ML20294A104

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NRC-2018-000103 - Response 11 - Interim, Agency Records Subject to the Request Are Enclosed, (Part 3 of 4)
ML20294A104
Person / Time
Issue date: 06/10/2020
From:
NRC/OCIO
To:
Shared Package
ML20294A094 List:
References
FOIA, NRC-2018-000103
Download: ML20294A104 (102)
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Text

' U.S. NR C Uni1rJ St3 ln Nud ..-ar R~1t1..1'2torr C.:,,mm h\iOII Prott:f'tr1lK Ptopk and d,e E1wfromne11 t

Agenda 7 U.S.NRC Unired Scates Nuclear Regulatory C ommission Proucting People and the Environment

  • 1. Reactor Internal Basics
  • 2. Fuel Basics
  • 3. Engineering Concerns for Reactor Internals
  • 4. NRC Reviewer's Guidance to provide Oversight for Reactor Internals
  • 5. Licensing
  • A. Initial

Reactor Vessel and Internals - - . U.S.NRC U nited Srares N uclear Regularory Commission Typical Westinghouse-desi_qn PWR P..nt-,rl;.,,, p,,,,.,,.,,,, Affrl 1-I,,. Pnvironment Upper Support Plate Upper Support Column

.----Hold Down Sprint

---Control Rod Guide Tube Inlet Nozzle Lower Core Plate + - - Pressure Vessel

--.1a--t-~--- Former Plate lower Support Column Body---..

Bottom-mounted Instrumentation - - - - Lower Core Support Plate

Looking Down Into the Core Barrel - U.S.NRC United States Nuclear Regulatory Commission During Refueling Protecting People and the Environment

Reactor Vessel Internals '.

7 U.S.NRC Functions United States Nuclear Rcgufatory Commission Protecting People and the Environment

  • Support core (fuel)
  • Direct flow
  • Alignment

Pressurized Water Reactor - ?'U.S.NRC Internals - Materials & Operating U nited Staccs Nuclear Retiulatory C ommission Pruucting People and the Environment Conditions

  • Material is mainly Type 304. stainless steel
  • Reactor coolant water chemistry is carefully controlled to limit oxygen and impurities to very low levels
  • High levels of neutron irradiation
  • Some forms of materials degradation not seen elsewhere in the plant can occur in RVI due to the high radiation levels.

Pressurized Water Reactor 7 U.S.NRC Internals - Aging Effects in RV/ United States Nuclear Regulatory Co.mmissfon Protecting People and the Environmmt

  • Embrittlement
  • Neutrons damage the material over time:
  • Material strength increases
  • Ductility decreases
  • Combination of a corrosive environment and sustained stress causes cracking in certain materials
  • Not usually a problem in PWR internals because of good water chemistry

Aging Effects in RV/ (cont.) 7 U.S.NRC United States Nuclear Regulatory Commission Protecting Peopk a,ul the Environment

  • Fatigue
  • Repetitive stress cycles can eventually cause cracks. Usually not a problem in RVI because it is prevented by design procedures
  • Void swelling
  • Materials forms voids causing a volume increase
  • Only expected in the most highly irradiated parts of internals
  • Irradiation Creep
  • Under high irradiation an a constant stress, material can deform causing dimensional changes.
  • Irradiation Stress Relaxation
  • Caused by the same process as irradiation creep, but causes a loss of preload in bolts instead of dimensional change.
  • Can cause bolted joints to loosen.

Pressurized Water Reactor U.S.NRC Internals - Design of Reactor United States Nuclear Regulatory Commission Protecting Peopf.e and the Environment Vessel Internals

  • ASME Code, Section Ill did not have rules for design of RVI in early 1970's when many current operating plants were designed
  • Therefore, RVI for many plants were not designed or fabricated using ASME Code rules.
  • Designers tried to follow the Code to the extent possible for RVI.
  • For example, Indian Point Unit 3, used the stress criteria of the ASME Code, Section Ill as a guide during design.

Pressurized Water Reactor - ,/ U.S.NRC U nited Stares Nuclear Regulatory Commission Internals ASME Code Protecting Peopl.e and the Environment

  • American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code) provides the "rulebook" for design and inspection of nuclear plant piping and vessels.
  • Section Ill - Rules for Construction of Nuclear Facility Component

- Covers materials, design, fabrication , examination and testing during construction

  • Section XI - Rules for lnservice Inspection of Nuclear Power Plant Components

- Specifies periodic inspection of piping systems and vessels

ASME Code Cont.- Other Important -;J U.S.NRC United Stares Nuclear Regulatory Commission Sections for Nuclear Plants Protecting People a,uJ the Environment

  • Section 11 - Materials

- For example - Provides material properties such as allowable stress to be used in the design process

Pressurized Water Reactor .

71 U.S.NRC Un ited Stares N uclear Regulatory Co mmission Internals - lnservice Inspection Protecting Peopk and the Environment (ISi)

  • Plant systems are classified as Class 1, 2, or 3 based on safety significance.
  • Section XI requires ultrasonic examination of welds in Class 1 piping and vessels.
  • This is done on a sample for piping welds, but typically 100°/o of vessel welds.
  • Examinations are done on a ten-year interval.
  • For reactor internals, only visual examination is required by Sect~
  • Visual examination for general structural and me ical conditio~

called VT-3 by the Code. }

Pressurized Water Reactor Internals -

Licensing Basics U.S.NRC United States Nudea.r Regulatory Commission Protecting People and the Environment

  • Commercial power reactors are licensed under

- 10 CFR 50 (currently operating)

- 10 CFR 52 (new reactors - combined operating license)

  • Original operating license for all reactors is for 40 years.

Pressurized Water Reactor . U.S.NRC Un ited Sta,c.s Nuclear Regulatory Commission Internals - License Renewal Protecting Peopl.e and the Environment

  • Not all systems, structures and components (SSCs) are within the scope of 10 CFR 54.

- safety related

  • Prevent offsite radiation exposures in excess of certain limits

- Non-safety related whose failure could impact safety function of safety related components

- Needed for other regulated processes, such as fire protection or station blackout

Pressurized Water Reactor Internals - U.S.NRC License Renewal Un ited Stares Nuclear R.egulacory Commission Protecting People and the Environment

  • One of the key requirements for the Commission to issue a renewed license is that the applicant must provide reasonable assurance that the aging effects are managed, such that there is reasonable assurance that the systems, structures and components can perform their "intended functions" until the end of the 20-year license extension.
  • "Aging Effects" include corrosion, cracking, fatigue, and embrittlement
  • Embrittlement of a material means it takes less energy to break
  • Exposure of steel to neutron irradiation from the reactor causes the steel to become embrittled over time.
  • "Effects" vs "Mechanism"

LR In-Scope Components Subject .,;;U.S.NRC to Aging Management Uni ted States Nuclear Regulatory Commission Protecting People and the Environment s,eam Generator

License Renewal - What .7U.S.NRC United States N uclear Regulatory Commissfon components are subject to aging Protecting People and the Environment management?

  • Only systems, structures or components (SSC) that are "passive" and "long-lived" are subject to aging management.

- *"Passive".means an SSC performs its function without moving parts or a change in configuration or properties

  • Example -piping, reactor vessel, containment building, valve and pump bodies
  • However, valve internals such as stems and discs are excluded, because they move, as are pump impellers.

- "Long Lived" components are not subject to replacement based on a qualified life or specified time period. In other words, things expected to last the life of the plant.

Current Licensing Basis .

7 U.S.NRC United States Nuclear Regula tory Commission Protecting Peopk and the Environment

  • Per§ 54.3 Current licensing basis (CLB) is the set of NRC requirements applicable to a specific plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect. The CLB includes the NRC regulations contained in 10 CFR parts2, 19,20,21,26, 30,40,50,51,52,54,55, 70, 72, 73, 100and appendices thereto; orders; license conditions; exemptions; and technical specifications. It also includes the plant-specific design-basis information defined in 10 CFR 50.2 as documented in the most recent final safety analysis report (FSAR) as required by 10 CFR 50. 71 and the licensee's commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well licensee commitments documented in NRC saf ~ luations or licensee event reports. ~~

Time-Limited Aging Analyses -..5'U.S.NRC (TLAA) per§ 54.3: United $rares Nuclear Regulatory Commission Protecting People and the Envirunment

  • (1) Involve systems, structures, and components within the scope of license renewal, as delineated in§ 54.4(a);
  • (2) Consider the effects of aging;
  • (3) Involve time-limited assumptions defined by the current operating term, for examp_le, 40 years;
  • (4) Were determined to be relevant by the licensee in making a safety determination;
  • (5) Involve conclusions or provide the basis for conclusions related to the capability of the system, structure, and component to perform its intended functions, as delineated in § 54.4(b): and

License Renewal and Current . U.S.NRC United States Nucle.ar Regulatory Cor.nmission Licensing Basis Protecting People and the Environment

  • For license renewal, applicants need to show the CLB will be maintained for the period of extended operation, but are not required to make changes to the CLB.
  • For example, a structural analysis of the RVI is described in the IP2 UFSAR.
  • This analysis does not consider the properties of the RVI after irradiation, and has no time-limited assumptions.
  • This analysis is therefore not TLAA.
  • The applicant does not have to revise the RVI structural analysis to consider irradiated properties of the steel for license renewal.

Aging Management Progran,s 7 U.S.NRC U nited SurcJ Nuclear Regulatory Commission Protecting People and the Environment

  • Programs may be existing plant programs such as lSI or new programs just for license renewal
  • Programs may rely on inspection such as the ISi program, or may be preventive, such as Water Chemistry.
  • For RVI, staff determined ISi program wasn't sufficient by itself to manage all the aging effects.

License Renewal Guidance  ?'U.S.NRC United States Nuclear Regulatory Commissio!l Protecting People and the Environment

provides guidance for

  • aging management review - what aging effects should be managed for specific SSCs, and what programs manage them?
  • Analysis of Time-Limited Aging Analyses
  • Staff periodically revises these documents.
  • Staff may issue Interim Staff Guidance (ISG) if portions of GALL or SRP-LR need to be revised between major revisions of the NUREGs.
  • NRC guidance is considered one acceptable method for applicants to comply with 10 CFR 54; however, it is not regulation and applica may use other methods. J

Industry RV/ Aging Managen,ent .

7 U.S.NRC United Stares Nuclear Regulatory Commission Program Protecting People and the Envir011ment

  • The industry, led by the Electric Power Research Institute (EPRI),

began development of an aging management program for PWR internals around year 2000.

