IR 05000250/2019011

Design Basis Assurance Inspection (Programs) Inspection Report 05000250/2019011 and 05000251/2019011

ML19274C217
Person / Time
Site:	Turkey Point
Issue date:	09/30/2019
From:	James Baptist NRC/RGN-II/DRS/EB1
To:	Moul D Florida Power & Light Co
References
IR 2019011
Download: ML19274C217 (16)

Text

ber 30, 2019

SUBJECT:

TURKEY POINT NUCLEAR GENERATING STATION - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000250/2019011 AND 05000251/2019011

Dear Mr. Moul:

On August 16, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Turkey Point Nuclear Generating Station and discussed the results of this inspection with Mr.

Stamp and other members of your staff. The results of this inspection are documented in the enclosed report.

Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance or severity of the violations documented in this inspection report, 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 II; the Director, Office of Enforcement; and the NRC Resident Inspector at Turkey Point Nuclear Generating Station.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 05000250 and 05000251 License Nos. DPR-31 and DPR-41

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000250 and 05000251 License Numbers: DPR-31 and DPR-41 Report Numbers: 05000250/2019011 and 05000251/2019011 Enterprise Identifier: I-2019-011-0003 Licensee: Florida Power & Light Co.

Facility: Turkey Point Nuclear Generating Station Location: Homestead, FL Inspection Dates: July 29, 2019 to August 16, 2019 Inspectors: J. Corujo-Sandin, Reactor Inspector T. Fanelli, Senior Reactor Inspector T. Su, Reactor Inspector Approved By: James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at Turkey Point Nuclear Generating Station in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations Failure to Qualify D.C. Combination Starter Panel 4N1405-A for Minimum Operating Time Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None 71111.21N Systems NCV 05000250, 251/2019011-02 Closed The NRC identified a Green Non-Cited Violation (NCV) of 10 CFR 50.49.(e)(1), "Temperature and Pressure," for the licensee's failure to qualify direct current (D.C) combination starter panel 4N1405-A for a period of at least one hour in excess of the time assumed in the accident analysis.

Two Examples of Failure to Verify Design Inputs to Qualification Criteria Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None 71111.21N Systems NCV 05000250, 251/2019011-03 Closed The NRC identified a Green Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion III,

"Design Control," for the licensee's failure to verify the adequacy of qualification testing and installation requirements in accordance with Quality Instruction (QI) 1.7, "Nuclear Engineering Design Input/Verification."

Additional Tracking Items Type Issue Number Title Report Section Status URI 05000250, Potential Harsh 71111.21N Open 251/2019011-01 Environments from High-Energy Line Breaks

INSPECTION SCOPE

===Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

===71111.21N - Design Bases Assurance Inspection (Programs)

The inspectors evaluated the licensees environmental qualification program implementation through the sampling of the following components:

Select Sample Components to Review - Risk Significant/Low Design (Inside/Outside Containment) (IP Section 02.01) ===

(1) 3V30C, JOY Manufacturing Company Emergency Containment Cooling Fan

(2) POV-2604, Main Steam Isolation Valve (MSIV) Opening Solenoid Channel B AVCO

(3) FY-3-1457B-6, Fisher Current to Pressure (I/P) Converter CV-2832, 303

(4) MOV-4-1405, Limitorque AFW Steam Supply

(5) SV-3-6319B, Target Rock Pressurizer Vent Discharge Solenoid Valve

(6) SPL SLEEVE-NPK, Raychemcor SPL Sleeve-NPK Nuclear Plant Splice Kit Select Sample Components to Review - Primary Containment (Inside Containment) (IP Section 02.01) (1 Sample)===

(1) MOV-4-751, Limitorque Normal RHR Inlet from Reactor Coolant System

INSPECTION RESULTS

Failure to Qualify D.C. Combination Starter Panel 4N1405-A for Minimum Operating Time Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None 71111.21N Systems NCV 05000250, 251/2019011-02 Closed The NRC identified a Green NCV of 10 CFR 50.49.(e)(1), "Temperature and Pressure," for the licensee's failure to qualify direct current (D.C) combination starter panel 4N1405-A for a period of at least one hour in excess of the time assumed in the accident analysis.

