ML20135C441

Summarizes 961031 Meeting W/Florida Power Corp to Discuss Corrective Actions to Address Weaknesses in Engineering Performance.List of Attendees & Handouts Encl

ML20135C441
Person / Time
Site:	Crystal River
Issue date:	11/06/1996
From:	Landis K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Beard P FLORIDA POWER CORP.
References
NUDOCS 9612060306
Download: ML20135C441 (4)

Text

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i November 6, 1996 s

i i Florida Power Corporation Crystal River Energy Complex 4 Mr. P. M. Beard, Jr. (SA2A) i Sr. VP, Nuclear Operations ATTN: Mgr.. Nuclear Licensing 15760 West Power Line Street Crystal River. FL 34428-6708

SUBJECT:

MEETING

SUMMARY

ENGINEERING PERFORMANCE MEETING

CRYSTAL RIVER - DOCKET NO. 50-302

Dear Mr. Beard:

4 This refers to the meeting on October 31, 1996, at your Nuclear Administration Building (NAB) Conference Room 101. The purpose of the meeting was to discuss

your corrective actions to address weaknesses in engineering performance. It j is our opinion, that this meeting was beneficial.

1 Enclosed is a List of Attendees and Florida Power Corporation Handout. The

discussions included the following topics

Root Contributers to Engineering

! Performance. Corrective Actions. Measures of Effectiveness, and Outage Scope.

In accordance with Section 2.790 of NRC's " Rules of Practice "Part 2.

Title 10 Code of Federal Regulations, a co)y of this letter and its enclosures 1 will be placed in the NRC Public Document Room.

a

Should you have any questions concerning this letter, please contact us.

] Sincerely.

Orig signed by Kerry D. Landis Kerry D. Landis, Chief Reactor Projects Branch 3 Division of Reactor Projects Docket No. 50-302 License Nos. DPR-72

Enclosures:

1. List of Attendees

2. FPC Handout cc w/encls: Gary L. Boldt. Vice President Nuclear Production (SA2C)

Florida Power Corporation Crystal River Energy Complex  !

15760 West Power Line Street  !

Crystal River. FL 34428-6708 l cc w/encls: Continued see page 2 060071 o m c m , copy 9612060306 961106 k l

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FPC 2 cc w/encls: Continued B. J. Hickle. Director Nuclear Plant Operations (NA2C)

Florida Power Corporation Crystal River Energy Complex 15760 West Power Line Street Crystal River. FL 34428-6708 L. C. Kelley Director (SA2A)

Nuclear Operations Site Support Florida Power Corporation Crystal River Energy Complex 15760 West Power Line Street Crystal River. FL 34428-6708 R. Alexander Glenn Corporate Counsel Florida Power Corporation MAC - ASA P. O. Box 14042 St. Petersburg FL 33733 Attorney General Department of Legal Affairs The Capitol Tallahassee FL 32304 Bill Passetti Office of Radiation Control De)artment of Health and lehabilitative Services 1317 Winewood Boulevard Tallahassee. FL 32399-0700 Joe Myers. Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee. FL 32399-2100 Chairman Board of County Commissioners Citrus County i 110 N. Apopka Avenue Inverness. FL 34450-4245 Robert B. Borsum B&W Nuclear Technologies i 1700 Rockville Pike. Suite 525 Rockville. MD 20852-1631

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LIST OF ATTENDEES Florida Power Corocration P. Beard. Senior Vice President. Nuclear Operations G. Boldt. Vice President. Nuclear Production '

B. Hickle. Director. Nuclear Plant Operations L. Kelly. Director. Nuclear Operations Site Support F. Sullivan Manager. Nuclear Operations Engineering G. Halnon. Assistant Director. Nuclear Operations Site Support J. Baumstark. Director. Quality Programs J. Terry, Manager. Nuclear Plant Technical Support Nuclear Reculatory Commission R. Butcher. Senior Resident Inspector. Crystal River S. Cahill. Resident Inspector. Watts Bar T. Cooper. Resident Inspector. Crystal River S. Ebneter. Regional Administrator A. Gibson. Director. Division Reactor Safety F. Hebdon. Director II-3. Office Of Nuclear Reactor Regulations (NRR)

J. Jaudon. Deputy Director. Division Reactor Safety J. Johnson. Deputy Director. Divsion Reactor Projects K. Landis Chief. Branch 3. Division of Reactor Projects L. Raghaven. Project Manager. Project Directorate II-1. NRR Members of the News Media Enclosure 1

FPC 3 Distribution w/ encl:

L. Raghavan, NRR -

B. Crowley, RII G. Hallstrom, RII PUBLIC NRC Resident Inspector U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, FL 34428 Orritt s!GNATURE 5, NAME LMellen DATE 11 / fa / 96 11 / / 96 11 / / 96 11 / / 96 11 / / 96 11 / / 96 com A no vcs wo vos e vEs e vts no vcs No Of flCIAt htwRD COPr 00ahtNi NAMt: 6:\CR5uM7.99

i Florida Power CORPORATION EO October 28, 1996 4

3F1096-22 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Subject:

Crystal River Unit 3 Forced Outage

Dear Sir:

On September 2,1996, Florida Power Corporation (FPC) shut down the Crystal River Unit 3 (CR-3) nuclear plant due to a leak in the turbine lube oil system. During this forced outage, FPC determined that a modification had been made to the plant during the Spring, 1996 Refuel 10 outage which created an Unreviewed Safety Question (USQ) regarding emergency diesel generator (EDG) loading. This USQ involved a reduction in the margin of safety described in portions of the Technical Specification Bases.

On October 4,1996, while still shut down, FPC was preparing a submittal to request NRC approval of a license amendment to change the affected EDG Technical Specification Bases when additional questions arose regarding the change to the '

emergency feedwater (EFW) system which created the diesel loading USQ. These questions involved failure modes with the EFW system which needed to be evaluated to ensure the system could perform its safety function and reliance on the turbine-driven, "B" train emergency feedwater pump for "A" train EDG load management. Due to the EFW/EDG issues, and some other design-related issues, FPC management made a decision to keep CR-3 shut down until these issues are adequately addressed. The purpose of this letter is to inform the NRC of our plans to address these issues prior to restarting the plant.

