Text

Summary of 970320 Meeting W/Util in Rockville,Md Re Outage Activities Identified in Attachment Four & Proposed TS Change 194, Reactor Coolant Flow Rate Requirements. W/List of Attendees & Viewgraphs
	ML20137L353
Person / Time
Site:	Maine Yankee
Issue date:	03/27/1997
From:	Dan Dorman; NRC (Affiliation Not Assigned)
To:	; NRC (Affiliation Not Assigned)
References
	NUDOCS 9704070206
	Download: ML20137L353 (37)
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,y UNITED STATES j

j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20066 0001 March 27,1997

. LICENSEE: Maine Yankee Atomic Power Company FACILITY: Maine Yankee Atomic Power Station

SUBJECT:

SUMMARY

OF MARCH 20, 1997. MEETING WITH REPRESENTATIVES OF MAINE YANKEE ATOMIC POWER COMPANY On March 20. 1997, pursuant to notice. the NRC staff met with representatives of Maine Yankee Atomic Power Company (the licensee) at NRC headquarters in Rockville. Maryland. A list of attendees is provided in Attachment 1.

Materials provided by the licensee at the meeting are provided in Attachments 2 through 5.

'During the first part of the meeting. the licensee discussed outage activities identified in Attachment 4.

The activities related to fabrication and loading of new fuel are currently identified as critical path items for a projected restart date in mid-July 1997. Other items were identified as having the potential to affect the critical path including core analysis documentation.

CCW [ component cooling water] cooler bypass trip modifications, atmos)heric steam dump modifications, circulating water pump trip modifications. iELB

[high energy line break] fixes. ECCS [ emergency core cooling system] valves leak. test. EQ [ equipment qualification] submergence modifications, spray pump NPSH [ net positive suction head] improvements, and 115 kV system enhancements.

The staff noted that it is evaluating several of these issues to determine whether they may involve unreviewed safety questions pursuant to 10 CFR 50.59.

h -taff also noted that it will review the licensee's core reload analysis.

1 The lics.nsee provided insights into its implementation of Section V. " Extent of Condit.*

of its March 7.1997. Restart Readiness Plan (reference Accessica A,. 9703130367.) The licensee discussed Attachment 3 to illustrate i

its approach to identification of equipment and process issues.

The organization chart was provided for information.

The individuals identified in the shaded blocks are new in these positions. All but two of these are Maine Yankee employees under the management services agreement between the licensee and Entergy Nuclear.'Inc. The other two are from Yankee Atomic Electric Company and a reassigned Maine Yankee employee.

The Nuclear Quality / Safety Culture diagrams in Attachment 3 identify the barriers used to provide defense-in-depth in the safety culture. The licensee will assess plant systems, programs and human interfaces to identify failures in each of the barriers. The licensee will then evaluate the identified barrier failures i

across all areas to identify trends and/or more fundamental root causes for performance weaknesses.

During the second part of the meeting. the licensee discussed Maine Yankee h

proposed Technical Specification (TS) Change No.194. " Reactor Coolant Flow

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4 Rate Requirements." dated August 15, 1995 (reference Accession No.

9508230024;)' The licensee's discussion generally followei Attachment 2.

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i 9704070206 970327 PDR ADOCK 05000309

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licensee noted that the addition of approximately 230 effective tube plugs per

' steam generator during the current outage would reduce the reactor coolant system (RCS) flow rate below the current minimum requirement in the TS (360.000 gallons per minute (gpm)). The licensee proposed to reduce the l

reactor power limit at a linear rate of 1.5% of rated thermal power (RTP) for i

every 1.0% (of 360.000 gpm) reduction in RCS flow rate below 360.000 gpm to a 3

minimum RCS flow rate of 342.000 gpm at power levels'less than or equal to 92.5% of RTP.

.The licensee indicated that the basis for the TS is to ensure margin to departure from nucleate boiling (DNB). The licensee stated that the DNB i

margin under the proposed conditions is demonstrated by. analysis using approved licensing methods.