  • Industry program needed results from materials testing that was not complete yet
  • NRC staff allowed applicants for renewed licenses to make a commitment to implement the industry program once it was issued.
  • The report describing industry's recommended program was submitted to NRC for review and approval on 1/12/2009.
  • MRP-227, Rev. 0, Materials Reliability Program: Pressurized Water Reactor Internals Inspections and Evaluation Guidelines

MRP-227 U.S.NRC United Stares Nuclear Regulatory Corn mission Protecting People and the Environment

  • MRP-227, Rev. 0 is primarily an inspection-based program.
  • Prescribes inspections of certain components of the RVI using different methods than specified by .the ASME Code
  • Higher-resolution visual examination (EVT-1)
  • Ultrasonic examination (UT)
  • Components inspected based on likelihood and safety impact of aging
  • Specifies inspections for three different generic designs: Babcock &

Wilcox (B&W), Combustion Engineering, and Westinghouse.

  • NRC staff approved MRP-227 with several conditions and action items, in safety evaluation dated 12/16/11 .

Exa111ples of RV/ Component * / U.S.NRC Un ired States Nuclear Regulatory C o mmission Inspections - Baffle-Former Bolts Protecting People and the Environment (UT)

External Hex

Examples of RV/ Co111ponent . U.S.NRC United Stacces Nuclear Regulato ry Corn.mission Inspections - Control Rod Guide Protecting Peopk and the Environment Tube Lower Flange (EVT-1)

Thermal Shield Flexures (VT-3) .- us.NRC United States Nuclear Regulatory Commission Protecting Pet,pk and the Environment

Safety Evaluation Report - License U.S.NRC United Staces Nuclear Regulatory Commission Renewal Protecting Peopk and the Envinmment

  • Staff reviews the LRA against criteria in the GALL report and SRP-LR, and the regulations (10 CFR 54):
  • NRC staff may issue Requests for Additional Information (RAI) to the applicant when it needs more information to make a safety determination.
  • NRC staff may perform audits, particularly to verify applicant's claims that AMPs are consistent with GALL.

§ 2.309(f)(1) Says to be admitted a . U.S.NRC contention must: United S1accs Nuclear Rcgulacory Commission Protecting People and the Environment

  • (i) Provide a specific statement of the issue of law or fact to be raised or controverted . . .
  • (ii) Provide a brief explanation of the basis for the contention;
  • (iii) Demonstrate that the issue raised in the contention is within the scope of the proceeding;
  • (iv) Demonstrate that the issue raised in the contention is material to the findings the NRC must make to support the action that is involved in the proceeding;
  • (v) Provide a concise statement of the alleged facts or expert opinions which support the requester's/petitioner's position on the issue . . . ; [and]
  • (vi) . .. . [P]rovide sufficient information to show that a genuine dispute exists with the applicanUlicensee on a material issue of law or fact.

- I.e., you can't just say "Nuclear is bad"

§ 2.309 - Requirement for Standing - ,.-:; U.S.NR C U nited Srarc:s Nuclear Regulatory Commission Protecting People and the Environment

  • Parties filing contentions must have "standing"
  • Key factors for the ASLB to consider with respect to standing are:
  • The nature and extent of the requestor's/petitioner's property, financial or other interests in the proceeding; and
  • The possible effect of any decision or order that may be issued in the proceeding on the requestor's/petitioner's interest;

Conduct of License Renewal ~-, U.S.NRC United Scates Noclcar Regulatory Commission Hearings Protecting People and the Environment

  • Three-judge panel
  • Prefiled written testimony is filed by all parties.
  • Hearing is conducted similar to a trial
  • Expert witnesses testify for each party
  • Attorneys are present for each party
  • Judges ask questions of the witnesses
  • Judges may or may not allow cross examination
  • The ASLB will rule on the contentions after the hearing.
  • Parties may appeal.

Case Study- Indian Point * .. .

7 U.S.NRC United States Nuclear Regulatory Commission Protecting People ana the Environment

  • SER for LR of IP2 and IP3, NUREG-1930, was issued in August 2009 after a 28-month review period.
  • Normally, the commission would issue the renewed license within about 6 months of the SER publication.
  • Due to contentions, IP2 and IP3 renewed license has been delayed.
  • For IP2 and IP3, after the LR SER was published, but before a .

renewed license was issued, the applicant submitted new information.

  • The new information was an amendment to the LRA containing a RVI AMP based on MRP-227, Rev. 0.

Case Study - Indian Point ,. U.S.NRC United Stares Nuclear Regu lacory Commission Protecting Peopk and the Environment

  • Since the staff had not yet approved MRP-227, Rev. 0, the review of the amendment went through several iterations.
  • Applicant also submitted its RVI Inspection Plan on 9/28/2011, 24 months before IP2 end of original 40-year license, to fulfill its commitment.
  • A revised RVI inspection plan submitted on 2/17/2012 consistent with NRC-approved version of MRP-227, published 1/9/2012. (MRP-227-A)
  • NRC reviewed RVI AMP and Inspection Plan using Interim Staff Guidance that updated Rev. 2 of GALL to reference MRP-227-A.
  • NRC issued Supplement 2 to the SER on 11/7/2014 to document its review and approval of IP2 and IP3 RVI AMP and Inspection Plan.

Hear ings ./ U.S.NRC United States Nuclear Regulatory Commission o.._..-,.._ 0 -~1- __ ,1 J.. Environmmt

Case Study - Indian Point United Staces Nuclear Regulatory Commission Hearings Protecting People a,u,/ the Environment

  • Atomic Energy Act Section 189 provides allows Commission to grant a hearing for granting, suspending, revoking or amending any license
  • Atomic Energy Act Section 191 establishes the Atomic Safety and Licensing Board and authorizes it to conduct such hearings
  • § 54.27 allows for hearings under §2.309, "Hearing requests, petitions to intervene, requirements for standing, and contentions"
  • Under§ 2.309, Any person whose interest may be affected by a proceeding and who desires to participate as a party must file a written request for hearing and a specification of the contentions which the person seeks to have litigated in the hearing.
  • ASLB also conducts the hearings related to granting licenses and amendments under § 50, petitions under §2.206. All hearings are conducted under the rules of §2.

Case Study - Indian Point - u.S.NRC Contention NYS-25 United Staccs Nucle.ar Rcgul.arnry Commission Protecting People and the Environment

  • The LRA does not include an adequate plan to monitor and manage the effects of aging due to embrittlement of the reactor pressure vessels. ("RPVs") and the associated internals at both plants, pursuant to 10 C.F.R. § 54.21 (a), and an evaluation of time limited aging analysis, pursuant to 10 C.F.R. § 54.21 (c).
  • Amended - Acknowledged that Entergy has a program for managing aging of reactor vessel internals, but argues that it is inadequate.

Focus shifted away from RPV to RVI.

  • 2 nd Amendment - After staff SER was published, new informatioA4 available, so ASLB admitted J

Case Study - Indian Point U.S.NRC United States Nuclear Regulatory Commission Contention NYS-25 Protecting People and the Environment

  • One of New York's bases for NYS-25 was that Entergy should revise the RVI structural analyses to consider embrittlement.
  • This is not required by § 54 since the analysis in the CLB didn't consider embrittlement.
  • Other bases included:
  • Failure to considered combined or "synergistic" effects of multiple aging effects, particularly fatigue and embrittlement
  • Inadequacy of inspection methods,
  • Components not being preventively replaced
  • Fatigue calculations did not include an error analysis

Case Study - Indian Point U.S.NRC United Sraues Nuclear Regularory Commission Hearing Aftermath and Tin,e/y Protecting People and the Environment Renewal

  • Hearing on NYS-25, plus two other contentions, held November 16-20, 2015 in Tarrytown NY.
  • The ASLB will probably take several months to issue its decision on NYS-25 and the other contentions.
  • Due to contentions, the commission has still not made a decision on whether to issue a renewed license for IP2 and IP3
  • Meanwhile, the original license for IP2 expired on 9/28/2013, and for IP3 expired on 12/12/2015.
  • Because the NRC has not made a final determination on whether to issue a renewed license, under the regulation called "timely renewal",

the plants are allowed to continue to operate until the NRG makes a final decision.

Case Study - Indian Point .

7 U.S.NRC United States Nuclear Regulatory Commission

/P2&/P3 License Renewal Protecting People anJ the Envirt1nment Timeline (Backup)

  • Contained commitment to implement industry program for reactor vessel internals
  • New York fifed petition to intervene - 11/30/2007
  • Industry guidance published for RVI inspection - MRP-227, Rev. 0 -

1/12/2009

  • Safety Evaluation Report (SER) issued - 8/11/2009
  • Amendment 9 to LRA submitted to NRC - 7/14/2010

IP2&/P3 License Renewal U.S.NRC Timeline Un ited States Nuclear Regulatory Commission Protecting People and the Environment

  • NRC Staff Final Safety Evaluation of MRP-227, Rev. 0 - 12/16/11
  • Entergy submits RVI inspection plan consistent with NRC Final SE -

2/17/2012

  • NRC staff issues Supplement 2 to Safety Evaluation Report for License Renewal, in which the staff approved the RVI program - 11/7/2014
  • New York filed motion to add more based to NYS 2/13/2015
  • ASLB granted motion to add bases - 3/31/2015
  • Hearing on Contention NYS-25 could not be scheduled until staff made its finding on the RVI program for IP2 and IP3.
  • Hearing was held November, 16-21, 2015, in Tarrytown, NY (aloJJar with two other contentions). J

OFFICIAL US! Ol~Li - S!l~SITIOI! 114Tl!PU4Al 114f6RMA"f'i614 00 f467f pre"nAfte Jlcf4¥ EMeERPTS Ol:IT81BE QF ~*RS 'NITtiSI.IT FIRiT OiTAl*11w. PliRMISSIQN FROM ORIGINA'l'O" OpECOMM Operating Experience Communication Good Judgment Comes from Experience Information Security Reminder - OpE COMMs contain preliminary information in the interest of timely internal communication of operating experience. This information is olten subject to change and is not intended for distribution outside the agency in this form.

St. Lucie Unit 1 - Rebuilt MSIV Fails After 230 Days in Use Executive Summary:

To support an Extended Power Uprate (EPU), new internals were installed in both Main Steam Isolation Valves (MSIV). It was not known at the time, but new tail links did not meet design specification dimensional drawings (they were oversized). The oversized tail links did not allow the valves to backseat fully, leaving the disc partially in the flow stream. This exposed the valves to unintentional dynamic loading, which ultimately resulted in the failure of internal parts. For the "B" MSIV, the failure caused the valve disc to seat uncontrollably while at 100% power, tripping the reactor. The direct cause was that the valve disc was not backseated which allowed unintentional loading of internal parts. The root causes were that the tail links did not meet design specification dimensional drawings and that the engineering change package did not include a verification that the modified valves would open fully.

COMM Groups Notified:

All Communications, Inspection Programs, Main Steam & Condensate/Feed Systems, Materials/Aging, New Reactors, Power Uprate, Pump and Valve Performance, and Vendor & Quality Assurance.

Background:

St. Lucie Units 1 and 2 are two loop Combustion Engineering plants. St. Lucie 1 received a Power Uprate approval on July 9, 2012 (ML12191A220). The approval authorized a 10.0% extended power uprate (EPU) and a

1. 7% measurement uncertainty recapture, increasing the maximum steady-state reactor core power level from 2700 megawatts thermal (MWt) to 3020 MWt.

