CP-202000263, Reactor Vessel Head, Bottom Mounted Instrumentation, and Hot Leg Ultrasonic Testing Examinations Relief Request

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Reactor Vessel Head, Bottom Mounted Instrumentation, and Hot Leg Ultrasonic Testing Examinations Relief Request
ML20100G562
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 04/09/2020
From: Sewell S
Luminant, TXU Energy, Vistra Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-202000263, TXX-20028
Download: ML20100G562 (25)
v • d • e


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to TXX-20028 Page 1 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

--Hardship without a compensating increase in quality and safety--

1. ASME Code Component(s) Affected Components: Reactor Vessel (RV) Bottom Mounted Instrumentation (BMI) Nozzle Penetration Code Class: Class 1 Code Case: N-722-1 (Ref. 2)

Examination Category: B-P [Class 1 Pressurized Water Reactor (PWR)

Components Containing Alloy 600/82/182]

Item Number: Code Case N-722-1 Item Number 15.80

Description:

58 BMI Nozzles welded to the inside surface of the RV with partial penetration J-groove welds

2. Applicable Code Edition and Addenda ASME Section XI, 2007 Edition through the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components. (Ref. 1)
3. Applicable Requirement The Code of Federal Regulations, Title 10, Part 50, Section 55a (g)(6)(ii)(E)(1) states:

All licensees of pressurized water reactors must augment their inservice inspection program by implementing ASME Code Case N-722-1, subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through (4) of this section. The inspection requirements of ASME Code Case N-722-1 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay or stress improvement.

ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials. (Ref. 2).

4. Reason for Request The U.S. Federal Government made a COVID-19 declaration of emergency pursuant to the Stafford Act on March 13, 2020. The U.S. Center for Disease Control (CDC) has determined that COVID-19 poses a serious public health risk. The CDC has identified many U.S. States reporting community spread of COVID-19 with the anticipation of conditions worsening over the coming weeks.

In response to the COVID-19 Pandemic, Vistra Operations Company LLC (Vistra OpCo) has established the following guidelines and restrictions at Comanche Peak Nuclear Power Plant (CPNPP):

1 to TXX-20028 Page 2 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4

1. Employees who do not have a critical need to be at CPNPP facilities must work remotely.
2. Employees who must work from a CPNPP facility are to practice strict social distancing.
3. Outage scope shall be reduced to limit the number of supporting contract personnel.

These guidelines and restrictions were established to minimize the potential of inadvertently spreading the COVID-19 virus to CPNPP personnel and the surrounding community from outside contractors who perform examinations and tests at CPNPP. Also, due to recent travel restrictions and quarantine requirements imposed by both the U.S.

Government and the State of Texas, the availability of outside contractors to provide the NDE services may be limited. Further complicating this situation is the Vistra OpCo reliance on outside NDE resources to supplement the small CPNPP NDE Staff.

Many of the planned spring 2020 refueling outage (2RF18) activities have been postponed until future outages based on these guidelines and restrictions; however, the Reactor Vessel Bottom Mounted Instrumentation (BMI) Nozzle Penetration examination which is required every other refueling outage by Code Case N-722-1 as mandated by the Code of Federal Regulations in Title 10, Part 50, Section 55a, (10CFR50.55a) cannot be postponed without prior NRC approval.

5. Proposed Alternative and Basis for Use Proposed Alternative Delay required rescheduling BMI visual examinations (VE) performed by contract personnel from 2RF18 to the fall 2021 refueling outage (2RF19). This interval extension is being requested in accordance with 10CFR50.55a(z)(2) as a hardship without a compensating increase in quality and safety.

In lieu of performing BMI VEs, CPNPP personnel will perform a boric acid inspection of the bottom head from the peripheral of the reactor mirror insulation package by removal of select insulation panels to gain access which will allow inspection for any gross active boric acid leakage.

Basis for Use Code Case N-722-1 requires a VE of the CPNPP Unit 2 reactor pressure vessel BMI nozzle penetrations every other outage. The BMI visual inspections are performed in order to detect leakage as a result of Primary Water Stress Corrosion Cracking (PWSCC) operating experience (OE). As identified in MRP-206 (Ref. 3) the inspections are performed to address safety concerns attributed to nozzle ejection as a result of a large circumferential crack below the bottom of the J-groove weld and the structural loss of the ferritic head material due to boric acid wastage.

In addition, leakage from the Lower Head would be detected in many ways due to CPNPPs extremely low tolerance for Class 1 operational leakage. Some of these detection methods are discussed below:

2 to TXX-20028 Page 3 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4 Operations Monitoring Technical Specifications (TS) 3.4.13, RCS Operational LEAKAGE, for each Unit, limits system operation in the presence of leakage from Reactor Coolant System (RCS) components to leakage amounts that do not compromise safety. Surveillance Requirement (SR) 3.4.13.1 requires the performance of RCS water inventory balance to verify RCS leakage is within limits to ensure that the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained.