The licensee performed a qualitative assessment of the impact of the proposed operating conditions on the small-break loss-of-coolant accident (SBLOCA) analysis. The licensee's assessment concluded that.

since power would always be decreased more than flow, the resulting changes in i

the parameters affecting the SBLOCA analysis would increase the margin to the

-peak cladding temperature limit.

Therefore the licensee concluded that calculations were not necessary to conclude that the SBLOCA analysis results l

are bounded by the proposed TS change.

i The licensee stated that if the number of effective tube plugs per steam 1

generator is increased by more than 580. the plugging will exceed that assumed i

in the design basis analyses including SBLOCA.

If that occurs, the licensee will need an NRC-approved SBLOCA evaluation model to perform the necessary re-analyses. The staff is currently reviewing the licensee's March 10. 1997.

response to-staff questions regarding the use of the'ANF-RELAP model.

(Reference Accession No. 9703170131.) The staff acknowledged that, depending on the results of the steam generator inspections during the current outage.

the licensee may need the staff's approval of the ANF-RELAP model before restart for use in licensing basis analyses of SBLOCA.

(

Daniel H. Dorman. Project Manager Project Directorate I-3 Division of Reactor Projects - I/II

)

Office of Nuclear Reactor Regulation i

Docket No. 50-309

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Attachments:

1.

List of Attendees 2.

RCS Flow Technical Specification Change 3.

Maine Yankee Orga ization (3/17/97) 4, 1997 Refueling 0 age Overview 5.

Maine Yankee Memrrandum dated March 13. 1997 cc w/att:

See next page 4

March 27,1997 I

9508230024.) The licensee's discussion generally followed Attachment 2.

The licensee noted that the addition of approximately 230 effective tube plugs per steam generator during the current outage would reduce the reactor coolant.

system (RCS) flow rate below.the current minimum requirement in the TS (360.000 gallons per minute (gpm)). The licensee proposed to reduce the every 1. power limit at a linear rate of 1.5% of rated thermel power-(RTP) for reactor 0% (of 360.000 gpm) reduction in RCS flow rate below 360.000 gpm to a minimum RCS flow rate of 342.000 gpm at power levels less than or equal to 92.5% of RTP, The licensee-indicated that the basis for the TS is to ensure margin to departure from nucleate boiling (DNB).

The licensee stated that the DNB margin under the proposed conditions is demonstrated by analysis using approved licensing methods. The licensee performed a qualitative assessment of the impact of the proposed operating conditions on the small-break loss-of-coolant accident (SBLOCA) analysis. The licensee's assessment concluded that, since power.would always be decreased more than flow, the resulting changes in the parameters affecting the SBLOCA analysis would increase the margin to the-peak cladding temperature limit. Therefore the licensee concluded that calculations were not necessary to conclude that the SBLOCA analysis results are bounded by the proposed TS change.

The licensee stated that if the number of effective tube plugs per steam generator is increased by more than 580. the plugging will exceed that assumed in the design basis analyses including SBLOCA.

If that occurs, the licensee will need an NRC staff-aparoved SBLOCA evaluation model to perform the necessary re-analyses.

T1e staff is currently reviewing the licensee's March

10. 1997. response to staff questions regarding the use of the ANF-RELAP model.

(Reference Accession No. 9703170131.) The staff acknowledged that, depending on the results of the steam generator inspections during the current outage. the licensee may need the staff's approval of the ANF-RELAP model before restart for use in licensing basis analyses of SBLOCA.

(Original Signed By)?

Daniel H. Dorman Project Manager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-309 Distribution Attachments:

E-MAIL 1.

List of Attendees SCollins JYerokun 2.

RCS Flow Technical Specification Change FMiraglia JLyons 3.

Maine Yankee Organization (3/17/97)

RZimmerman SSun 4.