To support the approved higher power levels, the Main Steam Isolation Valves (MSIV) internals had to be upgraded to insure they could withstand the impact stresses associated with spurious closure at the higher steam velocities. It was determined that the original MSIV body (Schutte and Koerting, Co (now Ametek, Inc.), model M?0-00656-V) could be retained and did not need to be replaced. Kalsi engineering provided the design and analysis (Engineering Change Package EC246556) with an independent third party design review by Zachary Engineering. Zachary Engineering (a separate group from the independent design analysis) provided the installation and post maintenance testing requirements. Flowserve manufactured and performed the field work to install the new internal valve components. Enertech provided and installed a new electro-hydraulic actuator to replace the existing pneumatic actuator. Bechtel provided QA services at Flowserve during parts manufacture.

[NOTE: the licensee's root cause evaluation noted that this complex project structure created an error likely environment.]

Ol'l'ICIAL t131! 014LY - SE'4BITIVE lflTERUil<<L IPIF8RMATIQ~l 89 ~l8:r F8RWil<teCRPf3 e~T81BE er URS WITI ,e~T FIRST 8BTAIP41f48 PERMl8818P4 FR9M 0"10110tt6R AlA PISTON INST RUM ENT A IA VIA ACCUMULATO RS (2perMS1V)

INLET

.. OUTLET TRIP VALVE CHECK VALVE Figure 1 St. Lucie Unit 1 Main Steam Isolation Valve Shown in the Open Position The new internals and actuator were installed during a planned shutdown on July 18, 2012 (See Figure 1). The work orders that installed the internals and actuator required stroke length measurements and provided acceptance criteria (as specified in the modification acceptance test plan). This was identified as a critical step to insure the valve would stroke fully closed to fully open. The work orders were closed even though the acceptance criteria were not satisfied for either MSIV. FPL procedures requlr,e an action request (AR) for "unexpected or unwanted conditions". The discrepant stroke length was documented, but no AR was initiated.

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o prpr1el)!ct tJ!f'. 6f4LY : SEHSl'fl'o'c 1Wffftf4AL lfff8RMA-Fl8U ee UO'f F8RWARB AP4¥ E:M8ERP:PS 6t:l'fS1BE eF l4fte 'fil'fl 16t:IT pr1ft!T 015TAllfllfe fl !l'tMl!!IOl4 l'l'tOIO'I Ol'tl61f4A-F8R During post installation limit switch adjustments, the limit switches could not be set within existing limits. Change request notices (CRN) documented this. The engineering resolution to the CRNs focused on limit switch set point adjustments and limit switch mounting changes, and did not address why the changes were needed. (Later investigation determined that this was a missed opportunity to detect the problem. A more thorough engineering review may have prevented the problem.)

From July 26 through July 29, startup testing was performed, and unit 1 reached the new 100% power level on July 29, 2012.

Issue

Description:

On March 12, 2013, after 230 days on line at the new rated power level, St. Lucie Unit 1 tripped on thermal margin/low pressure (TMLP) from an asymmetric steam generator (SG) transient (275 psid between the SGs) signal (See EN 48818). After the trip, the licensee determined that, with the valve positioned to "OPEN", the valve did not pass enough flow to maintain steam header pressure. Steam header pressure could only be maintained with the MSIV bypass valve in service. The licensee cooled down to Mode 5 to inspect the MSIVs internally. Internal inspection found that the "B" MSIV had a spindle to disc separation. A similar inspection of the "A" MSIV found the disc still attached to the spindle, but the spindle itself was damaged. See Figure 2 below.

INLET OUTLET TRIP VALVE CHECK VALVE Figure 2 MSIV Close Up Showing Tail Link, H-Link and Spindle OFFICIAL USE ONL t - SENSI I IVE IN I ERNAL INFORMA I ION 00 140T f'OPtoVA"" Al4'1' !)(Cl!ftl'T! Otl'f!lt,! er ro~.e Wl'fl 16t:l'f FlftS'f 6B'fAIUIU0 PERM1SSl0t4 FROM 8R1SIPJM8R

OFFICIAL OSE ONLY - Sf,qS,TIVE 114TEfU.ilcL itffBRMATIOOil DU NO I FORWARD ANY EXCERP IS OU I SIDE OF NRC WI I AOO I FIRS I 081 AINING PERMISSION FROM OPU6,UM8R The license discovered that the back of the tail link for both MS IVs made contact with the valve body before the back seat stopped valve motion (notice the worn portion of the tail link in Figure 3). The tail link was dimensionally checked against the design requirements, and det,ermined to be oversized.

Figure 3 MSIV Tail Link With the disc not fully backseated, it remained partially exposed t,o the steam flow through the MSIV. Excessive stress was caused by the valve disc remaining partially in the flow stream. These stresses caused the shear pin for the "8" MSIV to fail (See Figure 4). When the shear pin failed, the spindle separated from the disc. System flow then uncontrollably forced the valve disc into the seat, blocking flow from the steam generator (Inlet side) to the main steam system (Outlet side).

oFFICIAL USE ONLY . StNSI rIVE IN fERNAL lNfORMAl ioN B6 140T l'OfhfAft.O Af4 t EXCEftF IS OtJTSIOE OF IQRC vvi I HOO I FIRS I 08 I AINING PERMISSION FROM Oft.181P4A:rBR

Ol'l'ICIAL 1:1!1! 014LY - !1!14!1Th/l! 114Tl!PU4AL 1141'0"MATIOl4 QQ ~IQ+ FQR't'.'AA8 AtlY EXQl!:!RP+B QijT818E Qf tlRQ IJIITI 181:!if FIRSif 8BTAIPHtl6 PERMISS18PI FR8M ORIGINAi OR Figure 4 "B" MSIV Broken Shear Pin They also discovered the same oversized tail link and interference on the "A" MSIV. However, instead of the shear pin failure seen on the "B" MSIV, the spindle tab had partially torn (See Figure 5). The licensee determined that, once the partial tear occurred on the spindle tab, the "A" MSIV disc fully backseated, which prevented the in-service failure.

8fFlelAL l:ISE 8HLY SEPISlifl'/li! l~ITlsR~lAla 1*11aORU UIOM DO NO f FORWARD ANY EXCERP IS OU I SIDE o.- IU~C VUITl"IOl:IT .. ,"a' 68TAIP41P48 PERM1SBl8~l FRQM ORlf;i l~I AIOR

QliiliilCIAb YSE 8JJbY SDJSFFI¥[ IPffERIML ll~reKMl<TION ee ,~e, l'O"hl<"" AW( l!!,Cel!!fltl'TS OtJTSIDI!! 01' l~"e HI, .. ou I FIRS I OB I AINING PERMISSiON FROM ORIGll41t1f6R Figure 5 "A" MSIV Damaged Spindle Connection When the back of the tail link made contact with the valve body prior to the valve being fully backseated, the valve disc was not fully in its open position. When fully opened and backseated, the valve disc must be open at least 80 degrees with a 1/8 inch clearance to the body to account for thermal expansion. The newly installed tail link was oversized, making contact with the valve body before the seat was backseated. This caused excessive, unintended stress on the valve components, ultimately resulting in the shear pin failure on the "B" MSIV causing the disc to separate from the spindle. When the pin failed on the "B" MSIV, the valve shut uncontrollably.

The excessive stresses imposed on the valve from the uncontrolled shutting of the MSIV from 100% power required extensive engineering analysis and replacement of several internal components, including the valve seat, and some external supports. The engineering analysis determined that the valve body was acceptable for continued use.

The licensee determined that there were 2 root causes for this event:

1. The tail links provided by Flowserve did not meet the design specification dimensional requirements.

The contract with Flowserve required all parts be provided per detailed machine drawings. Shop inspections and dimensioning did not identify the tail link discrepancies. QA inspections by Bechtel did not OFFICIAi Pili 0*11.v ilitilillU<lii 1*1:rliRCtlALo lm7QRMMIQ*l DO *10:r 5'.0RWARQ 01¥ liXCliiRR:ri Oll+iltlli or;: CtlRC w1:rwo11:r r.lRiT 01no1*11w.. RliRHliilO*t r.ROH 8Rl61PJAlQA

0FFl01AL l:JSE 0HLY SEP4BITIIJE IP4TERPJitcL IPJF0RMitcTl9PJ B8 PJST F8R\1(itcR8 AIH' EKeEftf'l! 6t:IT31t,I!! 01' rcfte Wlrl"IOtlT l'lft!T O!TAIIQING Pl!RIOll551014 FROIVI 6Rl61~lA'iQR identify the dimensional discrepancies. FPL QA surveillance was waived for this hold point, as allowed by FPL procedures. This cause was characterized as a personnel performance issue.

2. The engineering change package did not include verification that the modified valve would open fully, i.e.,

to the back seat. The EC process includes requirements to specify required implementation instructions and post maintenance testing. Although multiple engineering groups reviewed the EC, the package did not contain specific instructions to assure the valve is fully open to the backseat. This cause was characterized as a personnel performance issue.

Licensee Event Report 3352013001RO (ML13142A200) was issued by the licensee on March 12, 2013.

The licensee completed an extent of condition review and determined that no other valves were affect~d. It should be noted that the Unit 2 MSJVs are a completely different design. The Unit 2 MSIVs are 34 inch Rockwell International (Model Number PD-153115) angled valves with an internal pilot and balance chamber. Licensee discussions with Flowserve indicate that they don't believe that any other plants have had this type of modification performed.

lf anyone has additional information related to this OpE COMM, please contact Bob Bernardo at Robert.Bernardo@nrc.gov or 301 -415-2621.

The following insights associated with this event apply to operating reactor and new construction inspection activities:

1. Oversight: Adequate oversight and inspection are critical to ensuring quality. In this case, the licensee established a complex project organization to complete this design change that relied on the use of multiple vendors. The inspections performed did not ensure that the tail links provided by Flowserve met the design specification dimensional requirements, as required by Criterion X of 10 CFR 50 Appendix B. Appendix 10 of IP 35007, "Quality Assurance Program Implementation During Construction and Pre-Construction Activities", IP 71111 .18, "Plant Modifications" and section 03.10 of IP 43002, "Routine Inspections of Nuclear Vendors", provide guidance for inspecting independent oversight.
2. Testing: Per Criterion XI of 10 CFR 50 Appendix B, verification testing must incorporate the requirements and acceptance limits contained in applicable design documents. Testing anomalies, whether at the vendor shop or in-situ, must be thoroughly investigated and dispositioned. In this case , post modification testing to verify that the MSIVs would fully open was not specified in the EC Package. In addition, the anomaly caused by the stroke length measurements not meeting the acceptance criterion defined by the work order instructions was not thoroughly investigated. I P-65001.D, "Inspection of the ITMC-Related Operational Testing Program", IMC 2513, "Light Water Reactor Inspection Program - Preoperational Testing And Operational Preparedness Phase", IP 71111.19, "Post Maintenance Testing" and IP 43002, "Routine Inspections of Nuclear Vendors", provide guidance for inspecting testing associated with construction activities and post maintenance testing for operating reactors.

