IR 05000413/1987021

Insp Repts 50-413/87-21 & 50-414/87-21 on 870617-19.No Violations & Deviations Noted.Major Areas Inspected:Closeout of Open Insp Items

ML20236B862
Person / Time
Site:	Catawba
Issue date:	07/23/1987
From:	Lawyer L, Long A, Shymlock M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20236B310	List: IR 05000413/1987021 ML20236B306
References
50-413-87-21, 50-414-87-21, NUDOCS 8707290262
Download: ML20236B862 (11)

Text

,

p UNITED STAT ES

]

[ p RFoog'o, WUCLEAR HEGULATORY COMMISSION 1

*

& 's s REGloN ll

,*k I \b 101 MARIETTA STREET, '

I .t ATLANTA. GEORGI A 30323

. %,...,..,/

Report Nos.: 50-413/87-21 and 50-414/87-21 Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba 1 and 2 Inspection Conducted: June 17-1q,1987 Inspectors: UL /_ 13$) 2 7 A. R,.'Lodgi Date Signed ,

A lu 11 3 YDk']

).L.La'wydr Date ' Signed Approve by:hw 1

!

LL h 7/j1/J'7 D' ate / Signed

%gM.B.'Shymlock,SectionChief Operations Branch ,

Division of Reactor Safety l I

SUMMARY l

!

Scope: This routine, announced inspection was in the area of closeout of open I inspection item Results: No violations or deviations were identifie B707290262 B70723 PDR ADOCK 05000413 G PDR

. -

- -

_ _ _ _ _ _ _ _ _ _ - - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ - - -- - _ - _ _ _ _ _ - _ - - _

.

REPORT DETAILS Persons Contacted Licensee Personnel:  !

G. Barrett, Training Records Document Control Specialist  ;

• H B. Barron, Superintendent of Operations  !

• H. Barron, Director of Operations Training l

• A. Cote', Licensing Specialist  !

• J. R. Ferguson, Unit Scheduling Engineer

'

'C. L. Hartzell, Compliance Engineer M. Janeski, Operations Training Instructor

'

.

i R. Neigenfind, Staff Engineer G. C. Rogers, Project Engineer ,

R. T. Simril, Assistant Operations Engineer  !

• G. T. Smith, Superintendent, Maintenance  !

• R. F. Wordell, Superintendent, Technical Services i Other licensee personnel contacted included engineers, technicians,.

operators, mechanics. security office menibers and office personnel.

l NRC Resident Inspectors

• Lesser, Resident Inspector

• Attended Exit Interview ) Exit Interview a

The inspection scope and findinos were summarized on June 19, 1987, with those persons indicated in paragraph I abov The inspectors described the areas inspected and discussed in detail the inspection findings, including those listed below. No dissenting comments were received from the license Item Number Status Description / Paragraph IFI 414/87-21-01 Open Design and Implementation of Corrections to Identified Human Engineering Deficiencies (Paragraph.3.a)

VIO 414/86-27-01 Closed Procedural Errors and Failures to Implement Procedures on Loss of Control Room Test (Paragraph 3.a)

UNR 414/86-27-02 Closed Failure to Provide Adequate Operator Requalification Training on Loss of Control Room (Paragraph 3.b)

.

2 i

,

l'

UNR 413/86-05-05 Open Environmental Qualification of Hydrogen Skimmer Fans - Open Pending NRC Policy Determination (Paragraph 3.c)

IFI 413/86-05-01 Open Revision of Station Directive 3.2.2 to i Require Shift Supervisor Notification o i Missed Surveillance Tests (Ptragraph 5.a)

IFI 414/86-07-03 Closed Review and Implementation of Environmental i Qualification Maintenance Program (Paragraph 5.b)

l The licensee did not identify as proprietary any of tne material provided to or reviewed by the inspectors during this inspection.