1997 Refueling Outage Overview SVarga SBrewer 5.

Maine Yankee Memorandum dated March 13, 1997 JZwolinski PMilano HARD COPY cc w/att: See next page DDorman Docket File EPeyton PUBLIC Dross FDI-3 RF RConte OGC DOCUMENT NAME: G:\\DORMAN\\HYMTGSUM.702 CAnderson ACRS T3 eeceive e espy of this document, Ind6cate in the box: 'C' = Copy unthout attachment / enclosure

E' = Copy with attachment / enclosure

h* = No copy i

0FFICE PM:PDI-:L LA:PDIV-2 l (A)D:PDid A DD:DRPE L-m NAME-DDorman,KHPf5 EPeyton n @

PMilan& ffV JZwolinski X, DATE 03/29/97 '

03 M /97 03JF/97 1

03/f/97 0FFICIAL RECORD COPY

t Maine Yankee Atomic Power Station Maine Yankee Atomic Power Company f

cc w/ encl-Mr. Charles B. Brinkman Mr. Robert W. Blackmore Manager - Washington Nuclear Plant Manager Operations Maine Yankee Atomic Power Station ABB Combustion Engineering P.O. Box 408 12300 Twinbrook Parkway, Suite 330 Wiscasset, ME 04578 Rockville, MD 20852 Mr. Michael J. Meisner Thomas G. Dignan, Jr., Esquire Vice-President Ropes & Gray Licensing and Regulatory Compliance One International Place Maine Yankee Atomic Power Company Boston, MA 02110-2624 329 Bath Road Brunswick, ME 04011 Mr. Oldis Vanags State Nuclear 'iafety Advisor Mr. Bruce E. Hinkley, Acting State Planning Office Vice-President, Engineering State House Station #38 Maine Yankee Atomic Power Company 4

Augusta, NE 04333 329 Bath Road Brunswick, ME 04011 Mr. P. L. Anderson, Project Manager Yankee Atomic Electric Company Mr. Patrick J. Dostie 580 Main Street State of Maine Nuclear Safety Bolton, MA 01740-1398 Inspector Maine Yankee Atomic Power Station Regional Administrator, Region I P.O. Box 408 U.S. Nuclear Regulatory Commission Wiscasset, ME 04578 475 Allendale Road King of Prussia, PA 19406 Mr. Graham M. Leitch Vice President, Operations First Selectman of Wiscasset Maine Yankee Atomic Power Station Municipal Building P.O. Box 408 U.S. Route 1 Wiscasset, ME 04578 Wiscasset, ME 04578 Mary Ann Lynch, Esquire Mr. J. T. Yerokun Maine Yankee Atomic Power Company Senior Resident Inspector 329 Bath Road Maine Yankee Atomic Power Station Brunswick, ME 04578 U.S. Nuclear Regulatory Commission P.O. Box E Mr. Jonathan M. Block Wiscasset, ME 04578 Attorney at Law P.O. Box 566 Mr. James R. Hebert, Manager Putney, VT 05346-0566 Nuclear Engineering and Licensing Maine Yankee Atomic Power Company Mr. Michael B. Sellman, President 329 Bath Road Maine Yankee Atomic Power Company Brunswick, ME 04011 329 Bath Road i

Brunswick, ME 04011 Friends of the Coast P.O. Box 98' Edgecomb, ME 04556 1

1 LIST OF ATTENDEES MEETING WITH MAINE YANKEE ATOMIC POWER COMPANY ROCKVILLE. MARYLAND MARCH 20. 1997 E

John Zwolinski, NRR Patrick'Milano. NRR Dan Dorman, NRR Rich Conte Region I Cliff Anderson, Region I Jimi Yerokun. Region I Jim Lyons, NRR Summer Sun, NRR Sarita Brewer, NRR MAINE YANKEE ATOMIC POWER COMPANY Mike Meisner Dan Keuter Howard F. Jones, Jr.