Other Operating Experience:

Public Web Page on Power Uprates: http://www.nrc.gov/reactors/operating/iicensing/power-uprates.html Previous OpE COMMS:

OFFICIAL USE 01\IL I - 31!!1431Tl't'E lf4lERPJAL ltff8RMMl9M DQ tl9:r F8R'NARB AtJY E>E8ERPTS et:l'fSIBE er f4"e Wlfl 16t:tT l'lft3T 08TAll411qG Pl!RIVIISSION FROM l>RIGllqAf8R -

Ofil'lelAL t:ISE 8ULY 6Etl81?1VI! IHH:R*IuA.1.. 1*11ao~aUTION DO NO:r ~O~'A'.o:Rg n,v iX:,liiRP:ri QUlilQi Qfa ~JR" '.t.'ITflQWT I-IRS+ 8BTA1Wt1S PERMISSIQPJ F'R8M Oftl61f4M"8R Main Steam Isolation Valve Failures at Farley Unit 1 http://nrr1 O.nrc.qov/forum/forumtopic.cfm?selectedForum=03&forumld=AIIComm&topicld=889 MSIV inoperable due to internal binding at McGuire:

http://nrr10.nrc.gov/forum/forumtopic.cfm?selected Forum=03&forum ld=AIIComm&topicld= 1049 Main Steam Isolation Valves Fail to Close (SIT) at Harris:

http://nrr10.nrc.gov/forum/forumtopic.cfm?selected Forum=03&forumld=AIICom m&topicl d=3841 Two MSIV Steam Failures Due to Thermal Embrittlement: at Vogtle:

http://nrr1 O. nrc.gov/forum/foru mtopic.cfm?selected F orum=03&forum ld=AIIComm&topic ld=4012 Issue for Resolution (IFR) 2005-069: Davis Besse - Potential lnoperability of AOVs to Function During Design Basis Conditions: http://nrr10.nrc.gov/ope-info-gateway/ifr/2005/1 FR%202005-069.pdf Previous Generic Communications NRC Information Notice 2004-15: Dual-Unit Scram at Peach Bottom Units 2 and 3 (MSIV failed to Close on Demand)

NRC Information Notice 1995-04: Excessive Cooldown and Depressurization of the Reactor Coolant System Following a Loss Of Offsite Power (MSIVs for steam generators A and B failed to close fully)

NRC Information Notice 1994-44: Main Steam Line Isolation Valve Failure to Close on Demand Because of Inadequate Maintenance and Testing NRC Information Notice 1994-08: Potential for Surveillance Testing to Fail to Detect an Inoperable Main Steam Isolation Valve NRC Information Notice 1990-79: Failures of Main Steam Isolation Check Valves Resulting in Disc Separation NRC Information Notice 1988-59: Main Steam Isolation Valve Guide Rail Failure at Waterford Unit 3 NRC Information Notice 1988-51: Failures of Main Steam Isolation Valves NRC Information Notice 1986-106: Feedwater Line Break (The event was initiated by the main steam isolation valve on steam generator C failing closed) - 4 fatalities NRC Information Notice 1986-81 : Broken Inner-External Closure Springs on Atwood & Morrill Main Steam Isolation Valves NRC Information Notice 1986-57: Operating Problems With Solenoid Operated Valves at Nuclear Power Plants NRC Information Notice 1985-84: Inadequate lnservice Testing of Main Steam Isolation Valves NRC Information Notice 1985-59: Valve Stem Corrosion Failures NRC Information Notice 1985-21: Main Steam Isolation Valve Closure Logic NRC Information Notice 1984-35: BWR Post-Scram Drywell Pressurization (inboard main steam isolation valve (MSIV) failed closed)

NRC Information Notice 1984-33: Main Steam Safety Valve Failures Caused by Failed Cotter Pins NRC Information Notice 1983-70, Supplement 1: Vibration-Induced Valve Failures NRC Information Notice 1983-54: Common Mode Failure of Main Steam Isolation Non Return Check Valves NRC Information Notice 1982-23: BWR Main Steam Isolation Valve Leakage 9FFl6h\l ~SE 9tll¥ 8Et1Slfl'IE IPffER~*AL IUF8RMM"18t4 ee ,.e, re"'i\1Al(e A14, l!!)(Cl!!lftP'TS OU I SIDI!! opi 1u,e 001 I HOO I FIRS I OB I AINING PERMISSION FROM

""161NA+QR.

OFFICIAL USE ONLY

  • SENSI I IVE 11q I !N.14At 114~0ft:MATl614 DO NOT FORWARD ANY EXCERPTS OUTSIDE OF NRC WI I ROOT FIRS I 08 I AINING PERMISSIOlq l'~OM QRl<.INAIQB NRC Information Notice 1981-38: Potentially Significant Equipment Failures Resulting from Contamination of Air-Operated Systems NRC Information Notice 1981-14: Potential Overstress of Shafts on Fisher Series 9200 Butterfly Valves with Expandable T Rings (maximum shaft stress underestimated}

IE Circular 1981-14: Main Steam Isolation Valve Failures to Close NRC Information Notice 1980-16: Shaft Seal Packing Causes Binding in Main Steam Swing Oise Check and Isolation Valves (b)(4) (b)(7)( 0 )

Licensee Event Reports (LER):

LER Link Date Site Description 4242012005 10/08/2012 Vogtle 1 Main Steam Isolation Valve Failure 4002012001 04/21/2012 Harris Delayed Closure of Main Steam Isolation Valves due to Corrosion 2372011003 10/17/2011 Dresden 2 MSIV Closure Times outside of Technical Specifications Limits 2492010002 11/01/2010 Dresden 3 MSIV Leakage Exceeds Technical Specifications Allowable Limits 3482006002 04/08/2006 Farley 1 Main Steam Isolation Valve Failure to Close Failure of Main Steam Line Isolation Valve 3702005005 03/02/2005 McGuire 2 (MSIV) to Close Main Steam Isolation Valve Failed to Close 2192004005 09/11/2004 Oyster Creek During Partial Valve Closure Surveillance Due to Mechanical Binding Closure Response Time Exceeded For Main 4131996008 03/07/1995 Catawba 1 Steam Isolation Valve 1SM1 , B Train Failure of Main Steam Isolation Valve 1MS-2018 2661992006 05/31/1992 Point Beach 1 to Fully Shut During Performance of IT-280 3011991001 09/29/1991 Point Beach 2 Failure of Main Steam Isolation Valves to Close Failure of Two Outboard Main Steam Isolation 4581989043 12/01/1989 River Bend Valves Main Steam Isolation Valves Failure to Close 2371988012 05/17/1988 Dresden 2 Due to High Stem Drag Forces Caused by Valve Packing 8fftetJltL t:JSE 8P4LY SEt4Si:f'llJE IPfFERU!tL IPffQRMA'flQPJ oa NOI FORWARD M:IY EXCERRI$ 01 !Til~i 0~ ~lRC IO!ITMOUT ~IRST 98'.fAIHIHG PeRMISSIQPJ FR0M ORIGINAIQB

QFFISIAb WSE OP4LY SEt*Sl'fPIE IH'fERtJA:L lfiF6RMilr'f16f~

DO 140 I FOROvARO 1(14 I !!,CC!ft.. T! OtlT!IO! o, 14fte 'ilt1ITl10tlT l'lft!T Ol!TAJl41fJ6 PERMISS16fJ FR8M Oftl61UMQR LER Link Date Site Description 2131986029 06/17/1986 Haddam Neck Main Steam Isolation Valve Closure Test Failure 2451986006 02/06/1986 Millstone 1 Failure of 1-MS-10 and 1-MS 20 to Close 2501986005 01/27/1986 Turkey Point 3 Main Steam Isolation Valve 3331985027 11/22/1985 Fitzpatrick Main Steam Isolation Valve Failures Failure of Main Steam Line Isolation Valves 821 -

3241985008 09/27/1985 Brunswick 2 F028A, F022C, and F028C to Fast Close During Operability Testing Failure of MSIV to Fully Close During 3181985008 07/24/1985 Calvert Cliffs 2 Surveillance Testing 4541985027 03/14/1985 Byron 1 Failure of MSIV to Close on MS Isolation Signal Failure of #12 MSIV to Fully Close During 3171984019 12/12/1984 Calvert Cliffs 1 Surveillance Testing Failure of Main Steam Isolation Valve (MO-1551984013 09/09/1984 Big Rock Point 7050) to Close 3181982050 10/17/1982 Calvert Cliffs 2 Update on MSIV Failure to Close Vermont 2711981027 10/16/1981 MSIV Failed to Close Yankee Vermont 2711981022 08/01/1981 The MSIV Failed to Close Yankee Failure of the "D" MSIV to Close within Required 3661981071 07/18/1981 Hatch 2 Time Frame Qliiliil~IAL. flili O~ILY SE~ISIIIVE INIFRNAI INEORMAJIQN DO t40'f FOR'NA:RB A:PJY E>t8ERP'fB O~'fBIBE 6F F4flte WPfll6l:ff PlfltS, 68TAlf41f48 PERM1SSl6f4 FR6M 8A1Slt4M8 R

An:*ragl'  :\umber of Q Numhc*r of 1"iu111her of .\umbl'r of  :\umber of

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ROP LER 2000-2004 14 F~-C?Wll 11 Fllldi!lg - NIA 6 NOD-Cited Vio1.tlim - Crem 50 Noa-Ci.wt Violati<n - N'/A 3 Violaticn - Cretn 1 Violatioll - SL-1\f l Violation - Wluta 1 2005-2010 30 Findin.i: - Cnm 8 Fincf.ng - NIA 3 Non-Cited Violation - Grem 101 Non-Cited Violation - SL-IV I Vio1Ki.cm - C.Nlll l Violaion - SL-I\' l Violmon - Whitt 4 Viotttion - Vlllcnr l 2011-2015

- - 58 Fit.ding - Cnia. 11 Finding - NIA l Ncn-Citttl Vioatiost

  • enc. 1i4 Nm-Cit!!d Viotarion - NIA 8 Non-Cit~ V:iot.tion - SL-IV l ViOUltioo - Gn,m i Violation - NIA 12 Violation - Red l

\'toi.tion - WJme 6

Percent of Time FCS spent in certain column.

2000-Q1 2010 °/o " 57.89474 " 34.21053 r 7.894737 0 12000-CAL Q3 2010 °/o 55 32.5 12.5 ..