CRYSTAL RNER ENERGY COMPLEX: 1576o W. Power Line St . Crystal River, Florida 34428-6708 . (352) 795-6486 A Florida Progress Company swr $3te*4t v fy.

U. S. Nuclear Regulatory Commission 3F1096-22 Page 2 of 7 The issues described in the attached list were identified through a review conducted by a multi-discipline team involved in reviewing the Emergency Operating Procedures (EOPs) and through design reviews by the engineering organization. The list was reviewed by CR-3 senior management and the items are considered necessary to ensure safety system operability or to increase design margins. Each issue has been documented in the CR-3 corrective action system and will be tracked to closure. Several of the issues have been determined to be reportable and Licensee Event Reports are being processed.

FPC will ensure the safety systems in question are capable of performing their design basis functions prior to restart from this outage. As an added level of assurance, FPC will be establishing an internal restart panel which will function similar to an NRC restart panel using NRC Inspection Manual 0350 as a guideline for conducting the restart readiness review. Upon completion of the work to resolve the issues, the panel will conduct a final review to confirm that all issues have been resolved adequately. When satisfied, restart of the unit will be recommended to the Senior Vice President, Nuclear Operations. In addition, the Nuclear General Review Committee (NGRC) will conduct an independent review prior to restart.

Project teams or individual lead responsibility have been established for each issue to support the design, licensing and installation activities necessary to complete the outage work scope. Final resolutions for some of the issues on the list have not yet been determined. Other resolutions require relatively long lead procurement activities. Therefore, an integrated outage schedule is not available at this time. However, we expect the unit to remain shutdown until at least mid-January, 1997. This will also likely move our next refueling outage, Refuel 11, to the fall of 1998 rather than the spring of 1998, as currently scheduled. The NRC will be kept abreast of the schedule and progress on these issues as the outage continues.

Sincerely,

. M. Beard, Jr.

Senior Vice President Nuclear Operations PMB/BG Attachment xc: Regional Administrator, Region II Senior Resident Inspector NRR Project Manager

U. S. Nuclear Regulatory Commission 1 l

3F1096-22 Attachment Page 3 of 7 1 l

CR-3 Design Margin Improvement Outage Scope of Work

1. Hiah Pressure Injection (HPI) Pump Recirculation to the Makeup Tank Concern: The HPI pumps draw suction from the Borated Water Storage Tank (BWST) during the initial phase of emergency core cooling system ,

(ECCS) injection. Once BWST level has reached a pre-determined i level, suction is switched to the reactor building sump with the HPI )

pumps taking suction from the discharge of the low pressure injection (LPI) pumps (piggyback operation). During piggyback operation, LPI pump discharge pressure keeps the check valve in the suction line from the makeup tank (MUT) to the HPI pumps closed (MUV-65). During long term small break LOCA (SBLOCA) cooling, HPI 1 flow may require throttling due to lower required ECCS flow. If i throttling continues, procedures will eventually direct the operators to increase total HPI pump flow by opening the HPI recirculation valves at a pre-determined flow rate to divert some  !

flow to the MUT. Since no flow is exiting the MUT, the tank could  ;

fill up with recirculation flow and lift the relief valves, dumping ,

fluid onto the auxiliary building floor. This would result in the transfer of RB sump fluid to the auxiliary building sump, which reduces the amount of water available in the RB sump from which the LPI and reactor building spray pumps take suction during the later stages of core and containment cooling. This could also create a l release path for post accident radioactive fluid outside containment.

Resolution: FPC is consulting with Framatome Technologies, Inc. (FTI) to confirm whether the scenario is valid and within the CR-3 design basis.

Although the resolution of this issue is still undetermined at this time, preliminary indications are that opening of a high point vent valve may preclude the need to open the HPI recirculation valves in the SBLOCA scenarios of concern.

Schedule: This issue will be resolved prior to startup from the current outage. -

2. HPI System Modifications to Improve SBLOCA Marains Concern: The CR-3 HPI-system currently meets all design and licensing basis functional requirements. However, the CR-3 configuration is not consistent with the designs at other Babcock and Wilcox (B&W) plants. As a result, HPI minimum and maximum flow limits are more restrictive and peak cladding temperatures for certain SBLOCA scenarios are higher. In addition, the reduced system design margin has created the need for several manual operator actions to ensure adequate core cooling. FPC intends to reduce the operator burden created by these actions and the system margin deficit through hardware modifications. These modifications would also make the CR-3 HPI system design more like other B&W plants.

U. S. Nuclear Regulatory Commission 3F1096-22 Attachment Page 4 of 7 Resolution: At this time, the following modifications are being considered:

a. Installing cavitating venturis to limit flow through any .

single injection leg due to a postulated break in that leg.  !

b. Installing cross-tie piping downstream of the HPI injection control valves to deliver increased core cooling flow should a failure prevent one or more of the injection valves from opening.

c. Modifying the normal makeup line to ensure automatic isolation occurs upon ES actuation to eliminate the operator action now required to perform this function. This involves modifying the power supply to the existing isolation valve (MUV-27) and

• adding another isolation valve powered from the opposite train in series with MVV-27. (Note: the proposed installation of the cavitating venturis could preclude the need for this I modification).

Schedule: Since the HPI system is fully capable of meeting its design function, these modifications are not considered necessary to complete during the current outage. However, FPC is developing the design packages and determining whether equipment can be procured in a time frame to install in the current outage given the schedules for other activities.

3. LPI Pump Mission Time Concern: During the IPAP inspection, an issue was raised regarding the need to establish flow through the decay heat removal (DH) drop line to j the decay heat removal (LPI) pumps as part of small break LOCA CR-3 has two redundant, independent LPI trains which I mitigation.

I can take suction from the RB sump during long term recirculation core cooling. However, certain small break LOCAs could result in '

l long-lasting, elevated RCS pressures such that the LPI pumps would have to operate in the piggyback mode at low flow rates for an extended period of time. As that period of time approaches the '

current low flow mission time for the LPI pumps, plant procedures direct the operators to trip one pump and open the DH drop line valves to the RB sump to provide additional flow through the remaining running LPI pump. There is only one DH drop line at CR-3 i (and many other pressurized water reactors) which has three motor- l operated valves in series. Failure of any one of the drop line j valves to open would prevent flow through the line. If the DH drop  ;

line was necessary to ful fill the ECCS long term core cooling function for small break LOCA mitigation, this would violate the single failure design criterion.