YANKEE ATOMIC ELECTRIC COMPANY J. R. Chapman Michael Scott Bob Harvey STATE OF MAINE Pat Dostie

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Maine Yankee RCS Flow Technical Specification Change E

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a NRC/NRR Meeting

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March 20,1997 i

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i AGENDA e

Introduction Mike Meisner

Background

a Relationship to Other Requests Submittal Status e

RCS Flow Technical Mike Scott Specification Change i

Purpose Basis SBLOCA Discussion e

Siemens SBLOCA Model Mike Scott e

Summary Mike Meisner i

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RCS FLOW TECH SPEC CHANGE Straightforward change request Technically well supported Similar changes and approaches routinely approved for other j

facilities Extending MY review beyond traditional bases to confirm no adverse mechanical effects due to reduced flow e

MY RCS flow change has additional significance One of several elements necessary to responsibly plan for

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Steam generatorinspection results i

Unnecessarily connected to RELAP5YA issue

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STEAM GENERATORINSPECnON 4

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Begins April 1 Additional tube plugging, depending upon extent, could lead to new f

RCS flow rate and additional analyses Prudent planning dictates timely review of pending requests RCS flow Tech Spec change i

Electrosleeving Tech Spec change Siemens SBLOCA methodology i

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i TUBE PLUGGING THRESHOLDS l

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of Plugged Regulatory App 1 oval Needed to Tubes Go Beyond Actual tubes plugged

~250/SG None (4.4%)

l Estimated point at which 480/SG RCS flow; electro-sleeving

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RCS flow is reduced to (8.4%)

i 360,000 gpm Current reload safety 830/SG Siemens SBLOCA methods; RCS analysis assumptions (14.6%)

flow; electro-sleeving

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a RCS FLOW SUBMirrAL STATUS Originally submitted in August,1995 e

e In October,1996, MY requested that the submittal be put on hold due to an apparent connection to RELAP5YA

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In fact, the supporting technical conclusions for the RCS flow submittal e

were drawn independent of any specific model or analysis

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PURPOSE OF TECHNICAL SPECIFICATION CHANGE e

Tech Spec Flow Limit - 1360,000 gpm for Power Operation i

e MY RCS Flow Measurements - 364,000 gpm (w/unc) i e

Minimal Margin to Low Flow Tech Spec f

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s CURRENT TECHNICAL SPECIMCATION LIMITATIONS

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3.10.E.2 Coolant Conditions i

Except for low power physics testing, the reactor coolant flow rate shall be maintained at or more than l

a nominal value of 360,000 gpm during steady-state operation whenever the reactor is critical 4

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2.1.1.d Low Reactor Coolant Flow:

Greater than or equal to 93% of 360,000 GPM 1

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MAINE' YANKEE Allowable Power versus Available Reactor Coolant Flow Rate Nure

' Technical 3.10-3 Specihcation 3.10-12 4-

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E PROPOSED TECHNICAL SPECIFICATION CHANGES I

e 3.10.E.2 Coolant Conditions t

Except for low power physics testing, the reactor coolant flow rate shall be maintained within the limits of Figure 3.10-3 during steady-state operation whenever the reactor is critical. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of entering the restricted region, perform one of the following:

a.

Be in the unrestricted region of Figure 3.10-3 b.

Reduce the allowable PDIL insertion, Excore Symmetric Offset LCO, Symmetric Offset Trip Limit, Thermal Margin / Low Pressure Trip Limit, Maximum Variable Overpower Trip Setpoint, Linear Heat Generation Rate Limit and Excore Symmetric Offset (LOCA l

i Limiting) by 1.5% RTP for each 1.0% of reactor coolant flow.

e 2.1.1.d Low Reactor Coolant Flow:

Greater than or equal to 93% of the power-dependent minimum reactor coolant flow required by Technical Specification 3.10.E.2.