'

, Licensee Action on Previous Enforcement Matters (92702) (Closed) Violation 414/86-27-01: Procedural Errors and Failures to i Implement Procedures on Loss of Control Room Test 1 During the Unit 2 Loss of Control Room Test on June 27, 1986, the transfer of contro1 *of Steam Generator Power Operated Relief Valves l (PORVs) to the Auxiliary Feedwater Pump Turbine Control Panel )

i

'

( AFWPTCP) erroneously commanded all four PORVs to open to seventy-five percent of full stroke. Reactor pressure and pressurizer level.,

which had been decreasing slowly as a result of the cooldown after the reactor trip at the start of the test, fell rapidly. Within a l minute of the transfer, pressurizer level indication was lost, and i l within two more minutes pressure had dropped below 1845 psig gener-ating a safety injection (SI) demand signal. By design, the transfer of control to the auxiliary panels had blocked automatic SI initia-tio After another three and one-half minutes of unsuccessful I attempts to manage the situation from the auxiliary panels, control )

I was returned to the control roo The transfer back to the control l room automatically intiated SI. By this time pressure had dropped as low as 702 psig.

'

The underlying cause of the event was the failure to specify in the Design Change Authorization or other documents that the mode of control of the steam generator PORV controllers at the AFWPTCP had j

'

been changed. This in turn led to a failure by station personnel to change procedures and train operators on this modification. The l situation was further exacerbated by human engineering deficiencies j introduced by the modifications. Other contributing factors included j the lack of a human engineering deficiency review of the shutdown

'

panels, inadequate training on shutdown panel instruments and controls, inconsistencies in labeling of instruments and controls, and reluctance to terminate the tes l i

__ -_ -_ _ -_ . _ - - _ - _ _ _ _ _ ____-______-_____a

I i

.

'3

The. NRC issued a Confirmation of Action Letter (CAL) on July 3,1986,  ;

containing corrective actions applicable to Units 1 and As- i documented in NRC Inspection Reports 413/86-27, 414/86-30, and I 413/86-36, 414/86-39 the corrective actions in the CAL were completed and the test was successfully repeated on July 11, 198 l NRC Inspection Report 414/86-27 identified as Violation 414/86-27-01  ;

five examples of inadequate procedures or failures to follow proce- I dures related to the June 1986 depressurization event. The five examples of Viclation 86-27-01 were subsequently cited as two violations in Escalated Enforcement Action (EA)86-147, issued November 12, 1986. The licensee responded on December 12, 1986, with an admission of Violation A with comments, and a denial of Violation As a result, the NRC modified Violation B.2, and stated in the April 14, 1987 letter to the licensee that additional corrective actions were necessary for the item. Corrective action commitments for the other items in the Notice were considered acceptable by the NR (1) EA 86-147 Violation A.1: Failure to Review Design Change Authorization for Impact on Established Operating Procedures Background: I The licensee's program for design controls had not assured that Design Change Authorization (DCA) CN-2-M-1527, which changed the design basis for the mode of control for the Steam Generator (SG) Power Operated Relief Valves (PORVs), was reflected in necessary procedural modificatio DCA CN-2-M-1527 was not properly reviewed by plant personnel as required by Station Directive 3.0.3, Mar,agement of Shutdown Requests, for the effects on existing operating procedure As a result, Procedure OP/2/A/6100/04 was not modified and incorrectly specified the setpoint of the SG PORVs. Instead of remaining closed, the SG PORVs opened to approximately 75 percent of full open, contributing to the depressurization even In the December 12, 1986, response to the. Notice of Violation, the licensee stated that the following corrective actions had been implemented prior to the restart of Unit 2:

-

A review of all Unit 2 design changes and shutdown requests implemented after Hot Fr tional Testing and prior to Fuel Load j

-

Revision of the Auxiliary Shutdown Panel operating proce-dure and the Loss of Control Room abnormal procedure to reflect the changes to the panel l

_ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ . _ __ _________J

_ __ - - ___ _ - - - _ __ __ ___

.

4 i

Resolution:

]

!

Prior to the restart of Unit 2, the licensee reviewed all Unit 2 )

Design Change Requests (DCRs) and Shutdown Requests implemented )

between Hot Functional Testing and Fuel Load, for effects on j established procedures and for human factors significance. The !

inspectors reviewed the findings presented in a letter from !

W. R. McCollum to File, dated July 3,1986. The . licensee DCR l review identified the need to _ replace a particular valve label I which contained a typographical error. The licensee verified for the inspectors, by checking the label on the valve, that this relabeling had been accomplished. The licensee review of ,

Shutdown Requests identified the need to add Lighting Panel and !