Resolution: The concern described above is time-dependent. If the time frame is long enough after the event, opening of the DH drop iine could be  ;

considered a long-term recovery action as opposed to an emergency j core cooling function. FPC considers the long term recovery phase I beyond the time frame implied by the regulations where applying the I single failure design criterien is necessary. At the time of the l 1

U. 5. Nuclear Regulatory Commission 3F1096-22 l Attachment

! Page 5 of 7 IPAP inspection, the low flow mission time for the LPI pumps was 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which was questionable from an ECCS versus long term recovery perspective. FPC is currently low-flow testing a pump which is  :

identical to the CR-3 LPI pumps. The test flow rate is approximately 100 gallons per minute (gpm). The design flow rate of the LPI pumps is 3000 gpm. The results of this test are expected to prove that the pumps could run for an extended period at very low ,

flows without damage. If the test is successful, procedures will be  !

revised to characterize opening the DH drop line in this scenario as a long term recovery action rather than an ECCS function.

Schedule: This issue will be resolved before startup from the current outage.

As of 3:30 p.m. on October 25, 1996, the pump had completed 18 days of continuous low-flow testing with no performance (head curve) degradation, no mechanical seal leakage, no indication of unexpected  !

bearing wear, and all vibration parameters stable and well below the +

action levels specified in the surveillance procedure. The testing is continuing beyond 18 days.

4. Reactor Buildina SoraY Pumo IB NPSH Concern: During the long term recirculation phase of core and containment  ;

cooling, the reactor building spray pumps (BSPs) take suction from '

the reactor building sump. Calculations have shown BSP-IS to have i little margin between required and available net positive suction  !

head (NPSH) during this phase of operation. A recent revision of i the calculation shows the margin to be approximately one foot of l water. It is desired to increase this margin.

Resolution: FPC currently plans to conduct factory testing and/or modify the l pump impeller to improve the margin between required and available NPSH. i Schedule: This issue will be resolved before startup from the current outage.

5. EmeroencY Feedwater System Uporades and Diesel Generator load impact. '

Concern 5.1: The CR-3 EFW system is comprised of two 100% capacity trains, with the "A" train pump (EFP-1) being motor driven and the "B" train pump (EFP-2) being steam driven. The steam for the EFP-l l

2 turbine driver is fed through redundant inlet valves (ASV-5 I l and ASV-204) to ensure the availability of steam given a failure of one of the inlet valves to open. Each pump feeds both steam generators. For a portion of the flow path from  :

the emergency feedwater tank (EFT-2), the two pumps share a common suction line. Under certain accident scenarios, there are failure modes which can cause the calculated NPSH available to both pumps to be less than required. For example, a failure of the DC control power source for the ,

injection control valves in one train of EFW can result in the i pump in that train producing high flows which result in excessive friction head losses through the common suction line.

V. S. Nuclear Regulatory Commission 3F1096-22 Attachment l Page 6 of 7 i

e concern 5.2: Motor-driven EFP-1 is powered from the "A" train ES bus and is connected to the "A" emergency diesel generator (EGDG-1A).

EFP-2 is steam driven and therefore does not affect "B" train  ;

EDG loading. However, portions of the load management scheme .

4 for EGDG-1A depend on the availability of EFP-2 to: 1) limit i the total flow produced by EFP-1 during the early stages of diesel loading and 2) permit EFP-1 to be shut down and the "A" train LPI pump (and other engineered safeguards features) to j be started in the later stages of accident mitigation.

Therefore, some postulated failure modes which cause EFP-2 to  :

, be unavailable invalidate assumptions made in EGOG-1A loading calculations and some accident analyses which may have taken credit for flow from EFP-2 after EFP-1 was shut down.

Resolution: At this time, the following modifications are being' considered:

a. Installing cavitating venturis in the EFW pump discharge lines to limit flow during the postulated failures which result in the loss of flow control for an EFW train. This will eliminate the NPSH concern.

b. Re-enabling "A" train Emergency Feedwater Initiation and Control (EFIC) system actuation of EFP-2 via automatically '

opening steam turbine inlet valve ASV-204. This feature was disabled by a modification in Refuel 10 and will be restored to ensure EFP-2 auto-starts given a failure of the "B" side initiate logic or ASV-5.

c. Installing motor operators on cross-tie valves EFV-12 and EFV-13 to allow remote manual opening of these valves. Opening these valves establishes a flow path allowing the pump from one train to feed the steam generators through the injection lines of the other train. This is desirable to ensure the operators can maintain EFW flow control and indication in certain single failure scenarios without requiring local manual valve operation.

Schedule: This issue will be resolved before startup from the current outage.

• We expect this issue to require additional interaction with the NRC prior to restart.

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6. Emergency Diesel Generator loadina l

Concern: The rated capacity of EGDG-1A is challenged by the continuous, automatically connected loads as well as the loads that are manually connected in the later stages of accident mitigation. Three  ;

concerns were created by the Refuel 10 modification which removed the "A" train EFIC automatic actuation of ASV-204. Calculated peak transient diesel loads were above the 3500 kW maximum engine rating documented in the FSAR and the ITS basis background for LCO 3.8.1, "AC Sources"; calculated peak diesel load at one minute was above the 3100 kW rating discussed in the basis for Surveillance Requirement 3.8.1.11; and the highest single rejected diesel load

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U. S. Nuclear Regulatory Commission 3F1096-22 Attachment l Page 7 of 7 discussed in the basis for Surveillance Requirement 3.8.1.8 increased.

Resolution: A combination of three efforts is being pursued to increase the load r

capability of EGDG-1A. They include an engine power upgrade to increase one or more of the load ratings; removal and/or reduction of connected loads; and improving the accuracy of the kW meters used to display the generators' output. We expect this issue to also require additional NRC interaction prior to restart.

Schedule: This issue will be resolved before startup from the current outage.

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7. Failure Modes and Effects of Loss of DC Power Concern: A number of CR-3 design and operating vulnerabilities have been identified on a case-by-case basis through design and E0P reviews postulating the effects of a loss of DC power. The loss of DC power also causes a consequential loss of emergency AC power since DC power is required for emergency diesel generator field flashing and bus breaker closure. A Failure Modes and Effects Analysis (FMEA) was performed for the CR-3 Class IE electrical distribution system (including DC) as part of the original plant design. However, it may not have fully considered system interactions, including effects on redundant trains and components.