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BASIS OF PROPOSED CHANGE - 1.5% Power /1.0% Flow i

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e Preservation of DNB Margin - Protection of Fuel.

Sensitivity Analysis Performed on SCU/DNBR vs. Flow f

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1.5% Power /1.0% Flow (vs. Sensitivity - 0.8% Power /1.0% Flow)

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i Reduce Trips /LCOs to Preserve Margin to Safety Limit i

e Impact of Power / Flow Reduction on Other FSAR Events Evaluation / Sensitivities Performed on FSAR Events i

Impact - Reduces RCS Temperatures and Sensible Heat l

Increases Margin to RCS/SG Overpressure' Limits (Loss of Load, Loss of Feedwater)

Increases Margin on Containment Peak Pressure, Loss of Feedwater/ Station Blackout (SG Dryout Times) l Reduces Hot Leg Temperature - SG Tube Corrosion

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Preserves Margin on LOCA'

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.v SMALL BREAK LOCA - DISCUSSION OF ISSUE Original SBLOCA Evaluation Based on Engineering Judgement of the ;

o Effects relevant to the SBLOCA event i

e Important Parameters Included:

4 Power Level LHGR Limits

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. Hot Leg Temperature SG Tube Plugging i

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i REVIEW OF RELEVANT PARAMETERS i

i Parameter Impact Power Level Defines Tw (given Tw& RCS flow) l Boiloff rate, RCS pressure response LHGR Limits Rod heatup rate Hot Leg Temperature Controls saturation pressure (RCS pressure response) i SG Tube Plugging SG UA, RCS loop resistance l

Conclusion was that impact of flow reduction Technical Specification o

Change would beinsignificant l

e Code calculations are not necessary to conclude SBLOCA results are bounded by Tech Spec change

PREVIOU.S EVALUATION RCS flow submittal conclusion was not made based on any specific e

RELAP5YA analysis Reference was made to RELAP5YA tube plugging sensitivities to show 1

relative importance 1

Overall evaluation was based on the full power conditions which bound the proposed flow Tech Spec change l

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CURRENT EVALUATION Current SBLOCA evaluation based on CE Cycle 4 results with adjustments to address current operating conditions e

Adjustments included:

Tube Plugging (maximum of 1000/SG)

T,a (increase from 554 F to 560 F) i Power level reduced (2630 MWt to 2440 MWt)

No credit taken for T reduction 1

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Tw for this evaluation bounds flow Tech Spec change

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Future SB LOCA Evaluation 4

Performed by Siemens Power Corporation e

Currently under review by the NRC e

Preliminary sensitivity results support proposed Technical Specification Change i

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REVIEW OF SIEMENS SBLOCA ANALYSIS Siemens model approval needed for future Maine Yankee operation and possibly for restart All outstanding questions have responses submitted e

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How can we facilitate completion of NRC review?

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SUMMARY

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.e Integrated approach to planning for steam generator inspection results e

RCS flow tech spec change Good technical support Consistent with similar submittals Independent of RELAP5YAissue e

Renewed focus on Siemens SBLOCA methodology review

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MAINE YANKEE MEMORANDUM s,

Reliable Electricity for Maine Since 1972 To:

Distribution Date:

March 13,1997 From:

Terry Maxey for File:

RJP-97-030 R. P. Jordan

Subject:

Initial Departmental Review of Programs (Restart Assessment)

- Additional Guidance A listing of programs to be assessed and the format for the assessment were i

distributed by R. P. Jordan on March 8,1997 (File: RPJ-97-27). The Non Hardware Restart Subcommittee is providing the attached information for your use in completing the program assessments. : Expectations for Program Ownership at Maine Yankee : Maine Yankee Program Assessment Guidance Application of this guidance to the program assessments will facilitate consistency and will help program owners focus on potential problem areas.

