IR 05000335/1996022

Insp Repts 50-335/96-22 & 50-389/96-22 on 961114.No Violations Noted.Major Areas Inspected:Aspects of Licensee Configuration Mgt & 10CFR50.59 Programs

ML20137R844
Person / Time
Site:	Saint Lucie
Issue date:	11/14/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137Q937	List: IR 05000335/1995014 IR 05000335/1996022 ML20137Q928 ML20137Q944 ML20137Q952 ML20137Q999 ML20137R008 ML20137R041 ML20137R279 ML20137R538 ML20137R587 ML20137R823 ML20137R883 ML20137R903 ML20137R914 ML20137R927 ML20137R960 ML20137R967 ML20137R973 ML20138G413
References
FOIA-96-485 50-335-96-22-01, 50-335-96-22-1, 50-389-96-22, NUDOCS 9704140264
Download: ML20137R844 (16)

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Docket Nos: 50-335, 50-389 License Nos: DPR-67, NPF-16 Report No: 50-335/96-22, 50-389/96-22'

Licensee: Florida Power & Light Co.

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Facility: St. Lucie Nuclear Plant, Units 1 & 2

. Location: 9250 West Flagler Street Miami, FL 33102

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Date: November 14,1996 j

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Inspectors: J. York, Reactor inspector ,

C. Rapp, Reactor Inspector M. Miller, Senior Resident inspector Approved by: C. Casto

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Chief, Engineering Branch Division of Reactor Safety r

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9704140264 9703d7 PDR FOIA b,.., B,1 NDER96--485 ,PDR ,

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REWRITE THIS SECTION !!!!!!!!!!!!!!!!!!!!!!!!!!!!!!!!!!Ill!!!!!!!!!

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EXECUTIVE SUMMARY ,

St. Lucie Nuclear Plant, Units 1 & 2 i NRC Inspection Report 50-335/96-22, 50-389/96-22

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. This special inspection included aspects of licensee's configuration management and 10 CFR 50.59 programs. Specifically, the inspection examined the extent to which plant  ;

changes were appropriately incorporated into procedures and drawings and the

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performance of 10 CFR 50.59 safety evaluations. Conclusions included the following:

A review of a number of screenings and evaluations performed pursuant to 10 CFR 50.59 resulted in the identification of four apparent violations:

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• One example of an apparent failure to perform a safety evaluation due to a failure to employ engineering controls in the construction of the Unit 2 Control !

Element Drive Mechanism Control System room and a continuing failure to recognize the nondocumented nature of the room (paragraph E1.1.b.1).

• One example of an apparent failure to identify that the installation of a .

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temporary fire pump represented a change to the plant as described in the  :

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Update Final Safety Analysis Report, resulting in a failure to perform a safety evaluation (paragraph E1.1.b.2).

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One example of an apparent failure to recognize that refueling equipment l l

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setpoints were included in the Updated Final Safety Analysis Report while l performing a safety evaluation screening, leading to a failure to perform a safety evaluation (paragraph E1.1.b.3).

• One example of an apparent failure to recognize an unreviewed safety .

question in the development of a safety evaluation for an Emergency Diesel ,

F Generator fuel oil transfer line valve lineup change (paragraph E1.1.b.4).

A review of off-normal operating procedures relating to safety-related annunciators  ;

identified a number of inaccuracies (paragrapb E7.1).

• Five apparent failures to properly incorporate Pla Change / Modification packages  ;

into drawings and procedures were identified (paragraph E7.2).

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ai inspection Scope

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The inspectors reviewed activities associated with the plant modification used to i replace the Nuclear Instrumentation drawers located in the Unit 1 control room. This i modification roulted in Ni channels A, B, C, and D being wired backwards due to a .

design error. This review included evaluations of the root causes and safety 4 significance from a core physics view poin ,

.i f b. Observations and Findinas

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On July 30,1996, St. Lucie Unit 1 was operating at approximately 100 % power

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when reactor engineering was analyzing the data taken during power ascension and

noted an anomaly in the results. The data indicated three of the four excore linear -

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detectors measured core power moving to the top of the core during power

- ascension. This was an unexpected phenomena and did not agree with the trend of l the power moving to the bottom of the core indicated by RPS. Channel B Linear

Range Detector, Control Channel #9 Linear Range Detector, and the BEACON Core Power Distribution Monitoring System. Evaluation of the data collected indicated that

- RPS Channels A.C,and D could have reversed (rolled) leads of the top and bottom chambers input to the RPS drawers.