ALL Initiating Event Mitigating System Findu!J

  • GNm IS Finding
  • GAtll g Pindin:
  • NIA 1 Pillding - CfflJl 3 Noo-Cit9d Violation - G1'ltll 121 Non-Cited Vielati.on. CNm 6i NCD-Citecl Vio!.ttum
  • Gl'IIUI. 13 Non-Cited Viet.tiofl . SL-I\'

lioo-Cited VioladoD- NIA 1 2 NOD-Cited Vioi.tioo

  • 51.-l\' 1 Non-Cltld Violation
  • SL-I\' 4 Violation
  • Cffll!ll 2

\ti<.>Jati.on

  • c,._ 2 Vi.o1ation - Wtillt 4 Violation
  • NIA 1 Findul; - C rea. 1 Noi>-Cited Viol.tlion
  • Cffll!ll 5 'Nm-Cittd Viot.lion - Gn:o. l Fii>dinJ - NIA 4 Vioi.ticn - Whitt 1 Non-C!Cld Vlolltioa
  • s1.--n* 1

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lnitulling Evtnts Mitigating Systems !JI Sanier Integrity (5)S~nce

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Sclcctm:cn tbc list (6) J1em Types

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Sensitive

Initiating Event (Green 2011-2014)

Finding Non-Cited Violation Component cooling water system for temporary off normal system conditiqns ***

Ensure leak before break commitment ****

Loss of reactor coolant *****

Loose maintenance carts, improperly storred ladders, excessive transient combustible material, scaffolding, removal of foreign material ****

5 alarm response procedure 480 VAC breaker

HPSI pump flow imbalance

( changing load on the main turbine) *****

Mitigating Systems (Green 2011-2014)

Finding Non-Cited Violation CB20, panel A 18, Window C3 Reactor protection system Diesel generator ****

channel A trip unit 6

  • Failure to generate complete Safety injection refueling Raw water strainer inspection list water tank vortex eliminator component ***

River sluice gates Auxiliary f eedwater ring ***** Auxiliary feedwater pumps FW-6 and FW 10 ***

Raw water pump anchor bolts Reactor coolant pumps oil Fuel oil consumption collection system calculations ***

Extent evaluation of Scaffolding procedure ***** Raw water pumps ***

emergency and abnormal operations procedure

Classifying component test for CCW heat exchangers failures Raw water pump AC-1 OC Vendor manual design control information ***

Adequate trending program Electrical supply cable insulation for ccw motors

  • Single worst cast spurious actuations Alternate shutdown capability post-fire safe shutdown procedure*

Safety injection tank *****

7 examples of procedure issues 6/7-2012 ****

Facility staff qualifications

~****

Equipment available to measure river level locally ***

Design bases documents ****

Written evaluations for two changes for flooding mitiqation strateoies **

AFW pumps discharge check valve leakage

  • Deficient evaluation of NRC bulletin 88-04 ***

Degraded/nonconforming condition evaluation and operability determination process*

CCW pump AC-3A "

Safety related pumps and valves up to code Emergency feedwater tank FW-19 ***

Store raw water to emergency feedwater storage tank fill hose****

CCW leakaqe criteria ***"

Auxiliartt feedwater back pressure protection trip ***

Containment air coolers VA-16A and VA-168 ***

Raw water pipe support RWS-117 Raw water piping and piping support calculations*

Pipe supports SIH-17, SIH-94,

and SIH-12 ***

480V and 4160 V buss switchaear cabinets ***

Raw water pump anchor bolt 18 alarm response procedures

  • Emergency operating procedure***'"'*

Air operated valve elastomers Failure to update calculations to account for non-safety related loads suppli~d by the EDG***

Fuel oil in tank F0-1 into applicable design documentation ***

Fuel oil inventory calculation Raw water flow into intake structure ***

Implement maintenance rule 120 Vac system

Digital low resistance ohmmeter values ***

6 instances of failure to identify a deficiency or a condition adverse to quality and enter them into CAP

  • 11 instance of failure to initiate condition reoorts *(**** )

Loss of raw water *****

Failure to identify significant condition 8 Low river level ***

Risk-based operability determinations ****

Operabil'lty determinations that lacked adequate technical justification ****

Intake cell level sluice gate leakaae ***

480 Vac 1B4A bus breakers

  • Structural calculations related to RCS***

Overstressed components****

Turbine driven auxiliary feedwater pump FW 10 ***

Operability determination process****

480 V breakers quality assurance records Switchaear room coolinq ***

Station procedure FCSG-24-4

        • (2)

Fire orotection proqram Sluice aate leakaqe valves ***

Piping in the intake structure raw water vault*

MSPI SSFF degrading trend

  • Removal of motor for raw water punip B on the intake cell level control during a potential site flood ****

Procedure for intake eel\ level control****

Class 1 raw water piping in non-class 1 service building Pipinq and pipe supports ****

Evaluate safety impact of deqraded conditions ****

Raw water oipinq supports **

Containment air cooler pipe supoorts VAS-1 & VAS-2 ***

HPSI iniection valve ***

Failure to request a license amendment for a required chanae **

HPS I pump design and runout***

CS desian chanae **

Operabilitv orocedure ****

Class 1 structures wall thickness defiiciences

  • Raw water electrical pull boxes PB-128T and PB-129T Reactor Vessel Head structural elements ***

Containment internal structure and auxiliarv buildina

PMT procedure ****

Operability determination ****

Model flow path for external flood mitigation ***

Raw water pump 1OC **

DG starting air system_

  • Software classification issues Switchgear room cooling ***

Failure to initiate condition reports for gaps identified in resolving NRC non-cited violations****

Equipment not in state of readiness Raw water strainer AC-128 Refill CCW system **

Gas voiding of ccw

Safety-related pipe stress calculations ***

Evaluate operability of degraded or non-conforming conditions ****

CS system surveillance test Design of F0-10 to F0-1 fuel oil transfer system ***

Temperature limits in auxiliary building Installation of flood barriers caused potential inoperability of auxiliary feedwater system Internal flood analysis*

Adverse design changes ***

Thermal lag analysis ***

Non-conservative values ***

Failed to monitor HCV-2875A Barrier Integrity (Green 2011-2014)

Finding Non-Cited Violation Decontamination work in Containment pray runout

  • spent fuel pool canal *****

HE-2 crane

Containment spray system **"

Reactor coolant pumps 01-RC-9,annunciator response procedure, control element assembly, reactor coolant system high activity **"'"'

Spent fuel pool area charcoal filtration system VA-66 *****

Containment internal structures "'**"'

Containment electrical penetration assemblies ***

Workers failure to follow work control procedures*****

Criterion Symbol Corrective Action

  • 50.59 **

Design control ***

10 CFR part 50 Apendix B, Criterion V ****

Technical Specifications *****

USAR ******

lnltlatinQ Event (Green 2000-2010\

Finding Non-Cited Violation Inadequate Operator Control Socket weld on discharge durinq low power ops pipina for the charainq

  • parameters ****

LCV 1190 condensate control Venting the reactor vessel valve* head ****

Bus bars due to high wind Transient combustion limit in room 59 *****

161 kilovolt power to a safety related busses *****

Inadequate internal flooding procedure from pipe break Inadequate procedure for plant cooldown (2700 gallon steam void in res) *****

Re-pack pressurizer spray valve PCV 103-1 *"***

Fai\ure to perfrom risk assessment in vicinity of T1 transformer Inadequate maintenance on OP-PM-FP-1000 (fire protection system flushing)

Mitigating Systems (Green 2000-2010)

Finding Non-Cited Violation lntrument Air Check Valve IA- Reactor coolant gas system Diesel generators ****

HCV-386-C Fire area 32 ****** Raw water strainer 480 volt Motor Control Center components design basis ***

MCC-3C2 (portable heater loads Low pressure safety injection Inadvertent manual start of

system header diesel driven auxiliary feedwater pump Fire protection sprinklers CCW inlet isolation valves

  • Fuel oil inventory not being Nonlaod shed welding verified receptacles
  • Containment tendon stressing Failure to update surveillance gallery door
  • procedure following valve configuration changes Nonconservative controls of Failure to obtain shift manager containment cleanliness approval *****emergency diesel generator fuel oil inventory *****

Auxiliary feedwater pump **

Emergency diesel generator air starting system air relay valves*

Auxiliary feedwater pumps Diesel driven auxiliary Containment tendon stressing feedwater pump coupling gallery work *****

guard Raw water/CCW heat Containment tendon stressing exchangers AC-1 A, A-1 B gallery work *****

Raw water system Drum heater plugs *****

Diversion of internal flood Nonload shedding electrical water to ECCS pump rooms outlets*

Control room air conditioner Safety related 4 kV bus ground detection circuitry ****

Train of charging pump Frazil ice buildup *****

Diesel generator test procedure****

Diesel generator test procedure *****

Containtment piping penetrations M-9 and M-12 Long-term loss of instrument air RAW water pump AC-10B Long term loss of instrument air***

Inadequate tech spec 2.4 ****

Room 62 and 69 (fire barriers)

Fire protection to features for components important for cold shutdown Fire hose station FP-7G *****

Failure to follow documented instructions *****

Potential compromise of scenario requalification examinations Fire protection program implementation btwn rooms 1 and 58 and rooms 1 and 30 Containment protective coatinas insoection ****

Turbine driven auxiliary feedwater oumo***

Loss of raw water *****

Fire water as a backup for raw water Fire protection orooram

Fire protection program(fire door 1007-10 between fire area 20.1 and fire area 20.4 Station fire plan, revision 61, attachment 14 *****

Decay heat removal loops Alternate shutdown procedure CVCS and HPSI pipinq Raw water pumps *****

Component cooling flow element*

Two high pressure safety injection pumps***

Comoonent isolation valve ***

Raw water strainer component ***

Loss of CCW *****

Turbine driven auxiliary feedwater oum o

strainers*

Emergency diesel generator-1 fuel oil transfer pumps set Points***

Procedure 01-EW-1, Extreme weather revision 13 *****

Component cooling heat exchangers AC-1 A-0 CCW bypass line isolation valve***

Redundant trains of auxiliary feedwater *****

Boron acid leaks

  • Raw water pump AC-10 D packina leakaae
  • Diesel generator control cabinets ***

Boric acid corrosion control procedure ****

Post fire safe shutdown procedure*

Fire water supply system pipina

  • Turbine building concrete floor I<

Emergency diesel generator relavs ****

HPSI header-alternate header isolation valve HCV-2987 *****

Raw water pumps ***

Raw water pump power cables*

Safe shutdown at flood level 1009.3 ****

Cold shut down impossible if river exceeds 1009 feet Repeated tripping of turbine driven auxiliary feedwater pump FW-10*

Turbine driven auxiliary feedwater rJump *****

Turbine driven auxiliary feedwater pump exhaust backpressure trio lever *****

Inadequate desing margin Engineerina chanae 45105***

4160 V and 480V *****

Fuel oil pumo F0-37 ***

Connection btwn cable lugs an cables*****

Apply code ASME section XI code case N-513-2 raw water pipe Operating test given to licensed operators Barrier Integrity (Green 2000-2010)