Resolution: FPC will perform a DC power FMEA which includes evaluations of system interactions.

i Schedule: The FMEA review will be completed to the extent that FPC is satisfied that we have identified any safety significant problems.

Such problems will be addressed prior to startup from the current 4

outage.

I 8. Generic letter 96-06

! Concern: This Generic Letter (GL) identifies three issues regarding the '

effect of post-accident containment heatup on containment coolers,

. piping, and penetrations. CR-3 is susceptible to the piping

? overpressurization phenomenon and is evaluating the water hammer and two-phase heat transfer problems.

Resolution: FPC is installing thermal overpressure protection devices on containment penetrations affected by this phenomenon. Actions to j address the impact of the other two issues, if any, will be determined after the review is completed.

Schedule: The overpressure protection devices will be installed prior to startup from the current outage. Actions to address the impact of the other two issues, if any, will be scheduled according to the safety significance of the findings.

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10/29/96 Rev. O CRYSTAL RIVER UNIT 3 l RESTART PLAN i Objectives e Achieve Significant In-Plant Safety Margin Improvement.  !

e Improve Materiel Condition of the Plant with Emphasis on the Secondary Side. l l e Accomplish all Restart Milestones on Time. l e Have No OSHA Reportable Injuries. j e Have No Violations Resulting in Escalated Enforcement.

e Perform Restart Event Free.

l e Achieve Prompt Sdf Disclosures of Problems.

1 e Successfully Implement the new Procedure Change Process and Corrective Action System.

Mananement Exnectations Safs e The plaat will be maintained in a safe condition at all times. All work will be performed using defense, in-depth strategies.

e All design deficiencies that impact safety will be addressed to assure adequate safety margins exist before restart. -

l e Procedures will be followed exactly as written or work will be stopped and the procedure corrected.

i e Strict attention and adherence to CR-3 industrial safety procedures and the Accident Prevention Manual will be maintained at all times. These safety standards will be strictly enforced.

i e- The plant will not be returned to power until scheduled work is complete and an independent restart safety review is conducted which clearly demonstrates that )

safety margins are acceptable for continued operation.

I e Radiation doses will be controlled as low as reasonably achievable. l I

1Agal l e All regulatory requirements and legal commitments will be fully met. l l

l e All interactions with our regulators will be timely, candid, and thorough.

o Regulatory concerns will be promptly communicated within the organization and addressed thoroughly.

l e All work step verifications will be performed without omission.

j Efficient e The schedule will control all work performed during the outage.

l e Plant equipment problems will be corrected so as to minimize operator burden l and ensure reliable operation until Refuel 11.

o Work problems will be communicated as soon as possible to the Nuclear Shift Manager.

e Workmanship will be of the highest quality achievable.

e All required training will be conducted on plant modifications and associated j l

procedure changes prior to power operations.

Organization / Responsibilities l

See Attachments I & II. -

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Conununications H I

During the restart of CR-3, communications with employees will be enhanced by: l i

i e Ali Hands Kick-Off Meetings. These meetings will ensure that all employees are

presented with the Restart Plan and a' ford each employee with the opportunity to

have questions / concerns addressed.

T +r

e CR-3 Internal News Bulletins will be issued to ensure employees are made aware l

of the current status of restart activities.

e Restart objectives will be displayed on posters throughout the plant to help ensure l

employees remain focused on the objectives.

Ontase Scone & Schedule See Attachment III for summary of outage scope. The outage schedule is currently under development. The outage target completion date (plant in Mode 1) is February 28,1997.

Logistics Meetings:

e An 0615 along with a 1615 Daily Schedule Coordination Meeting will be held to coordinate emergent activities and confirm schedule direction.

e An 0800 Plant Manager's Safety Review Meeting will be held to review plant status, ensure safety focus, and oversee restart objectives.

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e An 0830 Senior Nuclear Officer's Meeting will continue to be held to brief the ,

Senior Nuclear Officer on Nuclear Operation's status and establish coordinated l priorities for resolution of important issues. I o A Restart Command Center has been established in the Nuclear Administration Building Conference Room 203 for making key restart decisions. I Work Schedules:

e Normal work schedules will consist of two 10-hour shifts 5 days per week.

e Critical Path work will be scheduled for two 10-hour shifts 6 days per week.

e Sundays will not normally be used for catchup. m Budget e The budget will be developed after design details are finalized.

i e All expenditures related to this outage are to be charged to accounting number 657000.

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Oversleht Activities l l

Restart Panel NOTE: Attachment IV depicts the reporting chain of the Restart Approval Authority.

A Restart Team has been initiated. Membership includes:

l Regulatory Checklist Coordinator Containment Checklist Coordinator  ;

Programmatic Checklist Coordinator . '

Secondary Plant Checklist Coordinator Surveillance Test Checklist Coordinator ECCS Design Checklist Coordinator The Restart Team will report to a Restart Panel. Membership includes:

Director of Nuclear Plant Operations Director of Quality Programs Outage Manager Independent Panel Member The Restart Panel Chairman is the Vice President Nuclear Production.

The Restart Panel will report to the Restart Authority, Senior Vice President Nuclear Operations who will provide restart authorization.

Plant Review Committee Plant Review Committee is responsible for:

o maintaining oversight of the restart restraim check list, e reviewing 50.59 evaluations associated with outage work, and e monitoring AI-256, Outage Restart Readiness Guidelines.

Nuclear General Review Committee The NGRC provides broad safety oversight of the restart effort and provides recommendations to the Senior Vice President Nuclear Operations.

Technical Issues See Attachment V.