TM:AMG Distribution:

cc:

Non-Hardwate_ Subcommittee:

Pete Anderson Bruce Hinkley Bob Arsenault Jim Connell Dan Keuter Sean Day John Frothingham Graham Leitch John DeBartolo Bob Fraser Mike Meisner Bob Jordan (Chairperson)

Joe Grant Bill Metevia Steve LeClerc Bob Hayward Brian O'Grady Jim Hebert John McCann Matt Marston Steve Nichols Art Shean Steve Smith Eric Soule Mike Veilleux

f Attcchm:nt 1

f...

l Expectations fsr Program Ownership at Maine Yankee i

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Program ownership is having single point accountability for the success of a program. Although i

there may be a number of individuals and organizations involved with program implementation,-

there is only one owner, j

2 The program owner is expected to understand the program fully, the bases for its existence, the

]

overall objective (s), the technical content and the process hsvolved in meeting the objective (s).

This knowledge base is to be expanded over time. The owrer should ensure that changes to 4

regulations or industry standards which may impact the program are incorporated.' INPO has a j

s

^j series of good practices that apply to many programs, and EPRi ;wblishes a wide spectrum of documents that relate to program technical aspects; the program owner should be aware of put$ cations which address his/her program. Program owners should dso interface with peers in the industry to be knowledgeable of industry trends and to gain the lu:owhxige to perform effective self assessment of the MY program strengths and weaknesses.

The program owner is expected to coordinate the actnnties involved with achieving success. Since

)

programs cut across organizational lines, the owner must be aware of the interaces, and assure that individuals fulfill their respective roles and responsibilities. Furthermore, the program owner should be aware of the interrelationship between other MY programs which may impact the

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success of their program (e.g., the interface between the Preventive Maintenance, Equipment' Qualification, Commercial Grade Dedication and Shelf Life programs).

The program owner is expected to be continually aware of the program status and overall j

performance. Performance indicators should be established for each program, and should be used by the program owner to determine trends and pattems. Upon noting adverse indicators, the program owner is_ expected to take the initiative to develop corrective actions that effectively resolve the downward trend. The relationship of the program owner with program performance is

. analogous to the relationship the system engineer has with assigned systems. It is not enough to have performance indicators; there should be clear evidence that they are used to determine program status and to drive performance improvement.

The program owner should be an effective communicator within the organization, horizontally and vertically. The status of the program should be known to management and others who may be impacted by its performance. Poor program performance should never be a surprise delivered by y

an external organization, e.g., Quality Assurance, INPO, or the NRC.

I in summary, program ownership is not complex; it's a matter of basic stewardship. The program owner has the responsibility to assure that the program successfully fulfills the established i

objective (s). In performing this task, he/she needs to understand the objective (s) and how the i

program works to achieve them, in addition, there is a need to measure how well the program is L

performing, if performing well, the program owner should be looking for enhancements to make l

It better, if performing poorty, the owner should be looking for the underlying causal factors of the j

poor performance and taking the appropriate corrective actions.

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"P Attrchm:i:nt 2 MY Program Assessment Guidance

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r As outlined in the RRP, program assessments will be implemented using a staged approach. The initial phase (what we are doing now) is a screening to ensure that we don't have any unknown issues or problems. Program owners must be the ones to identify any weaknesses or key issues with our technical programs - not other site organizations, QA or extemal agencies. This j

assessment should be completed by the program owner and should not require extensive research.

in some cases, we have previously conducted detailed assessment that may fulfill the requirements-of this assessment. If, however, we identify concoms or significant weakness in the technical

. program, we may need to take a very detailed look at the progrsm to define the extend of the i

concem and to define and implement appropriate corrective actions. Obviously, the expectation is that all programs will eventually acnieve and maintain a recognized level of excellence; however, 4

the first step in meeting that objective is to recognize current deficiences and priontize actions to address any problems uncovered.

in completing f.he program assessment, it is important to provide the appropriate level of information. One word answers are inadequate, and pages of detailed descriptions are 3

i i

unnecessary and not required. The following paragraphs are intended to provide the program owner with sufficient detail to efficiently conduct the program assessment. The conclusions of the

- assessment should be documented in the format of the following sections.

i j

l.'