Per LCO 3.4.13 RCS operational LEAKAGE shall be limited to no pressure boundary LEAKAGE and 1 gpm unidentified LEAKAGE.

Procedure OPT-303, Reactor Coolant System Water Inventory, Attachment 10.2, RCS Live Leakrate Action Levels and Response Guidelines, provides the steps that would be taken to satisfy TS SR 3.4.13.1 by performance of an RCS water inventory balance. A summary of the steps is seen below in A through E:

A. Action Levels based on Absolute Value of Unidentified RCS Inventory Balance (From Surveillance Data)

Entrance Criteria ACTION LEVEL One seven (7) day rolling average of LEVEL 1 Unidentified RCS Inventory Balance values > 0.1 gpm Two consecutive Unidentified RCS LEVEL 2 Inventory values > 0.15 gpm One Unidentified RCS Inventory Balance LEVEL 3 value > 0.3 gpm B. Action Levels based on Deviation from the Baseline Mean:

Entrance Criteria ACTION LEVEL Nine (9) consecutive Unidentified RCS LEVEL 1 Inventory Balance values > baseline mean

[]

Two (2) of three (3) consecutive LEVEL 2 Unidentified RCS Inventory Balance Values > [ + 2], where is the baseline standard deviation One (1) Unidentified RCS Inventory LEVEL 3 Balance > [ + 3]

C. Exit Criteria following Action Level Entry:

Exit Criteria

  • Location of RCS leak has been identified AND Leak terminated (isolated or stopped) - confirmed by RCS leak rate and VCT level trends OR 3

to TXX-20028 Page 4 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4

  • A minimum of 2 daily leak rates with indicated leakage less than the applicable Action Level AND Ops Department and Engineering staff recommend exiting the applicable Action Level D. For any ACTION LEVEL response:
1) Run confirmatory leak rate calculation
2) Confirm indication
3) Evaluate trend of affected parameters
4) Check for abnormal trend of other leakage indicators E. IF confirmed:
1) Increase monitoring of leakage indicators
2) Initiate a Condition Report to document investigation and results in an evaluation
3) Commence a leak investigation
  • IF ACTION LEVEL 1, search for sources as resources allow
  • IF ACTION LEVEL 2, search for sources, continue 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day, 7 days/week until cause determined
  • IF ACTION LEVEL 3, search for sources, continue 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day, 7 days/week until cause determined. At 0.50 gpm sustained leakage, the Shift Manager should consult with the Duty Manager to consider an orderly shutdown to Mode 3.
  • Initiate log book detailing all systems and rooms inspected for leakage with findings. All rooms should be inspected with a thermographic camera (Fire Brigade or Engineering camera).
  • Track all inspections on copies of system drawings, denoting piping and valves inspected.
  • Review recent plant evolutions to determine any suspect source(s).
  • Check any components or flow paths recently changed or placed in service, shutdown, vented, drained, filled, etc.
  • Check any maintenance activity that may have resulted in increasing leakage.
  • Check RCS and seal injection filter seals, vents, and drains for leakage (most common source of leakage).
  • Check any filters recently alternated or changed for leakage from their vents or drains, inspect filter housings for gasket leaks, check seal injection filters and reactor coolant filters for signs of leakage.
  • Scrutinize sump pump run times, sample trends, and rad monitor readings.
  • Divide the primary plant into several large groups for investigation. (e.g.

Leakage to atmosphere inside containment, S/G primary to secondary leakage, leakage to atmosphere outside containment, inter system leakage)

  • Scrutinize tank levels and trends (e.g., RHUT, RCDT, PRT and SRST),

utilizing temporary indication if necessary.

  • Initiate outside Containment walk-downs of various portions of potentially affected systems.
  • Notify System Engineer to obtain input/assistance.

4 to TXX-20028 Page 5 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4

4) IF increased leak rate is indicated inside Containment, THEN:
  • Begin planning for Containment entry while carrying out other actions; obtain proper approval for Containment entry.
  • Obtain a Containment Sump sample (during pump out) and analyze for activity, a larger than expected boric acid concentration and other unexpected chemicals.
  • Evaluate other systems for indications of leakage.
  • Obtain a Containment atmosphere sample for indications of RCS leakage.
  • Perform a Containment entry to search for signs of leakage.
  • If necessary, utilize robot to perform loop room inspections.
5) Identify the source of the increase in leakage.
6) Quantify the leakage.
7) Initiate plan to correct the leak.
8) Monitor Containment airborne radiation levels as well as area radiation monitors and sample Containment atmosphere for indications of RCS leakage.
9) Monitor other Containment parameters (temperature, pressure, humidity, etc.).
10) If the leak source is found and isolated or stopped, re-perform RCS leak rate calculation.