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The modification performed during the outage associated with this problem was N .

PC/M 009-195. During the outage, the licensee replaced the power range NI '

j . drawers for the Reactor Protection System (RPS) with new Gamma Metrics drawers.

p' - This modification combined the linear power range input to the RPS and the  ;

logarithmic wide range channel into a single drawer, i.e. reduced the number of drawers on Unit i from eight to four. This modification increased the limits of the ,

instruments range and replaced aging equipmen )

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1) Evaluation of Root Cause A design error was responsible for the reverse connection (rolled leads) on four NI i

, safety related drawers on Unit 1. The Controlled Wiring Diagram (CWD), no. JPN- -

009-195-001/002 depicted the upper Uncompensated lon Chamber (UlC) connected )

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. to the lower UIC input at the NI drawer. The root cause noted that the designer and -

the lead engineer interpreted conflicting information on the existing CWDs and made ,

an assumption.'

The independent verification may have caught this error had the process been i properly performed. The drawings were prepared by the lead designer with input  ;

from the lead engineer. The drawings were then checked by a second designer who l

!. . had no special knowledge of the Ni design. This check was essentially. a drafting l

'< check. The drawings were then reviewed by the lead designer and then by the  !

engineering supervisor.- .

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Engineering Quality instructions (Ql) 1.7. Design Ir:putNerification, dated July 5,- l l  :

1995, states in part that " Design verification is the process whereby a competent

individual, who has remained independent of the design process, reviews the design .

.- inputs, ... and design output to verify design adequacy. This independent review is  !

provided to minimize the likelihood of design errors in items that are important to

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nuclear safety." Contrary to this requirement the first reviewer could not be

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considered as competent because he'was not an engineer as required by QI 1.7 and i the lead engineer as the third reviewer could not be considered to have remained- -

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- independent of this design project. : One of the action items to prevent recurrence

was to check all the l&C and electrical PC/M to see if all the drawing approval' ,

signaturesfcould qualify as independent verifiers. The licensee found three out of  ;

i eight open modifications where this was a potential problem, two of these  :

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. modifications were electrical and one was l&C. This therefore is not an isolated i case.: This failure to perform independent verification according to procedure is identified as example one of an apparent violation (eel 50-335/96-22-01, " Failure to

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Control the Design Process According to the Requirements of 10 CFR 50, Appendix  :

B, Criterion lil," EA 96-457). ]

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2) Evaluation of Safety Consequences l

The licensee had installed BEACON during this refueling outage to replace the older

IMPAX code used for in-core flux monitoring. BEACON provided several significant I improvements over IMPAX one being real-time flux profile monitoring. This improvement permitted reactor engineering to identify the NIS problem quickly and

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initiate prompt corrective actions.

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During power operations, reactor engineering used BEACON to obtain the actual in-core flux profile. The actual in-core flux profile was then used to verify compliance l with Technical Specifications and provide calibration information for the excore NIS

, drawers. As part of these routine surveillance, reactor engineering compares actual l i in-core flux profile to the in-core flux profile predicted by the core design cod Reactor engineering noted larger than normal errors between actual and predicted in-core flux profile. Because BEACON used the same neutronics engine as used in the core design code, reactor engineering could not explain the error and notified the corporate core design engineers. As part of the process to resolve these errors, it was discovered that a simplifying assumption, used to overcome limitations of the

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lMPAX, was not accounted for in the original design of BEACON. This simplifying assumption was used because the licensee had changed the fuel design to l incorporate a longer end cap to prevent debris induced fuel failures. This longer end l cap raised the overall core height by 2.64" causing an offset between detector  !

. . midplane and actual core midplane. The IMPAX code assumed detector midplane  !

- was along core midplane and could not accommodate the 2.64" offset. Therefore, l

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the licensee,' after discussion with the fuel vendor (Siemans), used this simplifying I assumption to essentially lower the core midplane by 2.64" so that final design

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- output would be referenced to detector midplane; not core midplane. However, the  !

