2014-002118 2014-003464 - - 2014-004215 2014-004284 2014-004331 A1-12 Procedures - - RPI-602 - - STA-660 Control of High Radiation Areas 13-04-0038 U--154A-D April 01, 2013 13-04-0981 U--154C April 14, 2013 -- - -- - -- -- -- - 2014-2600 Refueling Activities April 10, 2014 2014-2400 Steam Generator Primary work April 10, 2014 2014-2300 Steam Generator Secondary work April 10, 2014 2014-2215 Scaffolding Activities April 10, 2014 2014-2217 Insulation work in the RCA April 10, 2014 2014-2601 Refuel Support April 10, 2014 2014-2214 Reactor Annulus/ Seal Table and Incore Rooms April 10, 2014 2014-2208 Letdown Orifice Room 2-155A April 10, 2014 2014-2101 Operations April 10, 2014 2014-2227 Valve Work 2-077A&B / 2-080 April 10, 2014 A1-13 2013-004154 2013-004182 2013-004313 - 2013-004569 2013-004721 2013-004768 - 2013-007139 2013-007162 - - 2014-001123 2014-001179 - - 2014-002118 2014-003464 - - Procedures - - - STA-651 ALARA Program STA-657 ALARA Job Planning/Debriefing STA-660 Control of High Radiation Areas -- --- -- -- -- - -- - -- -- -- - -

A1-14 - - - - - - - - - - -- -- Procedure - 2013-001326 2013-001817 2013-001959 2013-004806 2013-006120 2013-007128 2013-010464 2013-010597 2013-012067 2014-000708 2014-001506 2014-003152 2014-003472 2014-003727 2014-004065 2014-004178 2014-004249 2014-004437 2014-004579 2014-004604 2014-004665 2014-004721 2014-004723 2014-004778 2014-005107 2014-005198 2014-005360 2014-006092 2014-006298 2014-006528 A1-15 2014-006530 2014-006600 2014-007319 Design Basis Documents 2323-ES-100 Specification Electrical Installation 108 DBD-EE-052 Cable Philosophy and Sizing Criteria 42 A1-0413, Sheet 1 Room/Area Designations Unit 1 and 2 Plans at -- CP-4 A1-0413, Sheet 2 Room/Area Designations Unit 1 and 2 Plans at -- CP-8 A1-0413, Sheet 4 Room/Area Designations Unit 1 and 2 Plans at -- CP-3 A1-0509 Primary Plant Electrical Control Building Floor Plan - CP-11 E1-0713-11 Auxiliary and Electrical Control Buildings Cable Tray -- 7 E1-0713-12 Auxiliary and Electrical Control Buildings Cable Tray -- 6 E2-0062, Sheet 22 Motor Operated Valve 2-8811A Sump to 1 Residual Heat Removal Pump CP-7 E2-0062, Sheet 23 Motor Operated Valve 2-8811B Sump to Number 2 Residual Heat Removal Pump CP-10 E2-0062, Sheet 24 Motor Operated Valve 2-8812A Refueling Water Storage Tank to Residual Heat Removal Pump 1 Isolation CP-6 E2-0062, Sheet 25 Motor Operated Valve 2-8812B Refueling Water Storage Tank to Residual Heat Removal Pump 2 Isolation CP-7 E2-0601-11 - 5 E2-0602-11 Safeguard Building Cable Tray Segments Plan Elevation - 3 E2-0602-12 Safeguard Building Cable Tray Segments Plan Elevation - 3 E2-0603-11 - 3 E2-0700-12 - 5 A1-16 E2-0701-12 - 5 E2-0702-11 - 2 E2-0703-12 - 6 E2-0706-11 Auxiliary Building Cable Tray Segments - 3 E2-0712-11 Cable Room and Auxiliary Building Cable Tray Segments -- 6 E2-0712-12 Cable Room and Auxiliary Building Cable Tray Segments -- 6 E2-0716-11 Electrical Equipment Area Cable Tray Segments Plan - 7 E2-0718-11 Electrical Equipment Area Cable Tray Segments Plan - 6 M1-0225, Sheet 3 Flow Diagram Safeguard Building Unit 1 Fire Protection CP-15 M2-0225, Sheet 3 Flow Diagram Safeguard Building Unit 2 Fire Protection CP-10 Procedures - - - -- -- - - ---

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A2-1 REQUEST FOR INFORMATION OCCUPATIONAL RADIATION SAFETY INSPECTION The following items are requested to support the occupational radiation safety inspection conducted during the week of April 7-11, 201 The areas of inspection are listed below in the attachmen Please provide the requested information on or before March 21, 201 Please submit this information using the same lettering system as belo For example, all contacts and phone - - If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at least 30 days later than the onsite inspection dates so the inspectors will have access to the information while writing the repor In addition to the corrective action document