Breaker numbers to Procedure HP/0/B/1001/18, EMF Sampling. The 1 inspectors verified that this was accomplished in Revision 3 to )

the procedure, approved December 15, 1986. Also as a result of )

the review of Shutdown Requests, Procedure OP/2/A/6200/01, Chemical and Volume Control System, was revised on February 3, 1987, to properly indicate new controller locations on the valve l checklist The inspectors terified that Operating Procedure OP/2/A/6100/04, l Enclosure 4.5 has been modified to specify the correct initial positions for 2NV-294 and 2NV-309 based upon the test data from .

TT/2/A/9100/03, Auxiliary Shutdown Panel and Turbine Control Panel Supplemental test, which was written to verify proper functioning of the various valves while at the Auxiliary Shut-down Panel (ASP). This test was performed satisfactorily prior to the Loss of Control Room retest on July 11, 198 The inspectors also verified that training on the aforementioned procedure changes has been included in operator requalification training (Paragraph 3.b).

(2) EA 86-147 Violation A.2: Failure to Adequately Review Shutdown Requests for Human Factors Considerations as Required l Background:

The control mode of the Steam Generator (SG) Power Operated Relief Valves (PORVs) had been changed through Design Change Authorization (DCA) CN-2-M-1527 without any visible change to the SG PORV controller or labeling at the Auxiliary Feedwater Pump Turbine Control Panel (AFWPTCP). This occurred as a result of the DCA not having been properly reviewed by design personnel.

(

The licensee stated in the December 12, 1986 response to the Notice of Violation that the following corrective actions had been completed or were ongoing:

i

.

l 5 i i i

-

A review of Main Control Board, ASPS and AFWPTCPs for both units to identify differences between units and to verify proper labeling nomenclature and units of measure (prior to Unit 2 startup)

-

Revision of Design Procedure EDP-3.17, Control Room Change

- Handling, to clarify the need for review of modifications 1 to the ASP and AFWPTCP and to clarify responsibility for initiating the Control Room Change Form

-

Revision of Instrumentation and Controls Workplace proce-dure PR-3 to assure emphasis on labeling and scaling of manual loaders, controllers, et Correction of all Human Engineering Deficiencies identified in NSM-CN-20227 Resolution:

The inspectors verified that prior to the restart of Unit 2, the licensee reviewed all Unit 2 Design Change Requests (DCRs) and Shutdown Requests implemented between Hot Functional Testing and Fuel Load, for human factors significance as well as effects on ;

established procedures (Paragraph 3.a 1).  !

'

The inspectors observed the ASP and the AFWPTCP labeling nomen-clature, meter unit designations, and controller position ,

I labeling changes and verified that significant human factors improvements had been made to the panels. This observation confirmed that "0" and "C" (0 pen and Closed) labeling had been added to the panels, as well as labels which clearly identified l the control mode of the Steam Generator PORV controller Design Procedure EDP-3.17, Control Room Change - Handling, was revised to clarify the need for review of modifications to the l

Motor Driven Auxiliary Feedwater Pump Control Panels, the ASPS or the AFWPTCPs by the appropriate Design Group. Revision 3, dated August 4,1986, clarified that the procedure applies to the subject panels when the arrangement of devices is modified; when operator interface devices are added, deleted, or modified; or when the appearance, labeling, or functioning of a device on the subject panels is modifie The Electrical Division Procedure ECPI-PR-3 was rovised to ensure emphasis on labeling and scaling of manual loaders, controllers, etc. This was accomplished in Revision 1, dated April 24, 1986, by changing Section 6.11, Operator Interface, to read, " Scaling and labeling of components to support the

__

.

~

functional description should be reviewed and documented on the I&C list and the Instrument Detail. When changes or additions to Main Control Boards, Auxiliary Shutdown Panels or Auxiliary pump Turbine Control Panels are required, a human factor review in accordance with EDP 3.17 shall be requested."