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. engineer preparing the design input for BEACON was not aware of this simplifying I assumption consequently BEACON was referenced to core midplane resulting in an

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4 3 I- increased error between the core design predicted in-co'.e flux profile and actual in- I core flux profil The licensee's root cause evaluation identified lack of cross-discipline review as the

, significant contributor to this design error. The inspector concurred with the .l 4: licensee's evaluation. . Engineering Qus!ity Instructions (Ql) 1.7, Design inputNorification, dated July 5,1995, states in part that " Design verification is the process whereby a competent individual, who has remained independent of the . J

< design process, reviews the design inputs, ... and design output to verify design .

. adequacy. This independent review is provided to minimize the likelihood of design l

' errors in items that are important to nuclear safety." Contrary to this requirement, . -I

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the design inputs were not adequately reviewed by a competent individual in that the core midplane offset was not identified as a design input for BEACON. This failure ;

2 to perform an adequate independent design review for the BEACON system is

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' identified as example two of an apparent violation (eel 50-335/96-22-01, " Failure to - !

Control the Design Process According to the Requirements of 10 CFR 50, Appendix  ;

B, Criterion ill," EA 96-457).

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The safety significance of reversing the detector inputs to the NIS drawers . _

l substantially reduced the safety margin between the TM/LP trip setpoint and the 'l analysis limit even considering the increased TM/LP margin to the trip setpoint due !

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to actual core operating conditions. The safety impact. of the failure to identify the l core and detector midplane offset on TM/LP or LPD safety limits was minima l The licensee also identified that BEACON was placed into service on Unit 1 without any benchmarking against IMPAX, the on-line core performance monitoring code BEACON was replacing, instead, BEACON was installed on Unit 2 and benchmarked against CECORE, which did not require any modifications to accommodate the core midplane offset. Technical Specification 6.8, Procedures and

- Programs, paragraph 6.8.1 requires in part that written procedures recommended in i Appendix A of Regulatory. Guide 1.33 revision 2, February 1978, shall be .

established, implemented... Engineering Quality Instruction (Ql) 3.7, Computer *

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Software Control, revision 1, Section 5.4. requires that SQA1 software shall be

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validated and verified (V&V'ed) in accordance with Section 5.6. Section 5.6 states that new software shall be V&V'ed prior to use. V&V includes the use of test cases to ensure the new software produces correct results. Item 4 of Section 5.6 states '

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that technical adequacy shall be determined by comparing the test case to results i from attemative. methods such as functionally equivalent and previously validated software. In the case of BEACON, IMPAX would have been functionally equivalent software. Benchmarking BEACON against IMPAX may have identified the design error concoming core midplane offset because the two codes would not have yielded

the same results. _ Contrary to this requirement, BEACON was placed into service on

Unit 1 without benchmarking against IMPAX. This is identified as an apparent i violation (eel 50-335/96-22-02, " Failure to Follow Procedure for Placing the BEACON System in Service," EA 96-457). .,

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3)- Corrective Actions

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4 i The inspectors reviewed the possibilities that the licensee had to determine that a ,

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design control problem existed by reviewing information from QA audits and from  :

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information in the root cause analysis. The QA monitoring report, QSL-PM-96-17,-

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. had a finding conceming problems with the Ni modification. This finding discussed

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the large number of Change Request Notices (CRN), a number of scope changes, ,

and the writing of four separate work orders for troubleshooting purposes. Fourteen >  !

Condition Reports (CR), four by the QA crganization, were written during the .  ;

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implementation of this modification.- Besides this being a qualitative indicator of a -

, problem with implementing the modification, the QA report noted that " workers on j

. the job frequently complained about the unmanageability of the implementation  !

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- documentation".

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The NRC SRI had a ' discussion with one o! the Reactor Operators (RO) who had

- noted and questioned an anomonly in the reading of the linear NI values at or below  !