lists provided for each inspection procedure listed below, please provide updated lists of corrective action documents at the entrance meetin The dates for these lists should range from the end dates of the original lists to the day of the entrance meetin Since more than one inspection procedure is to be conducted and the information requests appear to be redundant, there is no need to provide duplicate copie Enter a note explaining in which file the information can be foun If you have any questions or comments, please contact the lead inspector, Louis Carson, at (817) 200-1221 or Louis.Carson@nrc.go The other inspector will be Natasha Green She can be reached at (817) 200-1154 or Natasha.Greene@nrc.go . The date of last inspection was April 4, 201 a. List of contacts and telephone numbers for the radiation protection organization staff and technicians b. Applicable organization charts c. Audits, self-assessments, and licensee event reports written since date of last inspection, related to this inspection area d. Procedure indexes for the radiation protection procedures e. Specific procedures related to the following areas noted belo Additional procedures may be requested by number after the inspector reviews the procedure indexe Radiation protection program description Radiation protection conduct of operations Personnel dosimetry program Posting of radiological areas High radiation area controls Radiological controlled area access controls and radworker instructions Conduct of radiological surveys Radioactive source inventory and control A2-2 Declared pregnant worker program f. List of corrective action documents (including corporate and subtiered systems) since date of last inspection Initiated by the radiation protection organization Assigned to the radiation protection organization Potentially related to a performance indicator event NOTE: The lists should indicate the significance level of each issue and the search criteria can perform word searche If not covered above, a summary of corrective action documents since date of last inspection involving unmonitored releases, unplanned releases, or releases in which any dose limit or administrative dose limit was exceeded for Public Radiation Safety Performance Indicator verification in accordance with Inspection Procedure 7115 g. List of radiologically significant work activities scheduled to be conducted during the inspection period. Since the inspection is scheduled during an outage, also include a list of work activities greater than 1 rem with the dose estimate for the work activit h. List of active radiation work permits i. Radioactive source inventory list Occupational ALARA Planning and Controls (71124.02) The date of last inspection was December 17, 201 a. List of contacts (with official title) and telephone numbers for ALARA program personnel b. Applicable organization charts c. Copies of audits, self-assessments, and licensee event reports, written since date of last inspection, focusing on ALARA d. Procedure index for ALARA program e. Please provide specific procedures related to the following areas noted belo Additional specific procedures may be requested by number after the inspector reviews the procedure indexe ALARA program ALARA committee Radiation work permit preparation f. A summary list of corrective action documents (including corporate and subtiered systems) written since date of last inspection, related to the ALARA progra In addition to ALARA, the summary should also address radiation work permit violations, electronic dosimeter alarms, and radiation work permit dose estimate A2-3 NOTE: The lists should indicate the significance level of each issue and the search List of work activities greater than 1 rem, since date of last inspectio Include original dose estimate and actual dos h. Site dose totals and 3-year rolling averages for the past 3 years (based on dose of record) i. Outline of source term reduction strategy