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Attachment I Senior Vice President Nuclear Operations (SeniorNuclearOfficer)

Director Nuclear Plant Operations Director Quality Programs -

, o Independent Restart Oversight (Restart Director) o Safe Unit Restart o OutageObjectives Accomplishment ggg gg Director Nuclear Engineering & Projects ,

Operations Materials (VPNP)

& Controls ***' "*"i 8h' o Engineering o Outage Budget Preparation o MaterialPr9curement o Outage Contract Administration ,

Director Nuclear Director Nuclear Operations Site Support Operations Training o Licensiassupport o In-Processing i o MARTrammg l

Director Nuclear Plant Attachment H Operations Assistant Plant Director (Hickle) Assistant Plant Director Restart Director Maintenance &

Operations & Chemistry Radiation Protection (Davis) o Plant safety (Campbell) n o Operations / Chemistry

, o Industrialsafety j o QualityMaint1 Construction /HP g Outage Manager (Koon) o schedule Accuracy & Execution o MilestoneMonitoring Materials (GAC) Projects (KFL)

NSMs o safescheduleExecution PRC Chairman (Halnon) o startUpRestraints o 50.59 Reviews Communications Officer o RestartReadinessReview (Kurtz) o strategicCommunications Licensing Support Engineering Support (Gutherman)

(Sullivan)(Terry) o NRC Interface Coordination &

o Design Development o systemEngineeringsupport i

Attachment til OUTAGE SCOPE

SUMMARY

l l

l Technical Projects as defined in Attachment V.

OTHER MAJOR PROJECTS:

R.B. Ladders i

R.B. Coatings R.W. Spool Re,lacement MAJOR MAINTENANCE ITEMS:

RCP Motor Oil Leaks Polar Crane Work Items FHCR-1 Rebuild Fuel Hoist and Reinstall Over Due On-Line Preventative Maintenance (CS's) (24)

Over Due Outage Preventative Maintenance (CS's) (36) '

Over Due On-Line Calibrations (IC's) (20)

Over Due Outage Calibrations (IC's) (6)

Upcoming Preventative Mainicnance through May 97 (CS's) (278)

Upcoming Calibrations (IC's) thrcush May 97 (135) i Outage Related Preventative Mali:tenance (CS's) through October 97 (128)

Outage Related Calibrations (IC's) through October 97 (36)

System Engineering Ranking Lists Control Board Deficiency Tags (36)

Operator Work Around items (8)

Surveillance Procedures (SP's) Normal and Outage Related (Evaluating those due 2:

18 Months s 24 Months)

Fuse Control Program Retorque OTSG Primary & Secondary Manways & Handholes Multi-Case Circuit Breaker Changeout RWP-2A and Related items CHHE-1 A and Related items l

MSSV'S (6) Send off for Rebuild & Test .

R.B. Jib Crane Seal Oil System Work i Cl System Outage

l A"ehment IV Crystal River Unit 3 Restart Approval l

RestartAuthority SVPNO I

h RestartPanel Chairman VPNP I

V V V V Restart PanelMember Restart PanelMember OutageManager RestartPanelMember DNPO DQP Independent l

V Restart Team l

Regulatory hogrammade Containment Checklist Checklist Checklist Coordinator Coordinator Coordinator l

b Surveillance Secondary ECCS Design TestChecklist PlantChecklist Checklist Coordinator Coordinator Coordinator

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.- i Attachment V j TECHNICAL ISSUES Item Descriotion Comments I

l HPI Pump Recire Back to the MUT 1) Possible Procedure Fix, or

2) Possible Piping Fix.

HPI Modifications to Improve SBLOCA Evaluations Are On-going.

Margins Mission Time of LPI Pumps.- Evaluations Are On-going, i Improve BSP-1B NPSH Rebuild BSP-1B and Possibly BSP-1A.

EFW/EFIC Upgrade (Cavitating Venturis) A MUST for this Outage.

l Comp Module A MUST for this Outage.

ASV-204 A MUST for this Outage.

l EDG Lead Capability Rebuild Turbochargers and Modify Intercoolers, Evaluate I. mad Removals, and Evaluate Accuracy of KW Meters.

Complete FMEA for Loss of DC Power Evaluation Will Resolve Problems Raised via the Scenario, i

Assess Impact of GL 96-06 Approximately 13 Penetrations will be ,

Modified to Prevent Overpressurization.

Perform NPSH Review of EFP Suction Evaluation may ' Identify Worst Case as Swapover from EFT to CST Installing Two Check Valves, j i

l Identify and Complete Graded Setpoint 8 EFIC Hi-Range Transmitters to be Cale's Needed to Evaluate New Transmitters Installed. .

Resolve NPSH Concern when SFPs are used This Can be Accomplished with Procedure to Recirc the BWST Changes.

Address any Current Concerns with GL 96- Evaluation will Address any Concerns.

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l Item Descriotion Comments i

l Complete any OP, EP, and AP Changes Appropriate Procedure Revisions will'be Needed to Implement the Above Resolutions Made to Reflect Changes made to the Plant.

l

' EFV-12,13 MOV Crosstics Possibly Install AC or DC Motor Operators. l 1

Battery Safety Issues Considered as a Safety Issue and will be  !'

Resolved this Outage.

Safety Issues (Rotating Equipment Guards, Evaluation for Scope is On-going.

Ladders, Etc.)

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1 Em

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F Florida INTEROFFICE CORRESPONDENCE Power Nuciear site Suonort SA2A 240-4756 CORPORAT!ON OFFIE MAC TELEPNOME

> l

SUBJECT:

CR-3 SITE DIRECTION l T0: Nuclear Operations Personnel DATE: October 28, 1996 VPNP96 0062 I

One of the recommendations of August's Root Cause Analysis Team was the establishment of site-wide priorities to help establish our focus for us as we manage day-to-day and I near term work efforts.

The site senior management team has taken this recommendation and developed Crystal River Unit 3 Site Direction, a multi-colored poster being placed on site information boards and on the site video system.

Site Direction consists of two parts:

CR-3 Challenges - currently to improve our safety culture and ensure organizational and programmatic changes are effective.

CR-3 Top 10 Priorities - the 10 most significant issues we need to focus on in the near term.

Our most significant challenges and top priorities are expected to change from time-to- ,

time, and will be reviewed periodically by the senior management team for l appropriateness. As we determine an issue is being satisfactorily resolved, it may be removed from the list to be replaced by another significant issue. I Use Site Direction to help establish priorities in your daily efforts and to enhance teamwork as we work together to make CR-3 an exemplary performer.

. M. Beard, Jr.