Description:

The team identified that the licensee failed to qualify the D.C. combination starter panel, 4N1405, for MOV-4-1405 to meet the minimum required operating time as specified in Regulatory Guide (RG) 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants. The licensee committed to following the guidance in RG 1.89 by reference in their updated final safety analysis report (UFSAR), Appendix 8A, Environmental Qualification.

The Unit 4 D.C. starter panel 4N1405 controls the MOV 4-1405 which is the steam stop valve to the turbine-driven auxiliary feedwater pump. The starter panel environmental qualification report indicated a qualification time of 33 minutes. This time was consistent with the licensees high energy line break (HELB) blowdown time, not the required valve operating time including post-accident operating time (PAOT) for various accidents. The UFSAR, Section 14.2.7, Feedwater System Pipe Break, specified, in part, that it is concluded that the available auxiliary feedwater capacity is adequate for long-term decay heat removal, and the applicable acceptance criteria for the feedwater system pipe break analysis are met.

Final reactor coolant system cooldown begins (time of event turnaround) at one hour into the event (3600 seconds). For this accident sequence, the auxiliary feedwater pumps must operate for longer than 33 minutes. The pumps must continuously remove heat from the steam generators. The RG 1.89 Reg Position C.4, stated, in part, that the equipment should remain functional in the accident environment for a period of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the time assumed in the accident analysis. This time included HELB blowdown time and PAOT, including an additional one hour in accordance with the site design basis above.

Corrective Actions: The licensee entered these deficiencies into the corrective action program to restore compliance.

Corrective Action References: AR 02324645

Performance Assessment:

Performance Deficiency: The licensee's failure to qualify D.C. combination starter panel 4N1405-A for a duration with margin as specified by RG 1.89 was a performance deficiency.

Specifically, RG 1.89 stated, in part, that the equipment should remain functional in the accident environment for a period of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the time assumed in the accident analysis. This included PAOT.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to qualify 10 CFR 50.49 equipment to the most severe environment for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the operating time assumed in the accident analysis fails to ensure the reliability and capability of those components to perform their safety function when called upon during and following a design basis accident.

Significance: The inspectors assessed the significance of the finding using manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for mitigating systems, and IMC 609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency of a mitigating SSC and the SSC maintained its functionality.

Cross-Cutting Aspect: Not Present Performance. No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: 10 CFR 50.49(e)(1) requires, in part, that

(e) The electric equipment qualification program must include and be based on the following:

(1) Temperature and pressure. The time-dependent temperature and pressure at the location of the electric equipment important to safety must be established for the most severe design basis accident during or following which this equipment is required to remain functional.

Contrary to above, since February 25, 2013, the licensee failed to base qualification on the time-dependent temperature and pressure at the location of the electric equipment important to safety during and following the most severe design basis accident which this equipment is required to remain functional. Specifically, combination starter panel 4N1405-A was not qualified for the post accident operating time.

Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Two Examples of Failure to Verify Design Inputs to Qualification Criteria Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None 71111.21N Systems NCV 05000250, 251/2019011-03 Closed The NRC identified a Green NCV of 10 CFR 50 Appendix B, Criterion III, "Design Control," for the licensee's failure to verify the adequacy of qualification testing and installation requirements in accordance with Quality Instruction (QI) 1.7, "Nuclear Engineering Design Input/Verification."

Description:

Licensee instruction QI 1.7, Section 5.3, "Design Verification," required, that design verification is the process whereby a competent individual, who has remained independent of the design process, reviews the design inputs, design and safety analyses, and design output to verify design adequacy. This independent review is provided to minimize the likelihood of design errors in items that are important to nuclear safety. In the following examples the adequacy of the environmental qualification and installation was not verified.

Example one: The licensee committed to RG 1.89 and Institute of Electrical and Electronics Engineers (IEEE) 323-1974 by reference in their updated final safety analysis report (UFSAR), Appendix 8A, Environmental Qualification. The site failed to verify that Fisher current to pressure (I/P) transducers were qualified in accordance with the site EQ requirements in RG 1.89 and IEEE 323-1974. The qualification program and testing performed on Fisher I/P transducers did not meet the requirements in IEEE 323-1974 or RG 1.89 for replacement equipment. For instance, the recorded accident environment did not include a temperature and pressure transient. Further, the recorded cool down rate was uniform over ten hours rather than the abrupt temperature changes expected during accidents. Also, the transducers were not operated or cycled during a transient or throughout the testing as expected for a modulating valve. When asked, the licensee performed an analysis to demonstrate how the Fisher qualification met the IEEE and RG. The qualification was determined to be nonconforming. The testing was not adequate to support qualification of the transducers.