Finding Non-Cited Violation Failure to check Main hoist Spent fuel handling machine limit switches operator *****

Test backup nitrogen supply systems to ccw inlet and outlet valves to containment air cooling units Failure to certify nondestructive testing personnel Outside contractor properly qualified ****

Transferring fuel in reactor vessel *****

ASME code requirements for Air Accumulators Personal access lock door equalizing valve leakage

  • CCW valves to containment cooling units *"***

CS pumps

  • Auxiliary building crane operating instructions ****

FC 05561, CCW relief valve setpoints, PCS-212 and 413

Criterion Symbol Corrective Action

  • 50,59 **

Desiqn control ***

10 CFR part 50 Apendix B, Criterion V ****

Technical Specifications *****

USAR ******

Inspection Report 2005005 (2/5/06)

Equipment Findings 1R08 in-service inspection NIA N/A activities 1R12Maintenance rule

  • Emergency diesel None implementation generator 2
  • Electrohydraulic control Pump EHC-4B 1R 13Malntenance Risk
  • Raw water pump none Assessment and emergent breaker AC-1 OA work evaluation
  • Raw water.ccw Heat exchanger AC-1 C
  • Air compressor CA-1 A
  • Pressurizer level check surveillance
  • Condensate makeup control valve LCV-1190 backup nitrogen supply bottles
  • Motor driven Fir Pump FP-1A
  • LPS1 piping
  • Raw water pump AC-100 1R 15 Operability evaluations
  • TCV-202, RCS loop Non cited violation for LPSI 2A temperature control (green) valve
  • RM-063, accident range stack gas radiation monitor remote rate meter
  • 161 kV grid voltage
  • LPSI system 1R17 50.59 NIA N/A 1R19 post maintenance tests
  • Replace regulator None gage on IA-HCV-4899-FR, HCV-489 instrument air supply filter/regulator
  • Install diesel generator 2 jacket water filter skid
  • Remove Pl-325-1
  • Replace YS-351

change oil, and obtain oil sample

  • Calibrate Pl-3238 HPSI pump Sl-28 discharge pressure 1R21 CDBI NIA NIA 1R22 Surveillance Testing
  • Channel A safety None injection, containment spray, and recirculation actuation signal test
  • Fire protection system
  • Component cooling valves C & D 1R23 Temporary EC36650, reverse the logic of None modifications reactor vessel flange leakoff indication Pressure Switch PS-139 Inspection Report 2005004 (11114105)

Equipment Findings 1R08 NIA NIA 1R12

  • Engineering safety None features system switch CS-A/LS
  • Safety injection loop none injector valves HCV-331 & HCV 333
  • Charging pump 1C
  • EOG 1 & 2
  • Condensate makeup LCV and PUMP FW-54 (TBD) control valve LCV-1190 backup nitrogen supply
  • Containment cooling and filter unit VA-15A cooling coil VA-1 A
  • Pressurizer RC-4 relief isolation valve HCV-151

pump FW-10 back pressure trip latch FW-64 1R17 N/A N/A 1R19

  • Sl-123(S1-1 B drain) none
  • VA-46B hot gas valve
  • Condenser fan motor for VA-46B
  • AC-100 oumo 1R21 N/A N/A 1R22
  • Channel B safety None injection, Containment spray and recirculation actuation signal test
  • Pressurizer level instrument L-101 X &L-101Y
  • DG1
  • AC-100 Raw water oumo 1R23 Turbine EHC master trip None solenoid valves Inspection Repqrt 2005002 (5/12/05)

Equipment Findings 1R08 N/A N/A 1R12

  • Auxiliary building None ventilating and Air conditioning system condensing units VA-95 &VA-86
  • Intake Structure Sump Pumps, VD-2A & VD-28 1R13
  • EOG 1
  • Spent fuel pool cooling system
  • 345 Kv elevtrical supply lines 1R15
  • Lower portion of for safety-related coatings reactor vessel head
  • Safety injection tanks S1-6A-D
  • Fill/Drain Line Relief Valve to RCDT WD-1
  • Sl-22
  • EOG 1 1R17 NIA N/A 1R19
  • Rod RC-10-23 None
  • Temporary spent fuel pool cooling system
  • Disconnect Switch OS-T1A-1
  • IA-HCV-26038
  • Safety injection tank Sl-6A-D supply inboard isolation valve 1R21 N/A N/A 1R22
  • IA-YCV-1045-C None
  • Oil storage tanks F0-1

& F0-10

  • M-39 and M-53 1R23 Temporary spent fuel pool None cooling system Inspection Report 2005003 (8/5/05)

Equipment Findings 1R08

  • Pressurizer Lower None

girth Weld PRZ-SC 403

  • RPV closure head welds PRVCH-CRD-C0-41 &-41-2
  • Trapeze strut 8-AC-2003/01-PR
  • Circulating water pump None CW-1A
  • Heatless Air Dryer CA-12
  • Air compressor CA-1 B
  • Control Room air condition units
  • Circulating water pumps
  • Safety injection refueling water tank recirculation valves 1R13
  • Component cooling none heat exchanger AC-1 D
  • CCW oulet valve HCV-4828 solenoid
  • CA-7 air compressor
  • Slowdown tank FW-7 transfer pump FW-348
  • LCV-1109 condensate

makeup level control valve

  • EDG2
  • Charging pump CH-1 A
  • Control room ventilation system 1R17 ECCS containment sump NONE screen 1R19
  • Raw water pump AC- none 10C
  • Auto Load Shed Channel A control switch
  • CS geader isolation Valve HCV-344&345
  • Control element assembly RC-10-41
  • Vessel Seal Leakage instrument line waste drain line valve RC-163 1R21 NIA NIA 1R22
  • AFW injection check None valves FW-163 &164
  • RCS leak rate test
  • Auxiliary feedwater system floVI.I transmitters 1R23 Fuel assembly AA06 None Inspection Report 2005009 (1217105)

Equipment Findings 40A2 Inspectors reviewed 183 CR's over 2 year period 1R08 NIA NIA 1R12 NIA None 1R13 NIA none 1R15 NIA

1R17 NIA NIA 1R19 NIA none 1R21 NIA NIA 1R22 NIA None 1R23 NIA None Inspection Report 2005011 (10/14105)

Equipment Findings 1R08 NIA NIA 1R12 NIA None 1R13 NIA none 1R15 NIA 1R17 NIA NIA 1R19 NIA none 1R21

  • Raw water system
  • NCV green for raw
  • ASME code water system and requirements Air ASME code Accumulaots
  • Green finding for
  • Intake structure intake structure 1R22 NIA None 1R23 NIA None Inspection Report 2006002 (5115106)

Equipment Findings 1R08 NIA NIA 1R12

  • None 1R13
  • none 1R15
  • None 1R23 None Inspection Report 2005004 (11/14/05)

Equipment Findings 1R08 NIA N/A 1R12

  • Potable water system 1R13
  • CW-6A bearing water none

cooler

  • FW-3 condensate cooler
  • LPSI jockey pump SI-18 1R15
  • CS Pump Sl-38 Unresolved item with backup bearing cooler CW instrument air outlet flow controller
  • Backup air instrument air to condensate makeup control valve 1R17 LPSI jockey pump none 1R19
  • FW-57 startup aux none feedwater pump suction strainer
  • XSS control switch for CA-1C
  • Air compressor C
  • HCV-2851
  • Component IA-DPT-1039-81 1R21 N/A N/A 1R22
  • Channel A safety None injection, containment spray and recirculation actuation signal test
  • Room 22 safety injection/containment spray pumps and valve
  • Safety injection and refueling water \ank
  • Component cooling category B valve 1R23 Voltage recorder in contro l None room

Inspection Report 2006003 (8/14/06)

Equipment Findings 1R08 N/A N/A 1R12 * 'A' circulating water None cell

  • Coolant charging pump CH-1C 1R13
  • EOG 1& 2 none
  • CCW pump AC-3C breaker 1R15
  • Air supply riser BK
  • Incorrect heat sink calculation
  • Reactor vessel monitoring system 1R17 N/A N/A 1R19
  • Containment purge none isolation valve PCV-742A
  • Auxiliary f eedwater pump
  • Raw water pump AC-1OD backup seal water supply filter AC-220
  • Raw water strainer AC-12B and backwash valve AC-2805B
  • Raw water pump AC-1OB discharge valve 1R21 N/A N/A 1R22
  • Component cooling None valves C & D
  • DG1
  • AC-1 OB raw water pump
  • Fire protection system
  • Reactor manual trip

1R23 USAR (defeating annunciator None card CB-20 A 15 B3)

Inspection Report 2006004 (11/14106)

Equipment Findings 1R08 NIA NIA 1R12

  • Instrument air dryer None failures
  • Fuel oil tank F0-38 level switch LS-2120 1R13
  • Condensate storage none tank
  • Review of licensee's risk assement
  • DG2
  • YCV-8178 DG2 room fresh air supply damper
  • Containment duct relief port 1R17
  • Pressurizer NCV (green) pressurizer
  • Pressurizer heater cable replacement
  • Containment opning replacement SG
  • Reactor vessel head 1R19
  • IA-HCV-28838-FR none
  • SG RC-2A blow-down to blow-down tank FW-7 control valve HCV-1390
  • Spent fuel storage None facility
  • SG RC-28 channel 8

pressure Loop B/P905

  • RCS leak rate test
  • Safety channel C

Equipment Findings 1R08

  • Class 1 welds none
  • Pressurized water reactor vessel upper head penetration
  • SG tube 1R12
  • CS injection valve None HCV-345
  • CCW pump AC-38 1R13
  • Auxiliary FW-54
  • Condenser off-gas radiation monitor replacement RM-057 1R15
  • CS injection valve HCV-345
  • RCS flow intruments tubing separation
  • CH-143 &-155 1R17 N/A N/A 1R19
  • RCS leakage testing none
  • Type C local leak rate test of penetrations M-

39 and 53 CCW AC-3C I None Inspection Report 2007002 (5/15/07)

Equipment Findings 1R08 N/A none 1R12

  • Containment sump None strainer I
  • LPSI pump S1-1A 1R15
  • Containment none ventilation units VA-3 and VA-7
  • Safety related 4160 kV breaker
  • Normal range stack gas radiation monitor remote ratemeter RM-062
  • Filter regulator none assembly
  • SG RC-2A and RC-2B
  • Feedwater pump FW-4C recirculation valve FCV-1151C-20 1R21 N/A N/A 1R22
  • DG2 None
  • Raw water system
  • Fire protection system 1R23 NIA None Inspection Report 2007003 (8/14/07)