JSB:PMB/lf Attachment

i CRySTSLE RIVER 'UXIT3 SITE DIRECTION UNIT CHALLENGES:

e To strengthen the nuclear safety culture throughout the nuclear organization and in all elements of our nuclear program.

o To ensure organizational and programmatic changes are effective To meet these challenges, the following are the top 10 CR-3 priorities:

] a Improve safety system margin and methodically validate the design basis for key plant systems.

m Revise the corrective action process to include a single graded approach to problem identification, effective root and apparent cause determination, and meaningful performance monitoring and trending.

m improve human performance with emphasis on eliminating operator work-arounds, enhancing procedures, reducing administrative j burden, and improving human error reduction skills.

d a Revise and validate the 50.59 process against industry standards; communicate and effectively implement the new process.

m Achieve technically accurate and timely regulatory submittals.

a Evaluate current and projected site workload against resources to

achieve sustained backlog reduction rates.

I n Critically examine the integrated work process, including planning, scheduling and work control for both the on-line cnd outage l environments; establish an improvement plan.

a Establish and communicate standard methodologies by which we will i

manage change, including emerging issues and organizational and programmatic change.

! a Create and implement an effective Management and Supervisory Development Program.

a Improve adherence to nuclear security requirements.

4 l

4 10/31/96

i October 22, 1996 y s

'y i..

NUCLEAR E ** WMmiJ

,( ," ;

l OPERATIONS [ [ '

NEWS Reflecting on 1996, Our Core l

Values, and Our Principles For Conducting Business. . . Af ter e e eve at a wng teamwnka completing a record run in 1995 with a capacity factor of 100% which E"* "' *# "" # Y' '" "#E" '

continued into January 1996, we were fM people an usential to our

, on our way to Refuel 10 scheduled to "# " " '

begin in February 1996. Then the 1 challenges began. . .you know them as O M M CUM well as I. . . Forced outage in Januarya M M E E l. -

INPO plant evaluation, Extended Listed below are the key principles i

Refuel Outage, NRC IPAP Inspection, that we must follow in conducting the l Shutdown in September, Unusual Event business of nuclear power operations, in late September, Shutdown due to Along with the core values, they several design issues that require establish a philosophy for achieving

, resolution, and currently our long excellence.

range plan to discover and resolve all design issues as quickly as 1. Continual 17 reinforce in the minds possible. of each person, the dis tinctive l

• essence" of nuclear operations I You have dealt well wi th all professionalism whACh is nuclear i challenges faced. Remember, each Bafety -- the protection of the l challenge we have come thru is a reactor core --

through personal small victory. Now it is time to responsibility, conservative move on. As we move forward, we will decision-making, and a questioning stay with our " Game Plan"...Our Core attitude.

Values and Beliefs and Our Principles For Conducting Business. It's time 2. Ensure that regulatory requirements to revisit them: (Pat Beard) are fully met. This requires establishing and maintaining an open C_o_re Valucp_pnD__Beliefg1 dialogue with the NRC and other o We believe that low cost electrical ugulatry agencies.

I energy is a vital strategic ingredient of the U.S. economy. It . Demonstrate a strong crum4 tment to is our responsibility to preserve and training and accreditation as a deliver the full promise of the vehicle to avoid an attitude of nuclear option: safe, dependable, and complacency. Recognize that an competitively priced electricity. aganization and its people must continually strive to improve or o To a person a we are dedicated to perfumance will inevitably degrade.

conducting our business to the highest safety standards. 4. Focus on the iden tification of problems and their solutions with an

\ o We believe in the pursuit of attitude that small things that are excellence through critical self- wrong will likely lead to large assessments and the creation of a problems if not promptly corrected.

learning environment which leads to a requires com itment to ultical continuous improvement. ' 'U ~ * " '"*'" U " *" 'U"

• corrective action tracking system a and regular solicitation (continued)

L l

6 NON...page 2 scheduling for day-to-day work, plant outages, system outages, on-line of feedback irom craft and supervisory employees.

maintenance / projects, and major modifica tions . )

5. Emphasize the use of operating 13. Keep a strong configuration experience, both externally and management philosophy that is interna 11ys develop the capability reflected in modification and for rigorous follow-up and maintenance activities. (Maintaining iden tifica tion of root causes and the design basis of the plant is a corrective actions for problems. responsibility of each employee.)

6. Create an environment that promotes 14. Emphasize emergency preparedness teamwork among groups and through strong management support and individuals. This requires prompt the development and exercise of action to address instances where effective plant and corporate teamwork is lacking, emergency plans.

7. Establish and maintain strong and 15. Maintain superior material effective channels of communications condition of the plant with few

-- up, down, and across Nuclear operator workarounds.

Operations. Place particular emphasis on " upward communication

• so that 16. Keep a strong focus on improving personnel are able to freely bring industrial safety.

issues to management's attention. ' ********' "*****'*"*****"*"**

8.save clear assignments of responsibility and a rigorous CONGR A T UL A TIONS I . . . th e Radiological Emergency Response exercise of accountability throughout Program exercise was very successful.

Nuclear Operations.

A total team etfort that exhibited

9. Insist on high standards of excellcat performance was recognized personal performance and, by the NRC during their critique on October 18, 1996.

particularly, emphasize the importance of the role model This was a full participa tion portrayed by managers and supervisors. This exercise with participation by the includes an NRC, state, counties, and evaluated intolerance for poor performance. by FEMA.

10.Carefu11y select people wi th potential, and provide etiective The NRC considered the exercise a technically accurate and sufficiently initial and continuing training and challenging scenario which was well qualification to prepare them for controlled.

their duties and maintain their knowledge and skill level.

The exercise demonstrated successful

11. participation off site as well, which Stress the development of is a good reflection on FPC because personnel for higher positions of of interfaces developed with the responsibility such that a strong counties and state agencies.

management staff is established and maintained. Utilize selected training No violations or weaknesses were and job rotation as key development iden tified. Also, during the NRC tools.

critique, several positive comments

12. Establish within the various were received which will be reflected in their report.

groups of Nuclear Operations, a comprehensive and continuous planning This is the most positive interaction and scheduling function that looks with the NRC this year.

ahead, anticipates problems, I commend and all of you, avoids surprises. (Of particular importance is the planning and Thanks, Pat

)

1. " FPC / NRC i= _ ~ i Engineering Meeting Agenda to j October 31,1996 2

- Introduction Pat Beard, Jr.