As required by the Confirmation of Action Letter of June 27, 1986, the licensee' reviewed all Human Engineering Deficiencies (HEDs) identified in NSM-CN-20227 and their schedules for ,

implementation. As a result, complete re-engraving of all )'

nameplates on the ASPS and AFWPTCPs was accomplished prior to August 22, 1986. As discussed in a July 30, 1986 letter, the remaining portions of NSM-CN-20227 could not be implemented at j that time since the remaining section required de-energizing part or all of the systems on the ASPS or the AFWPTCPs. All l

HEDs are required to be corrected prior to restart from the j

'

first refueling outage in accordance with the Facility Operrting l License. The design and implementation of corrections to che HEDs identified in NSM-CN-20227 will be reviewed in a future inspection and will be tracked as IFI 413,414/87-21-0 The inspectors verified that training was provided to appro-priate personnel and inc1 ed labeling and surface changes made to the ASPS and Unit 1/Un 2 control difference (3) EA 86-147 Violation B.1: Erroneous Valve Setpoints Background:

During the Loss of Control Room test on June 27, 1986, depres-surization occurred due to an inadequate procedure. Enclosure 4.5 of Operating Procedure OP/2/A/6100/04, Shutdown Outside the Control Room from Hot Standby to Cold Shutdown specified initial settings for valves 2NV-294 and 2NV-309 which were inappropriate and resulted in these valves assuming an incorrect position when control was transferred to the remote shutdown panel Resolution:

The inspectors verified that Operating Procedure OP/2/A/6100/04, Enclosure 4.5 has been modified to specify initial positions for 2NV-294 and 2NV-309 based upon the test data from TT/2/A/9100/03, Auxiliary Shutdown Panel and Turbine Control Panel Supplemental Test, which was written to verify proper function of the various valves while at the ASP. This test was performed satisfactorily prior to the Loss of Control Room retest on July 11, 198 ,

.

(4) EA 86-147 Violation B.2: Failure to Transfer Control Back to Control Room Background:

Test Procedure TP/2/A/2650/03, Loss of Control Room Functional Test, was not properly implemented in that control was not transferred back to the control room when a situation arose that could not be adequately controlled from the auxiliary shutdown panel ,

Resolution:

Operator requalification training specifically addressing proper implementation of TP/2/A/2650/03 and lessons learned from the June 27, 1986 depressurization, has been completed for all reactor operators and senior reactor operators (Paragraph 3.b)

(5) EA 86-147 Violation B.3: Remote Shutdown Panel Labeling Background:

Operations Management Procedure OMP 1-6, Control Panel Informa- j tion Changes, dated May 10, 1982, stated that any informational 1 changes to the co trol panel will conform to human factor guidelir,es and agree with the setpoints, limits, and precautions established in approved operating procedures. Contrary to this ,

procedure, the labels on the controllers at the remote shutdown panel for valves 2NV-294 and 2NV-309 were reversed and indicated the opposite of the intended and anticipated meaning The NRC considered the effect of the valve mislabeling to be significant to the June 27, 1986 depressurization even Resolution:

The inspectors verified that the relabeling of the ASP control-1ers for valves 2NV-294 and 2NV-309, and other modifications to the ASP, had been approved and documented on the OMP 1-6 forms for control panel informational change Through interviews with licensee personnel, the inspector verified that OMP 1-6 is now being used to document control !

panel labeling change Except for the resolution of the identified Human Engineering Defi-ciencies, which will be tracked as IFI 414/87-21-01, the inspectors concluded that the licensee had corrected the previous problems and 3 developed corrective actions to preclude recurrence of similar problem Corrective actions stated in the licensee response to the Notice of Violation have been implemente The item is therefore close i

.

l

b. (Closed) Unresolved Item 414/86-27-02: Failure to Provide Adequate !

Operator Requalification Training on Loss of Control Room

.

Background:

Prior to the June 1985 depressurization event, an upgrade of the - ;

Steam Generator PORV had been performed in accordance with DCA ;

l CN-2-M1527. This change as applied to the AFWPTCP led to confusion '

because the operator was unaware that the same pressure setpoint varying controller was being used to indicate and control Steam Generator PORV valve positio Training was deficient in not adequately teaching the Steam Generator PORV design change to each licensed operator and senior operato NRC Inspection Report 414/86-27 identified as a violation the failure to provide adequate operator requalification training on facility J design changes in accordance with Technical Specification 6.4.1, The

'

NRC letter f rom J . Nelson Grace to Duke Power Company, dated November 12, 1986, stated that in accordance with the current NRC policy statenient on training and qualification of nuclear power plant personnel the violation was not cite Resolution:

i The Steam Generator PORV design change as performed under DCA CN-2-M1527 displayed several deficiencies, predominately in the human factors engineering aspects of the change. These deficiencies were i identified in Nuclear Station Modification NSM-CN-2022 l l Deficiencies relating to changes to nameplates and labels on the ASPS and AFWPTCPs were identified, incorporated into lesson plan transparencies and layout drawings, and instruction was provided to all licensed reactor operators and senior reactor operators.