Equipment Findings 1R08 N/A none 1R12

  • AC-1 OD raw water none pump
  • HCV-329
  • LPSI to loop 1A and HCV-333
  • LPSI to loop 2B injection valves
  • Switchgear and turbine plant cooling water
  • Replace CW-300&301 raw water pumps AC-1OB and 1OC sparging valves
  • FC-2820 HPS\ pump Sl-28 bearing cooler CCW outlet flow controller
  • FW-54 diesel AFW pump
  • Air compressor CA-1C
  • CA-366 air dryer CA-12 pressure relief valve 1R15
  • HCV-1749 none containment service air header outboard isolation valve
  • Auxiliary feedwat,er pump FW-10 steam traps flow indicator Fl-

1138

  • DG2
  • Reactor protective system 1R17 N/A N/A 1R19
  • 480 V BRKR for AC- none 3C
  • AC-3C
  • QC-10A
  • AC-100
  • HCV1749 1R21 N/A N/A 1R22
  • RCS flow rate None determination
  • Channel B safety injection, CS and recirculation actuation signal test
  • AC-1 OA raw water pump
  • YCV-1045 1R23 NIA None Inspection Report 2007004 ( 11 /13/07)

Equipment Findings 1R08 N/A none 1R12

  • DG 1 Green finding in 2007011 for
  • HCV-151 DG 1
  • DG2 none
  • Discharge valve FW-479
  • HCV-554
  • HCV-329
  • Channel A safety injection, CS and recirculation actuation test
  • DG-1

1R15

  • Ventilation Fan VA-648
  • Ccw system
  • Station battery posts/terminals 1R17 N/A N/A 1R19
  • Emergency response none facility computer system
  • 3CR auxiliary contacts on DG2
  • SA-194 primary starting air pressure regulation valve on DG 2
  • DG 1 1R21 N/A N/A 1R22
  • Raw water system None
  • AC-10B raw water pump
  • Instrument panel Al- None 110 Inspection Report 2007005 (2/11/08)

Equipment Findings 1R08 NIA none 1R12

  • CW-1A main Green NCV for raw water circulating water pump pumps and strainers
  • EOG and support systems
  • Raw water system
  • Containment recirculation
  • Safety electrical distribution system
  • Safety-related structures
  • EDG-2 none
  • Air compressor CA-1 B
  • CS pump Sl-3A 1R15
  • FW-6 and FW-10 NCV green for CCW heat
  • Reactor thermal limits exchangers bypass isolation
  • Underground diesel oil valve storage tank
  • Component cooling heat exchangers AC-1A-D CCW bypass line isolation valve 1R17 DG 1 fuel oil pump molded NCV green for DG 1 fuel oil case circuit breakers transfer pumps 1R19
  • DG-1 none
  • Air compressor CA-1 B
  • CCW heat exchanger inlet valves HCV-489NB
  • Auxiliary f eedwater pump FW-6 1R21 NIA N/A 1R22
  • RCS leak rate test None
  • eves pump check
  • 0-ring seal 1R23 Steam generator-2a None feedwater regulating bypass valve FCV-1105

Inspection Report 2007007 (9/7/07)

Equipment Findings 1R08 N/A none 1R12 N/A None 1R13 N/A none 1R15 N/A none 1R17 NIA N/A 1R19 N/A none 1R21

  • 4160 circuit breakers
  • NCV green for loss of
  • Station batteries - ccw including battery
  • NCV green for safety transfer switches injection pump room
  • Nitrogen admission 21 to component cooling
  • NCV green for raw water surge tank, water system pump pressure control discharge strainers Valve PCV-2610
  • NCV green for turbine
  • Component cooling driven AFW keep water shutdown heat warm line bypass exchanger inlet Valve throttle valves MS-366 HCV-480 & 368
  • Safety injection and
  • NCV green component refueling water storage cooling surge tank tank level indicators nitrogen and
  • Raw water strainers demineralized water
  • Air accumulators - supply line isolation outside containment valves
  • Safety injection pump room ventilation
  • High pressure core injection minimum flow recirculation isolation Valves HCV-385 and -

386.

  • Safety injection refueling water tank discharge Valves HCV 383-1, 2, 3, and 4
  • Safety injection recirculation - throuqh

recirculation sump and safety injection refueling water tank

  • High pressure core injection pump - net positive suction head and sequencing of valve manipulation (including safety injection refueling water tank vortex calculation needed to evaluate pump net positive suction head)
  • High pressure core injection valve, motor-operated Valve HCV-312
  • Raw water and component cooling water - interface

345.

  • Control element assemblies
  • Low pressure core injection - jockey pump
  • Discharge side of component cooling water pressure control Switches 412 and 413 1R22 N/A None 1R23 NIA None Inspection Report 2007010 (10/26/07)

Equipment Findings 40A2 155 Condition Two green NCVs reports reviewed 1R08 N/A none 1R12 N/A None 1R13 N/A none 1R15 N/A none 1R17 N/A N/A 1R19 N/A none 1R21 N/A N/A

I 1R22 N/A I None 1R23 None Inspection Report 2008002 (5/16/08)

Equipment Findings 1R08 N/A none 1R12

  • Instrument Air None compressor CA-1 C 1R13
  • Raw water pump AC-10A
  • Bearing water pump C-9A
  • Raw water pump AC-100
  • Heatless air dryer CA-12
  • Bearing water cooler CW-6A
  • DG auxiliary contacts Green noncited violation for 2-CR raw water pump AC-1 OD
  • Offs\te power low signal relays 27-74/T1A1 and 27-74/T1A2
  • Charging pump suction relief valve CH-180
  • Raw water pumps A/8/C (AC-1 OD) 1R17 NIA N/A 1R19
  • DG2
  • Raw water pumps AC-

10(B&C)

  • Pressurizer quench tank CH-223
  • DG 2 Turbo Lube Oil Circulating Pump LO-40-2 1R21 N/A N/A 1R22
  • Sl-1B LPSI pump discharge isolation (HCV-2938) &

recirculation check valve (Sl-304)

  • Safety related battery chargers
  • DG 1 1R23 N/A None Inspection Report 2008003 (8/12/08)

Equipment Findings 1R08

  • Reactor pressure
  • NCV green for (boric vessel closure head acid)
  • Pipe to elbow weld
  • Elbow to pipe weld
  • Valve to pipe weld and vice versa
  • Hot leg nozzle to safe end welds 22 & 28
  • Cold leg nozzle to safe end welds 24,26,30,32
  • Vessel upper head penetration (VT-2)
  • Raw water pump None packing gland

1R13

  • Refueling outage
  • Raw water pumps 1R15
  • eves none
  • Containment coolers
  • Air coolers VA-3A/B and VA-7A/8 and Dampers VA-56A/B and VA57-A/B 1R17 NIA NIA 1R19
  • FW-54 none
  • Pressurizer
  • HCV-1105
  • Raw water pump AC-100 1R21 NIA NIA 1R22
  • Refueling outage
  • 4160 volt breakers

Equipment Findings 1R08 N/A none 1R12

  • Safety injection water None recirculation tank suction valves LCV-383-112
  • Raw water pump AC-100
  • Instrument air compressor CA 1 B 1R13
  • HPSI pump Sl-2A none
  • CS pump Sl-3A
  • T1A1 and T1A2 transformers
  • Jacket water none temperature control valve JW-116
  • EOG
  • ECCS train A suction piping
  • Fire protection piping 1R1 7 NIA NIA 1R19
  • CS pump none
  • Control room air condition Unit VA-468
  • DG-2
  • FW-54
  • Containment cooling unit VA-88 CCW outlet valve HCV-403C 1R21 NIA NIA 1R22
  • Station batteries EE-8A/B
  • Auxiliary feedwat er pump FW-10
  • Relief valve tailpipe
  • Raw water pump AC-10A 1R23 NIA None Inspection Report 2008005 (2112109)

Equipment Findings 1R08 NIA none 1R12

  • Electro-hydraulic None control system
  • Intake cell A none
  • Inadvertent entry 1R15
  • Sl-1591160 check systems) valves
  • RCS vent to pressurizer quench tank HCV-180
  • Auxiliary f eedwater pump recirculation valve FCV-1369
  • Redundant equipment on opposite trains 1R17 NIA NIA 1R19
  • M-coil replacement in None rod control cabinets
  • Raw water system flow element FE-2890
  • CCW pump AC-3A 1R23 NIA None Inspection Report 2009001 (11113109)

Equipment Findings 1R08 NIA none 1R12

  • DG security None
  • Containment sump outlet strainer Sl- 12A 1R13
  • Condenser FW-1A none hotwell level controller LC-1190
  • $witchyard
  • DG-2 1R15
  • Crane HE-2 Green HE-2 crane
  • FW-10
  • DG-2
  • HCV-509A and B
  • HCV-400A 1R17 N/A N/A 1R19
  • Heat exchanger CCW none inlet valve HCV-481
  • Cooling coil VA-88 CCW inlet valve HCV-403A I
  • Crane HE-2
  • DG-2
  • Component cooling category B
  • Component cooling category A 1R23 N/A None Inspection Report 2009002 (4/30/09)

Equipment Findings 1R08 N/A none 1R12

  • Fire protection and None drainage systems
  • FW-54 none
  • AC-10C
  • Air compressor CA-1 B
  • Bearing water cooling pump CW-68
  • FW-10
  • FW-28 1R15
  • Seismic restraint SIS- none 185
  • HCV-2893
  • FW-10
  • DG 1R17 N/A N/A 1R19
  • HCV-28808 none
  • Raw water outlet valve 1A ccw/raw water exchange
  • FW-10
  • AC-108
  • HCV-2875A
  • FW-54 1R21 N/A N/A 1R22
  • FW-6 valve FCV-1368
  • 13.8 kV emergency power
  • Raw water system 1R23 NIA None Inspection Report 2009003 (8/3/09)

Equipment Findings 1R08 NIA none 1R12

  • DG-2 None
  • FW-10 1R13
  • FW-10 none
  • CS pump Sl-3A
  • DG-1 & 2
  • FW-54 1R15
  • FW-10 none
  • Isolation Valve HCV-8208
  • Radiation Monitor RM050/51
  • Raw water Heat Exchanger AC-1 D
  • FW-6 (auxiliary)
  • Valve MS-291
  • Inverters A & B 1R17 N/A N/A 1R19
  • FW-10 none
  • Valve HCV-2880A
  • HCV-2851
  • HCV-492A
  • Air Compressor CA-1 B 1R21 N/A N/A 1R22
  • FW-10
  • Fuel handling machine interlock
  • Raw water system
  • EOG
  • RCS 1R23 N/A None Inspection Report 2009005 (2/5/10)

Equipment Findings 1R08

  • Vessel upper head penetration
  • Boric acid corrosion control
  • Steam generator 1R12
  • Condensate pump None FW-2B
  • Main generator output breaker 1R13
  • Reactor vessel none
  • Auxiliary building fire header 1R15
  • DG-1 none
  • Charging pumps 1R17 N/A NIA 1R19
  • Containment purge air none

inlet inboard isolation valve PCV-742C

  • Bus tie breaker
  • Pressurizer power operated relief valve PCV-102-2
  • Relay Al-31-TEST-PB-K2 1R21 NIA NIA 1R22
  • Station batteries 1 and 2
  • Raw water system
  • Raw water pump AC-100
  • Penetration M-86 for valve Sl-176 1R23 NIA None Inspection Report 2009006 (12131109)