Program Ownership and Definition j

Consider the following in answering questions A through E.

3 Do we have multiple organizations working on the same program resulting in efficiency and effectiveness weaknesses?

ls it clear who owns the program and what procedures govem implementation?

=

Are input / output requirements from interfacing organizations defined and understood l

such that quality results are produced on a consistent basis?

Are the interfaces between the engineering organizations (MY and YAEC) defined and standards for outputs defined?

Are all program documents readily available on-site to support emergency operations or high priority work?

i 1

1.

Potential Consequences of Plant Restart with Undetected Weaknesses in this Program

{

Consider the following in answering this question.

Why does the program exist? The owner should have a clear understanding of 4

a the overall purper,e of the program and have the ability to articulate that i

understanding in the terms of describing the primary objective or mission of the program.

4 Is the program mandMed by regulation? Is it referenced or found in Technical Spec.htions, USAR or other licensing base documents? If the program isn't based on regulation, what is the basis for its existence?

Does the program have an impact on nuclear safety? Does the program have a regulatory compliance impact? Explain how the program impacts operations, and to the extent practicable, the degree of impact. Those programs with strong ties to plant operation should receive increased attention and scrutiny.

-. _ _. -.---..----._--__~ - --

I, Program Heath cod Effectiveness i,1 Consider the following in answering questions A through C.

4 I

How is the success of the program measured? What data are available? Are there j

formalized performance indicators? Are they consistent with industry standards? Are j

the indicators complete and comprehensive, i.e., are they capable of providing the true picture of program performance? How are these performance indicators being used to j

determine the status of the program? Consider recent trends. Describe any actions taken based on the information and analysis associated with performance indicators?

. What type of feedback loops have been constructed to assure the actions taken are effective?

Has the program has been the subject of an NRC inspection? Has the program been i

i part of a routine NRC resident report or it may have been the subject for a special

inspection? All portment extemal evaluations during the past 24 months should be included, if the NRC or INPO has looked at this program over the past 24 months, what i

were their conclusions? What was done to the program as a result of these findings?

What has changed (positive or negative) regarding the program since the time of the inspection? Do the extemal results correlate with intamal results? If not, can they be l

explained or the differences justified? -

Explain the significance of any NRC violations. MY's response to the NOV is a source of this information along with your understanding of the program and its status at the i

j time of the violation.

Has the program been the subject of an NRC Information Notice, Generic Letter, or Bulletin? Has the program been the subject of INPO SOERs or SERs? Has the i

operating experience review program provided any adverse performance input applicable to this program? Are there any known industry notifications pending on this program?

The number of corrective action program items has a bearing on program performance, but the number alone is insufficient to tell the full story. What are the relative ages of

)

the items? What is the breakdown of categones (significance)? How are actions being tracked and managed? In the aggregate, what does the corrective action process i

indicate about program adequacy?

p

1. List Actions Recommended for Restart or Post-restart and the basis for this Recommendation Actions may include a range of activities including training, procedural improvement, clarification of ownership, developing standards for inputs / outputs, establishing formal i

protocol for clarifying responsibilities between departments, process implementation improvements and other improvement actions. The restart criteria should be used in determining the actions required to be completed prior to restart.

r.

Also describe any major changes (completed / planned) in the program and why those changes were/are being implomonted.

l.

Restart Conclusion Provide your assessment of the overall status of the program. Given all of the information compiled above, how would you rate the current program adequacy. This should be done in terms of the adequacy of the program to support on-going safe and reliable operation of MY, and in terms of the adequacy of the program relative to successfully accomplishing the program's primary objective (s). Provide the recommendation and plan to review your results 1

with the Non-hardware subcommittee.

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ML20137L353

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