,

In addition, training included a detailed discussion of the June 27, 1 1986 incident. The lesson included the changes that were made to Procedures OP/2/A/6100/04, Shutdown Outside the Control Room from Hot Standby to Cold Shutdown, and AP/2/A/5500/017, Loss of Control Roo The training also covered labeling and panel surface changes made to the ASPS, control differences between Unit 1 and Unit 2, and proper use of the newly revised panels and procedures to shutdown the reactor and plan Based on this information, the item is closed.

l (0 pen) Unresolved Item 413/86-05-05: Environmental Qualification of Hydrogen Skimmer Fans - Open Pending NRC Policy Decision Status:

The item remains open pending an NRC policy decision on Environmental Qualificatio l i

!

. -l

'

I i Unresolved Items No unresolved items were identified during the inspectio . Licensee Action on Previously Identified Inspector Followup Items (92701) I (0 pen) Inspector Followup Item 413/86-05-01: Revision of Station Directive 3.2.2 to Require Shift Supervisor Notification of Missed Surveillance Tests  !

Background:

During an inspection in March, 1985 it was noted that Station Direc-  !;

tive (SD) 3.2.2 was inadequate in several respects and was not being followed in all case All of these problems have been resolved except for on It was noted that Station Directive 3.2.2 only addressed the notification of Performance and Compliance when a surveillance test could not be performed within the required time interval. The procedure did not state that the Shift Supervisor must i i

be immediately notified if a surveillance interval had passed. The licensee had stated that SD 3.2.2 would be revise The licensee noted that failure to meet a surveillance requirement was coverea l under the provisions of Station Directive 3.1.8.

'

i Status:

l As of June 19, 1987, Station Directive 3.2.2 had not been revised to ,

require notification of the Shift Supervisor when a test had not been I completed within the interval required in Technical Specification The licensee committed at the Exit Interview to complete this action by August 31, 198 (Closed) Inspector Followup Item 413/86-05-06, 414/86-07-03: Review l and Implementation of Environmental Qualification Maintenance Program Background:

10 CFR 50.49 requires that a record of environmental qualification (EQ) of electrical equipment important to safety must be maintained to permit verification that each item meets its specified performance requirements when it must perform its safety function up to the end of its qualified lif Implicit in this requirement is the constraint that records must be kept to substantiate that periodic maintenance activities required to maintain a piece of equipment in its qualified condition have been performe The licensee specifies these-periodic maintenance requirements in the Station Equipment Qualification Reference Index (EQRI), When the EQRI was first being implemented, it contained numerous references to instruction manuals and it was not clear which periodic maintenance

_ _ _ - - _ _ _ _ _ _ _ _ _ _ - _

.

10 activities were actually required for EQ. In a letter dated March 27, 1985, outstanding maintenance activities or alternative actions taken by the station were identified for a Design Engineering Review to

,

verify that all EQ-mandated maintenance had been accomplished. NRC l Inspection 86-07 identified that as of January 9,1986, the Design Engineering Review had not been complete Although an in-depth review and revision of the EQRI was currently in progress, the potential existed that the qualifications of some equipment may have been compromised or invalidated through a failure to perform necessary periodic maintenanc Resolution:

The inspector verified that the concerns presented in the March 1985 letter had been adequately resolved, as documented in a letter dated January 27, 1986. No equipment qualifications appeared to have been invalidated or compromised by the identified alternative maintenance i actions.

)

The EQRI manual has been completed by the licensee. The inspectors verified that EQ-mandated maintenance now is identified in the EQRI Manual. The sources of the EQ requirements are also reference Based on the above information, the item is close ________-____-_--_________a