Example two: The site failed to verify that environmentally qualified components were installed in the as-qualified configurations. Weep holes are required for all boxes containing terminal blocks, relays, devices with exposed terminals, visible contacts, etc. as per licensees drawing 5610-E-302, Rev. 22. The inspectors noted at least three examples where control and terminal boxes were installed without the requisite weep holes.

1. On Unit 4, the site failed to install drain holes for 1N1405 Starter panel as required per above referenced drawing.

2. On Unit 4, the site failed to install drain holes for the junction box associated with Automatic Valve Company (AVCO) solenoid valve, SV-4-2604B as required per the above referenced drawing.

3. On Unit 3, the site failed to install drain holes for the junction box associated with the Fisher I/P converter, FY-3-14578-6, as is required per the above reference drawing.

Some of the qualification records indicated that boxes with blocked weep holes failed the LOCA and HELB tests.

Corrective Actions: The licensee entered these deficiencies into the corrective action program to restore compliance.

Corrective Action References: AR 02324142 and AR 02324348

Performance Assessment:

Performance Deficiency: The failure to verify the adequacy of qualification testing and installation requirements to maintain qualification in accordance with QI 1.7 Section 5.3 was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify component qualification during testing and installation affected the component reliability when it is required to operate.

Significance: The inspectors assessed the significance of the finding using manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for mitigating systems, and IMC 609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency of a mitigating SSC and the SSC maintained its functionality.

Cross-Cutting Aspect: Not Present Performance. No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: 10 CFR Part 50 Appendix B, Criterion III, required, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, since 1984, the licensee failed to provide design control measures that verified or checked the adequacy of design, by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee failed to verify the adequacy of environmental qualification testing and equipment installation methods to maintain environmental qualification, which resulted in unqualified I/P transducers and terminal boxes being installed in the plant.

Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Unresolved Item Failure to Mitigate Potential Harsh Environments from 71111.21N (Open) HELBs URI 05000250, 251/2019011-01

Description:

The inspectors are concerned that the licensee incorrectly classified certain areas in the turbine building as mild environments when the areas contain high energy fluid lines that could create harsh environments if pipe cracks or breaks occur in the lines. The high-energy fluid lines are near safe shutdown equipment. For these areas, critical cracks may have been required to be assumed by the sites current licensing basis. Licensee calculations did not address the possible harsh environments from high-energy line breaks (HELBs) in the turbine building. This was based on the turbine building design being open to the atmosphere; its structure has floors with no sides. Calculation 87-264.6000_000_2, Section 7.3, Environmental Effects, stated, in part, since the blow down piping system is located outside (open to the atmosphere) flooding and temperature rise are of no significance and therefore not a concern to this analysis. All the postulated line breaks in this calculation had this same statement. The jet impingements temperatures could exceed 212 °F, and steam in an outside at atmospheric conditions could reach at least 212 °F, both of which would be classified as a harsh environment.

In addition, calculation PTN-BFJM-92-016_000 determined that a break in a three-inch main steam line would have a mass flow rate of 215,645 lb./hr. Main steam is 1034 psi and 547 °F.

A fraction of this flow jet could cover a large area with 212 °F steam. The calculation indicated that a break of this size would not cause a plant trip, so it could persist until it was noticed and manually stopped.

The high energy fluid lines in the turbine building are not safety related. The UFSAR Section 5A for Seismic Classification & Design Basis does not list the lines as Class 1 (seismically qualified) and they are adjacent to the safe shutdown electrical equipment. The inspectors are concerned that a HELB in certain turbine building areas could cause a plant trip and thus a potential loss of offsite power. There are feedwater and main steam lines in the area. The cracks in the high energy lines, as discussed in the guidance the licensee stated they used to develop their analysis (Giambusso letter errata dated 1/26/1973) range from four to twelve inches long, which are equivalent analytical pipe diameters of one to seven inches. The inspectors are concerned that either a feed water line HELB or a steam line HELB could create harsh environments in the areas. A HELB in these lines could create an initial transient overpressure that the doors in the area may not be designed for. These doors are to rooms that contain multiple trains of safety related switchgear.