In addition to surveillance required monitoring, Operations continually monitors the RCS leak rate through control board metering and trend graphing capability. Instrumentation such as pressurizer level, sump levels, containment pressure and humidity instruments, and many others are monitored on a continuous basis and are logged and reviewed each shift.

Administrative controls Administrative procedures require monitoring of RCS leakage under the Boric Acid Corrosion Control Program (BACCP) on a per cycle basis. The program also addresses abnormal trends in RCS primary system leakage indicators, which may provide indication of leaks much smaller than TS and RCS leakage levels. CPNPP monitors the following containment building/system parameters during the operating cycle to determine any potential leakage from borated or radioactive systems containing boric acid.

Unidentified Reactor Coolant System Leakage:

This parameter monitors unidentified (RCS leakage both inside and outside of containment. The purpose is to review the data graph for trends in leakage and assess if any increases are indication of borated system leakage in Unit 2 containment.

Containment Air Particulate and Gaseous Concentration:

Radiochemistry analysis are performed of the containment atmosphere for each vent process. The vent process is performed approximately every three days and can identify changes in containment activity. The isotope for detecting increase leakage is noble gas Argon-41. Argon-41 is a short-lived isotope that can be readily monitored to detect small increases or upward trends in leakage. From this parameter, an increase in borated system leakage in the Unit 2 containment can be detected.

Containment Air Cooler Condensate Flow and Containment Sump 5

to TXX-20028 Page 6 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4 Level/Pump Rates:

The containment air cooler condensate flow is not separately measured, but flows to the containment sumps. Therefore, the reported sump flow rates include the containment air cooler condensate flows. Review of monthly Unit 2 sump pump volumes for this operating cycle are performed to determine if any upward trend in daily pump volume rates are observed. An increased pump rate could indicate leakage from borated systems in Unit 2 containment.

Containment Humidity/Dew Point Temperatures:

The Unit 2 containment has five area dew point temperature sensors at containment elevations 832, 860, and 905, as well as the Control Rod Degradation Mechanism shroud and the reactor coolant pipe penetrations. Weekly dew point temperature measurements during plant operation are performed to monitor changes in temperature that could be indicative of leakage from borated systems in containment. Minor changes in temperature, both up and down, are gradual and are considered to be normal due to seasonal variations.

Continued monitoring of these parameters will provide early indication of any abnormal unidentified leakage during the upcoming cycle following 2RF18.

Previous CPNPP Unit 2 Internal Operating Experience (OE)

The last BMI inspection for Unit 2 with vendor assistance was performed in the Spring of 2017 during refueling outage 2RF16 with no relevant indications. During every other outage when the BMI visual examinations are not performed, (i.e., 2RF17) CPNPP personnel perform a boric acid inspection of the bottom head area. To date there have been no observed BMI leaks in Unit 2.

Other Considerations Visual examination of the exterior surface of the BMI penetrations will be performed in accordance with the BACCP during 2RF18.

The ISI Pressure Testing Program will implement the Class 1 System Leakage Test required by ASME Section XI Examination Category B-P, Item Number B15.10 at normal operating pressure and temperature during heatup from 2RF18 which includes the BMI area; VT-2 Visual Examination will be performed during the test.

Any gross leakage (i.e. water, boric acid, insulation deformation) or structural deformities would be identified during the above examinations/activities and captured in the Corrective Action Program.

In summary, extending the inspection frequency for direct visual examination of reactor bottom head and associated penetrations to the next refueling outage scheduled for 2RF19 would not adversely impact the function of the reactor bottom head or penetrations pressure boundary integrity or result in a reduction in plant safety. In the current pandemic environment, performing the required tests would result in an increased risk of virus exposure to plant personnel and a reduction in occupational health and safety without a compensating benefit. Therefore, this one-time request for relief meets the criteria in 10 CFR 50.55a(z)(2) for proposing 6

to TXX-20028 Page 7 of 7 Comanche Peak Unit 2 RELIEF REQUEST NUMBER 2A3-4 inspection alternatives on the basis that compliance results in hardship or unusual difficulty without a compensating increase in level of quality or safety during the current COVID-19 pandemic.

6. Duration of Proposed Alternative The duration of the request is for Unit 2 operating cycle 19, until refueling outage 2RF19 in the fall of 2021.
7. Precedents ML20088A533 supplemented by ML20090L944 - Relief Request 65 Unit 2, COVID-19, Request for Relief from Bottom Mounted Instrumentation Nozzles and a Pressurizer Nozzle to Surge Line Weld Overlay Examination - Palo Verde Nuclear Generating Station (PVNGS)
8. References
1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section Xl, 2007 Edition through 2008 Addenda.
2. ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 MaterialsSection XI, Division 1.
3. MRP206, Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants.
4. 10 CFR 50.55a, Codes and Standards, January 2020.

7 to TXX-20028 Page 1 of 7

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