. Root Contributors to Engineering Performance Gary Boldt

• Corrective Actions Gary Boldt Fran Sullivan

)

Greg Halnon ,

Jim Baumstark

) . Measures of Effectiveness JimTerry i

l Outage Scope Gary Boldt Closing Remarks Pat Beard, Jr.

3 J

Enclosure 2 J

The MESSAGE

_________z r""2EsaM=sesa:E e_ .

)

e We have a problem.

e It is our fault.

e it will be fixed by us or by a new team.

e You are part of the "new" team if:

5 You have resolve fire in your belly) to sustain c ange.

3 n You understand and can implement the triad, " Safety versus Production versus Cost".

3 n You can be continually objective of yourself and those who work

) for you.

3

Independent Design

> Review Panel Purpose

-- ,,gg ,g-.-

)

e Perform an independent assessment of the Crystal River Unit 3 Design Bases and the adequacy of the design l 3

bases management control processes to provide l 3

reasonable assurance of safe plant operation.

J e Identify other related areas that a warrant further consideration.

3 o Provide recommendations for improvement.

Independent Design

Review Panel Results

?

-wu_em ,_

i o

e Did not see a basis for a significant j safety concern.

b e importance, recognition, and b ownership of the plant's Design l Basis needs to be improved throughout the organization.

3 e Some elements of the process for 3 maintaining the Design Basis of the plant need to be upgraded.

J e Design margins in some CR-3 systems are lower than in other 3

B&W plants.

J

~

l

> Root Contributors to Engineering Performance nmemmyggenyggj cgn:cge = we 7

6 o Safety culture not

, appropriately emphasized e Insufficient communication of a management expectations l e ineffective oversight and self-3 assessment l 3

e ineffective change l management D

3 l

l 3 Root Contributors to l Engineering Performance 3

w m m mran n y r g r ur n:7:m m r w w p e Safety cu ture was not effectively

emphasized

b Attention to safety was not commensurate with that given to production and cost.

3 Design basis issues were primarily '

resolved by analytical means rather 3 than physical means (e.g. plant '

moc.ification or test):

- Some original design limitations had not 3

been improved

- Plant changes took advantage of 3

available margins Narrowly focused safety evaluations.

J 2

D

, Root Contributors to Engineering Performance  !

f .

mm==q=mmmm _ .

l 3

o Insufficient communication of management expectations, particularly with respect to l establishing a comprehensive:

3 Definition, documentation, and understanding of the plant design

) basis.

Program for inter-departmental 3

plant configuration control.  !

Program for networking with other g

B&W plants to maintain consistent designs and system safety margins.

?

)

i f

1

( Root Contributors to Engineering Performance

-m emenggr g rp e n.*e.

3 e ineffective oversight and self-assessment: l 3

Not sufficiently self-critical.

Slow to recognize extent of the a problem.

Identified symptoms rather than root 3

causes

- leading to inappropriate priorities.

NGRC, PRC and QA audits 3

(defense in depth) not sufficiently critical.

3 Insufficient performance monitoring and trending

- ineffective corrective action system 3

4 J

J

, Root Contributors to Engineering Performance D

+ 6 Trsaariabs m!ngzzrs +nz w D

e Ineffective change management.

Excessive organizational and f

programmatic change:

- Re-engineered business processes o - Reduced contractor support

- Downsized staffing

- Relocated corporate staff to site  !

- Extended surveillances 18 to 24 ,

months a - Implemented ITS l

- Sent many of the best engineers to SRO training Increased human error rate 5

3

Corrective Actions to Achieve a Turnaround memargg tg;gyg.syrgg<,wr ..

?

o Purpose b To assure Engineering is a role model for the safety 3

conscience of the plant.

D e What actions have been 3

taken and what are the results ? l D

D D

Corrective Actions to Achieve a Turnaround f

_ ,yg g7 _

f e Extended outage to l demonstrate corporate f support for safety culture s

i e Committed to improve design margins by physical means

? e Had Engineering stand-down l to review 50.59 problems e Completed an independent

( review of FPC's design basis issues (IDRP; D

D

Corrective Actions to 2

Achieve a Turnaround j we43mman;gsgsg. meg;myc.en -

.s 4

f e Strengthened the inter-departmental Design Review 3

Board o Strengthened networking with other B&W owners L e Emphasized expectations

o Increased use of

b outside peers / subject experts in self-assessments use of independent consultants for design reviews J

D

> Corrective Actions to Achieve a Turnaround

. _ ,, ,. y g g g g g y 7 m _ y .

a e Conducted FPI performance benchmarking / training o Established top ten engineering priorities D

o Improved effectiveness of 3 NGRC, PRC, QA l i

e Improving performance 3 monitoring, root cause, and corrective action programs D

D 9

.D

6

, Corrective Actions to Achieve a Turnaround 3

2" f"$$$$idh..._ _2155 5$$$5535 F*6"E" 3 e Increased engineering design effectiveness t7 rough a restructuring e Increased total engineering staffing level

Filled vacancies wit, ,

b l engineers ' rom ou': sic e FPC 3

e Acded two acministrative supervisors o

e 10 0

i

?  :

! i 6

Corrective Actions to 1 Achieve a Turnaround

., 7 . .- .

D o Forming contract partnership with Parsons (formerly Gilbert Associates}

3 On-site team t1 rough ' 1 R e Creating a formal change J 3

management process J

J J

11 J

?

d Restructuring D

e=esarw asrw -

D e New

Title:

3 Nuclear Operations Engineering Improve focus on operations 3 support Multi-faceted organization (not 3 just design} l e New Mission Statement:

P " Provide Superior Technical  :

Support for CR3 Operations '

with a Constant Focus on Nuclear Safety."

e

l Organization Chart  ;

1 3

X.O.E. l

_ _ a=~;-wav;a;= = x z;;sw-vcmm _

,wmem..........mx+:.......-.-.-.-. , ,

3 Mgr, NOE FX Sullivan Additional Direct Reports (3 Clerical and 2 Staff positions)

D I I I I I 1

Supv, I&C Supv. Elec Supv. Struct Supv. Mech  !

Nuc. Engr Nuc. Engr Nuc. Engr Nuc. Engr l WS KOLEFF JS ENDSLEY DL JOPLING CL MILLER 3

1 g

Supv. Nuc. Engr Supv. Nuc. Engr Supv. Nuc. Engr Admin. Fuels Mgmt & Safety Analysis Admin.