The feedwater jets may not immediately flash to steam, which could impact the doors in the area. The jets from steam or feedwater may have enough energy to physically open some doors and damage some electrical equipment such as exposed cables or MCCs. The fluid release from any of the lines could elevate the temperature to at least 212 °F and envelope the areas with high humidity that could condense inside electrical equipment for which they are not qualified.

The Turkey Point Nuclear Generating Station may not have met the following licensing basis requirements:

Turkey Point Nuclear Generating Station received information notice (IN) 2000-20, "Potential Loss of Redundant Safety Related Equipment Because of the Lack of High-Energy Line Break Barriers," ML003760571. The IN informed licensees of the requirements to address the environmental effects from HELBs (critical cracks) near safety related SSCs. In addition, the IN 2000-20 identified four conditions that must coexist to produce a risk-significant configuration.

1. lack of a HELB barrier between the redundant trains of a system that is needed to

mitigate accidents,

2. the lack of environmental qualification for the redundant components of trains located

in the same area,

4. the lack of a HELB barrier between adjacent piping and the redundant safety system

trains.

The Turkey Point Nuclear Generating Station may not have met the following licensing basis requirements:

10 CFR 50.55a(h)-IEEE 279-1968, stated in part, the design basis of protection system equipment shall be provided and based on, the malfunctions, accidents, or other unusual events (e.g., fire, explosion, missiles, lightning, flood, earthquake, wind, etc.) which could physically damage protection system components or could cause environmental changes leading to functional degradation of system performance, and for which provisions must be incorporated to retain necessary protection system action.

The Giambusso letter dated December 18, 1972, General Information Required for Consideration of the Effects of a Piping System Break Outside Containment, in conjunction with General Design Criterion 4, required, in part, that, structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. The licensee specified that they performed an analysis in accordance with the Giambusso letter and errata thereto. The errata specified, in part, that where pipes carrying high energy fluid are routed in the vicinity of structures and systems necessary for safe shutdown of the nuclear plant, supplemental protection of those structures and systems shall be provided to cope with the environmental effects (including the effects of jet impingement) of a single postulated open crack at the most adverse location(s) with regard to those essential structures and systems. The licensee responded, in part, that equipment is located sufficiently distant from the postulated high energy line breaks so as to ensure operability throughout cooldown in the event of a postulated pipe break. It was not evident to the inspectors how the licensee arrived at this conclusion given a lack of documented analysis concluding this.

10 CFR 50.49(d) dated 1984, stated in part, the licensee shall include the information in paragraphs (d)(1), (2), and

(3) of this section for this electric equipment important to safety in a qualification file.

1. The performance specifications under conditions existing during and following design basis accidents, 2. The voltage, frequency, load, and other electrical characteristics for which the performance specified in accordance with paragraph (d)(1) of this section can be ensured, 3. The environmental conditions, including temperature, pressure, humidity, radiation, chemicals, and submergence at the location where the equipment must perform as specified in accordance with paragraphs (d)(1) and

(2) of this section.

Planned Closure Actions: This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation and Regional Counsel, to determine if this issue of concern constitutes a violation.

Licensee Actions: The licensee captured the inspectors questions in their corrective action program.

Corrective Action References: AR2324737

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On August 16, 2019, the inspectors presented the design basis assurance inspection (programs) inspection results to Mr. Stamp and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.21N Calculations PTN-BFSM-11- NRC Generic Letter 89-10 MOV Design Basis Differential 0

20 Pressure Determination - Post-EPU

Corrective Action 01761767 W/O#40034286-01, Lube ECC Motor 04/12/2012

Documents 02305467 MOV-4-751: Inactive Boric Acid @Stem and Packing Gland 03/13/2019

Corrective Action 02324144 10 CFR 50.49 Check Box 08/12/2019

Documents 02324350 NRC DBAI EQ Program Inspection - Program Applicability 08/13/2019