Equipment Findings 1R08 NIA none 1R12 NIA None 1R13 NIA none 1R15 NIA none 1R17 NIA NIA 1R19 NIA none 1R21

  • EOG 1&2
  • 2 NCV green for intake
  • Containment Fan structure and 1 Cooler 7C & 70 unresolved problem
  • 125 Vdc Station with intake structure Batteries 1 & 2
  • Raw water pump Breaker 1A3-9
  • 161kV transform T1A-

3&-4

  • Power operated relief valve 102-1
  • CS pump 38
  • Intake structure 1R22 N/A None 1R23 N/A None Inspection Report 2010002 (5/12/10)

Equipment Findings 1R08 NIA none 1R12

  • Safety related inverters EE-8H and EE-BJ 1R13
  • West raw water none header outage
  • FW-54
  • Floor plug removal above room 21 1R15
  • Safety injection tank . none S1-6C leakage header
  • Heat exchange CH-7 backpressure control valve PCV-210
  • Power operated relief and pressurizer safety valves tailpipe
  • FW-10
  • Raw ater pump none discharge header isolation valve HCV-2874B
  • B-reactor protective system
  • DG-1
  • CS pump Sl-3A
  • FW-10
  • FW-6
  • HPSI Sl-2B 1R23 NIA None Inspection Report 2010003 (7126110)

Equipment Findings 1R08 NIA none 1R12

  • Pump CW-1A None
  • FW-54
  • Fuel oil transfer pump F0-37 1R13
  • 480 volt ground on bus none 183A
  • CCW isolation valve HCV-2893
  • PCV-102-1&2 none
  • DG-1
  • MCC-3A1
  • AC-3A
  • MCC-3A1 none
  • HCV-317, breaker
  • Al-3-M2
  • CH-10 inlet valve 1R21 N/A N/A 1R22
  • DG-1
  • raw water pump
  • FW-10
  • DG-2
  • Room 22 safety injection/containment spray pumps and valve 1R23 N/A None Inspection Report 2010004 (11/8/10)

Equipment Findings 1R08 N/A none 1R12

  • Heat drain tank level None control valve LCV-1199
  • Breaker 1A31 1R13
  • 345 kV to 161 kV none transformer T3
  • CS pump S1-3A
  • DG-1
  • DG-2
  • Air instrument air compressors 1R15
  • Control room air NCV green for storage tank conditioning unit VA- F0-10 to F0-1 469
  • Room 81
  • F0-10 to F0-1 1R17 6 samples of evaluations; 15 NCV green for pressure samples of changes, tests, control switches PCS-and experiments 412&413 1R19
  • Control room air none condition unit VA-468
  • Air compressor CA-1c
  • HPSI pump Sl-28 discharge valve
  • HCV-2908
  • LPSI pump Sl-1 B discharge valve 1R21 N/A N/A 1R22
  • Radial peaking factor
  • FW-6 1R23 N/A None Inspection Report 2008002 (5/16/08)

Equipment Findings 1R08 N/A none 1R12

  • Condenser off-gas None radiation monitor RM-057
  • FCS cycle 24 maintenance rule 1R13
  • T1 transformer NCV green for T1 transformer
  • CS pump Sl-3B 1R15
  • Main f eedwater pump NCV green for raw water FW-48 suction piping piping
  • F0-1
  • Ccw heat exchanger ac-1 C endbell
  • HCV-321
  • East raw water header 1R17 N/A N/A 1R19
  • HCV-1150A none
  • HCV-344&345
  • Raw water pump AC-10B
  • CS pump Sl-3B 1R21 N/A N/A 1R22
  • SG low pressure trip pressure unit
  • Channel A safety injection
  • Room 22

2000 2001 2002 2003 Findi11~ - NIA 1 Fillcbg - Crffll 4 Non-Cited Viot.tioll - Gn,m 7 Fincling - C,- 5 F!ndin! - NJA 3 Finding - NIA 1 Findir;g - NIA l NOll-Ciud Viollticm - NIA 1 Nm-Cited ViolltiOII - Cam 7 NOii-Cited Violaucm

  • Cum l3 Nco-Cited ViolatiOD- GNa 11 Non-Citecf Violatioft - NIA 2 ViG!a!ioo - l\'tile 1 Violation - SL.I\' 1 2004 2005 2006 2007 Filldillg - Cr.a 3 Finding - Cn,e:n F~-G:n,q 2 1 FmdinJ - l'{JA 1 Fillding - Crem l NM-Cited Violatiaa - GN& 12 NM-Cited Viotatiort
  • Gl'etll 16 NOD-Cif*I Violation - GrH:ll Ill Non-Cited Violation - Gnm 1 "i Yiolatioo. - SL-I\'

Violarioa - Gnu. 1 l Violmoll-~ l v-!OllltiOD

  • Wllife l Violatioa - ~ 2 2008 2009 2010 2011 NOA-Cit!d Violati~ - Gr.a 24 Non-Cited Viol.atiOD - GRl!ll 15 Findillg - Gn,m 1 NOII-Citfd Vio~ - Sirn' .5 Non-Cited Viobtiou - SL-IV 1 F~-Greoi 2 Findir.g - NIA 1 Vi.o!atioll - YeDotr 1 Viot.tioo - Wllilt 2 F'lllding- NIA l N<m-C'tted Violation - Gno 9 Non-Cit!CI Violltico - Gno 17 Non-Ctt<nc promotlhg ~ in. ~e.s:hoting Dynamic Web Page ROP PIM Reports - Event Dates: 01/0112012 - 06/11/2015-Generatcd on 06/17/lS By Types, Cornerstones, Event Dates, Sites Significance: All 203 Open/Closed Final items selected -

Site: 'FCS' Finding. Crten 11 Finding* NIA I Non-Ci1ed Violalion

  • Green 159 Non-Ci1ed Viol111ion
  • NIA 8 Violation - Green 7 Viola1io11
  • NIA 12 Violation
  • Red Violation. Whtie 4 Cross Cutting Areas:
  • SCWE - Safety Conscious Work Environment
  • HP' - Human Performance
  • PIR - Problem Identification and Reso/utwn Finding Mitigating Systems 02/02/2012 FCS *SCWE: N *HP:Y *PIR: N Docket/Status: 05000285 (C)

Open: ML12079A224 (PIM) Failure to Perform Extent of Condition Evaluation The NRC identified a finding for failure of the licensee to follow directions of an apparent cause evaluation to perform an extent of condition evaluation. Specifically, following the identification of an inadequate temporary design modification that rendered annunciator CB 20, Panel A18, Window C3 inoperable on July 5, 2011, engineering personnel failed to perform an extent of condition evaluation to identify other annunciator windows rendered inoperable by the design modification. The failure of engineering personnel to perform an extent of condition evaluation as directed by the apparent cause evaluation for a temporary modification following identification of an unexpected condition was a performance deficiency. The finding is more than minor because the failure to adequately implement the corrective actions associated with the temporary modification s identified deficiencies affects the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because it did not represent a loss of system safety function, did not represent the actual loss of safety function of a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the corrective action program component of the human performance area associated with work practices because engineering personnel failed to follow direction and ensure that an extent of condition review mandated by an apparent cause evaluation was perfonned.

Mitigating Systems 11/ 17/2012 FCS *SCWE: N *HP:Y *PIR: N Docket/Status: 05000285 (C) h1m://c inQ .nrc.ll'nv/nrr-nffice/ms/civn/mn ir 1.cfml 06/ 17120 15 4:10: 17 PMl

uynanuc Web :sate Open: ML12366Al58 (PIM) Failure to Properly Scope All the Pertinent External Flood Protection Features into the Walk.down List in Accordance with Industry Guidance NEI 12-07 The inspectors identified a finding of very low safety significance (Green) for the licensee s failure to generate a complete inspection list, with all the external flood protection features credited in the current licensing basis documents for flooding events, to comply with NRC endorsed NEI 12-07, Guidelines for Performing Walkdowns of Plant Flood Protection Features. These walkdowns were being performed in response to a March 12, 2012, letter from the NRC to licensees, entitled, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident. Specifically, the scoping list did not include several active components, which are an essential part of Fort Calhoun s design basis flood mitigation strategy. The licensee entered the issue into the corrective action program and revised the scoping list accordingly. The performance deficiency was determined to be more than minor because it is associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (Flood Hazard) and it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in addition to not scoping the sluice gates into the Flooding Features Walkdown List, fourteen additional active components would not have been scoped into the walkdown list. This would have prevented the licensee from identifying that preventive maintenance tasks needed to be created, and some active components that are an essential part of the flood mitigating strategy would not have been inspected and tested. The finding was screened as very low safety significance (Green) because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors determined the finding had a cross-cutting aspect in the area of human performance because licensee personnel did not properly apply human error prevention techniques such as peer checking and proper documentation of activities (H.4(a))

Mitigating Systems 12/31/2012 FCS *SCWE:N *HP:N *PIR: Y Docket/Status: 05000285 (C)

Open: ML13045B055 (PIM) Failure to Manage Functionality of the River Sluice Gates The team identified a finding for the failure to manage the functionality of the river sluice gates. Specifically, the licensee s preventive maintenance program requirements were not appropriately implemented for a period of 12 months and as a result, the functionality of the river sluice gates was improperly maintained. The team concluded that the failure to manage the functionality of the sluice gates was a performance deficiency that warranted further evaluation. Specifically, the licensee s preventive maintenance program requirements were not appropriately implemented for a period of 12 months and as a result, the functionality of the sluice gates was improperly maintained. The examples supporting this perfonnance deficiency are as follows: 1) Failure to perform preventive maintenance and monthly testing on the river sluice gates for four months 2) Failure to perform monthly testing on two sluice gates on September 2012 3) Failure to perform monthly testing on all the sluice gates on October 2012 4)

Failure to properly identify and timely enter conditions adverse to quality into the Corrective Action Program 5)

FaiJure to demonstrate effective control of performance of the river sluice gates and to place the system in a monitoring program 6) Failure to make appropriate functionality assessment when the river sluice gates failed the monthly testing during August 2012 The licensee entered these issues into their corrective action program under numerous condition reports described in the body of this report. Using the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined this finding affected the Mitigating Systems cornerstone. The finding is greater than minor because it is associated with both of the Mitigating Systems Cornerstone attributes of Equipment Performance and Protection Against External Factors and, it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to hutiating events to prevent undesirable consequences. The inspectors determined that the fmding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, and conducted a Phase l characterization and initial screening. Using Phase 1 Table 3, SDP Appendix Router, the inspectors answered httn://cio9.nrc.itnv/nrr-nflice/ms/<lvn/mn irl .cfmf06/ 17/20 I:'i 4:10: 17 PMl

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