O SK BALLIET RW KNOLL A PETROWSKY O

COLOR CODE KEY RED- New NOE Management Personnel O

BLUE-New NOE Management Positions 2

0

3

> Cultural Changes 3

e Safety related design margin 3 e 50.59 reviews e Reduce operator burden J

e Project scope teams a e Design Review Boards (conceptual / multi-organizational) a e 3rd party reviews e Integrated project schedule D

e Appropriate work hours D

3 D

> Work Load 1 Management

. rsygyyggyymn-

)

Request for Project &

3 Plant Modifications Modification Review Group l Must Do/ Approve High Want List

)

u Schedule Coordination Group D

U l J

Cycle Reaort i 4

J

__ _ _ _ _ _J

b

> Safety Evaluation

)

Reviews 1 um graggggg:grpm*rw .

e .

j p e Review of 50.59's being performed by the j newly formed Safety Analysis Group Multi-disciplined / background review i

i Consistency of reviews Consistency of final product

?

Lessons learned applied to all 50.59's D

50.59s are a stand alone product D

D 5

D

?

Design Review Board (DRB}

)

• ess;rtmes:ggn:cqgsgrewcw*+

b o Design Review Boards will be held on all design projects deemed important to plant safety and to plant operations.

e Design Review Boards will ae held at the direction of plant management, NOE p management and at the request of the j design engineer.

o Board members derived from:

-Core NOE membership

- Required inter-departmental members

- Additional inter-departmental 3

mem aers e Meetings not held unless the core NOE 3 members and required inter-departmental members are present.

6 J

O Design Review Boarc (DRB}

O MMifrG22 kza ~~ ::" . 2R M ???m? m v e The DRB mem3ers are individually responsible /

lO accountable for the aspects '

l of t7e project within their area lo of expertise.

k l e h o project reviewed ay the l:O DRt3 shall be issued "or construction until such time it o

is unanimously approved by

':he DRB core and required o members.

O

e i

g Why We Are Better Right Xow!

l e We admit we have problems!

3 No band-aids (design)

, No short term organizational and 3 programmatic fixes e Restructured!

p Management team changes l 12 net new positions Improved experience mix

! Integrated Safety Analysis into NOE to improve 50.59 reviews e Improved interface with operations.

8 3

> Why We Are Better Right Now!

p

o Cultural changes are happening !

o Better prioritizing of engineering work and development of integrated schedules i

)

S J

J 4

9 D

> PRC Effectiveness g

,mmm_;______z=m b

e Attributes of an effective PRC Independence from issue g

Diverse know edge l Diverse experience b Questioning behavior Empowered 3

Frequent Attendees e Specific PRC for ASV-204 .

i Not sufficiently independent 3 Weak in opera: ions experience Questioning Behavior not effective 3 Two attendees no': frec uent J

PRC Corrective Actions

, _m_;_____=mg o Issued Expectations to all a members and alternates e One-on-One discussions on-going with chairman and each potential quorum member

, o Requirement to have diverse g

experience built into meeting protocol 3

o Leve of significance of meeting .

agenda will factor into meeting protocol The higher the sig nificance, t7e more stringent at:endee requirement D

a ROOT CAUSE PROGRAM .

EXHAXCEMEXTS 3

rsaresses:__ _ ;;_se = ezzwe ~

i o Senior Management Training )

3 Completec October 3,1996 l o Precursor /Proalem Reaort Eva uation a Team initiated October 7,1996 o Root Cause Training October 29 -

3 Novem oer ' ,1996 o Apaarent Cause Training November 3

13 and 14,1996 o Procec ure C1anges in progress 3

Process f ow clart complete Interfaces identifiec 3

Revisec CP-111out for review o Communication of change to site aersonne J

o Prog ram Imp ementation Novem aer

'8,1996 J 1 Florida Power Corporation

a 3

OTHER XEAR TERM IXITIATIVES 3 .

g qg g _ y yamg g _ -

e Safety Culture Survey (FPI) conducted week of October 21,1996 e Suaervisor and Worker Human Error 3

Prevention Training 2 day course for Supervisors (90 total)

D 1 day course for Workers (120 total) e Management Team Assessment p 2 independent looks by outside sources l Results to Sr. Vice President e Wee < y management focus meetings h to c evelop/ affirm and fo d into ' 997 Business Plan Goals Core Values

• Priorities i Florida Power Corporation 2

l MEASURES OF EFFECTIVEXESS e TO MEASURE AND MONITOR f ENGINEERING PERFORMANCE IN THE FOLLOWING AREAS:

3 RESOURCE LOADING OF ENGINEERS b MAR QUALITY PLANT SUPPORT CAPABILITY 3

SAFETY CULTURE i

b l e SOME MEASURES OF EFFECTIVENESS ARE BEING DEVELOPED BASED ON MCAP 11 AND INDUSTRY EXPERIENCE.

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IX SCMMARY l

) l e RESOURCE CHALLENGE IS BEING MET

?

STAFF AUGMENTATION IN PROGRESS IMPROVED PROCESS FOR PC'S

) REA/ MAR BACKLOG REDUCTION IN PROGRESS e PLANT SUPPORT IS IMPROVING OPERATOR INTERFACE STRENGTHENED 3 OPERATORS WORKAROUND LIST /MCR 4 DEFICIENCIES IMPROVING i PROJECT TEAMS BUILD TEAMWORK b

e SAFETY CULTURE IS IMPROVING

) MANAGEMENT RE-EMPHASIS CONTINUOUSLY LOOKING FOR DB ISSUES.

" MARGIN IMPROVEMENT" OUTAGE 3 PROVIDES UNCOMPROMISING MESSAGE TO STAFF 3

s Extended Outage Scope

_ymme__

l f e Purpose i

To improve safety margins in f the top ariority systems using physical means (modification '

)

or test).

o Systems affected LPl/Dropline L HPI Upgrade Diese Upgrade

) EFW/EFIC Upgrade BSP-1B Upgrade

)

96-06 RB Penetration Upgrace 2

j

Extended Outage

, Scope

---

,=7g.,

l

)

e Other items l 5 FMEA of LOCA, LOOP, loss of DC power.

J Maintenance

-control board deficiencies 3 -secondary plant focus J

D D

13 g