Resulting from 02324618 DOC Pac 16.0 Eval of Test RPT Anomalies 08/15/2019

Inspection 02324654 DOC 17.0 Changes 08/15/2019

Drawings 5610-E-302 Electrical Installation Raceways, Notes, Symbols & Details Rev. 22

5614-E-26 SH Auxiliary Feedwater Pump Steam Supply D.C. MOV-4- Rev. 9

2G 1405

5614-E-377 Connection Diagram for Motor Operated Valves Rev. 7

5614-E-397 Connection Diagram - Control Board 4C02 Rev. 6

5614-E-486 Connection Diagram - Containment Isolation Rack 4QR51 Rev. 5

Engineering 1003483 EPRI Technical Report 1003483 Comparative Analysis of 0

Evaluations Nebula and MOV Long Life Greases for Limitorque Main

Gearbox Applications

600456 Limitorque Test Report No. 600456 Qualification Type Test 0

Report Limitorque Valve Actuators for PWR Service

B0058 Limitorque Test Report No. B0058 Limitorque Valve 0

Actuator Qualification for Nuclear Power Station Service

DOC PAC No. Generic Backup Documentation Rev. 10

1000

Documentation Nuthern Starters Rev. 1

Package 14.1

Documentation Doc Pac - Joy Manufacturing Co. Electrical Fans 13

Package 16.0

Documentation Doc Pac - Limitorque Corporation Valve Actuators 13

Package 17.0

71111.21N Engineering Documentation Limitorque Corporation Valve Actuators 7

Evaluations Package 17.3

Documentation Target Rock Solenoid Valves Rev. 10

Package 26.0

Documentation Automatic Valve Company (AVCO) Solenoid Valves 4

Package 41.0 Page 1

JPN-PTN-SEEJ-92- Engineering Evaluation to Develop an Enhanced EQ 0

006 Profile

X-604 Joy Manufacturing Co. Test Report X-604 Qualification 0

Testing of Joy Axivane Fan and Reliance Electric Motor

Miscellaneous 2375 TRC Report 2375 Qualification Test Report Aging, Seismic, G

& Accident Simulation Test of Target Rock

Corporation 1 Solenoid Valve, Model 77CC-001_

3854 TRC Report 3854 Qualification Extension Analysis Report 0

for Project 83UU Solenoid Operated Globe Valves_

3996 TRC Report 3996 Qualification Test Report for the A

Environmental Qualification of the Target Rock Corporation

Solenoid Operated Globe Valves

3J8345-A-L Certificate of Compliance - Bechtel Power Corporation 10/20/1983

44400R97 Qualification Test Report For Automatic Valve Solenoid Rev. A

and Air Operated Valves for Use in Various Nuclear Power

Plants

600456 Limitorque Test Report No. 600456 Qualification Type Test 0

Report Limitorque Valve Actuators for PWR Service

B0058 Limitorque Test Report No. B0058 Limitorque Valve 0

Actuator Qualification for Nuclear Power Station Service

B0058 Limitorque Valve Actuator Qualification for Nuclear Power Rev. 0

Station Service

B0212 Limitorque Test Report B0212 Qualification Type Test 0

Report Limitorque Valve Actuators with Type LR Motor for

Westinghouse PWR

71111.21N Miscellaneous F-C3271 Franklin Research Institute Test Report F-C3271 0

Qualification Test of Limitorque Valve Actuator in a Steam

Environment

FPL-12808R Nutherm Qualification Report for Nutherm Model 72978 & Rev. 3

Model 73010 DC Starter Panels

Procedures 0-ADM-540 Motor Operated Valve Program 4

0-ADM-703 10 CFR 30.49 Environmental Qualification 3

0-ADM-704 Environmental Qualification Maintenance Index Rev. 7

Documentation Environmental Qualification Generic Approach and 10

Package 1001 Treatment of Issues

ER-AA-112 Environmental Qualification Program 4

ER-AA-112-1000- Environmental Qualification Program Document Control Rev. 0

10000 and Retrieval Guidance

ER-AA-116 Motor Operated Valve Program 3

Work Orders 40016308 01 EQ-MOV-4-751 MOV & Grease Inspect 04/08/2011

40189647 01 MOV-4-751: MOV & Grease Inspect 10/08/2014

40470455 01 MOV-4-751: MOV & Grease Inspect 10/22/17

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