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five percent power. In a review of the licensee's root cause report in the. area of

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personnel interviews, it was noted that ROs questioned the readings below five i percent, and on increasing power above five percent. In addition, this report states - ,

that a Reactor Engineering representative also questioned the difference when in the L ' increasing power range of 70 to 90. Therefore, the results of the areas mentioned in

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- the QA monitoring report and opportunities afforded by the questions from the ROs/ Reactor Engineers should have allowed the licensee to identify design problems and to have taken effective corrective action. The failure to take effective corrective l 4 - ' action is identified as an apparent violation (eel 50-335/96-22-03, " Failure to Take l Effective Corrective Action to Prevent a Design Error," EA 96-457). l

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c. Conclusions on Conduct of Enaineerina l As a result of this inspection three violations were identified for the engineering area.

n The first violation had two examples of problems with the licensee's design control

process. One example involved the failure to independently verify Controlled Wiring I

l Diagrams which were in error and resulted in wiring the Ni drawers backward '

! Another example resulted in the core midplane offset not being identified as a design

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input to BEACON, the computer program used for real time flux profile monitorin The second violation involved the failure to validate and verify a new computer )

i program before placing it in use. The third violation. involved the failure to take

- effective corrective action to prevent a design error from being implemente Conduct of Maintenance

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a. inspection Scope The inspectors reviewed the maintenance activity for replacing the no. 6 detector for

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channel B of the Linear Range Detector. ' The maintenance activity allowed reversal i'

of the field cables.

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• . All four of the RPS Linear Range Detectors had the connectors reversed as  !

previously discussed but the B channel unlike the other three channels was giving l

- .the correct data. At the same time that the drawers were being replaced on Unit 1, j h ; the detector for channel B (detector no. 8) was being replaced as a maintenance

. . activity. During connection of the field cables, the connections were reversed for the -

3~ ' upper and .!ower detection chambers, thereby causing the B channel to record-

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i-D The root cause for the swap of the cables was that'the new detector had different

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labeling than the existing cables. The existing cables were labeled TOP SIG and 1

BOT SIG, and the new detector had A and B. The inspectors discussed this - .

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. maintenance job with the l&C supervision who had supervised the latter part of this'

- maintenance project. Several opportunities were presented to the maintenance

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personnel, one when the detectors were checked out in the warehouse and a <

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second time when this condition was noted in the fiel Mainionance personnel'should have resolved the labeling problem by writing a

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' Condition Report (CR) and having a formal resolution. Technical Specification 6.8,.

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Procedures and Programs, paragraph 6.8.1 requires in part that written procedures

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. recommended in Appendix A of Regulatory Guide 1.33 revision 2, February 1978,  ;

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shall be established, implemented... Administrative Procedure No. 0006130, Condition .

Reports, rev 4, dated March 22,.1996 Par. 8.1.1.A states in part that "Any individual . .

who becomes aware of a problem or discrepant condition ... should initiate a CR. If )

> doubt exists, a CR form should be initiated". This failure to comply with the t requirements of the administrative procedure is identified as an apparent violation-(eel 50-335/96-22-04, " Failure to initiate a Condition Report for Labeling on Safety

Related Detectors,".EA-457).

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c. Conclus' ions on the Conduct of Maintenance  ;

i The l&C maintenance personnel reversed the field cables for the no. 6 channel B

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detector. The cables were labeled differently than the existing ones and the  ;

maintenance personnel had two occasions to question this condition by initiating a Condition Report. An apparent violation for failure to conform to administrative I procedure for writing a CR was identifie .

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REWRITE LIST!!!!!!!!!!!!!!!!!!!Ill!!!!!!!!!!lll!!Il ,

PARTIAL LIST OF PERSONS CONTACTED Licensee t

Bladow, W., Site Quality Manager -

Bohlke, W., Vice President, Engineering Burton, C., Site Services Manager Dawsoni R., Business Manage ,

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Denver, D., Site Engineering Manager Fulford, P.., Opeistions Support and Testing Supervisor _ ,

li Holt, J., Information Services Supervisor 3

. Johnson, H., Operations Manager '

Scarola, J.; St. Lucie Plant General Manager

Weinkam, E., Licensing Manager >

. Other licensee employees contacted included operations, engineering, maintenance, and corporate personne ]

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P INSPECDON PROCEDURES USED  !

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j ITEMS OPENED, CLOSED, AND DISCUSSED -

Opened

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. 50-389/96-12-01 eel Failure to Perform a 10 CFR 50.59 Safety Evaluation for CEDMCS Enclosure

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- 50-335,389/96-12-02' eel Failure to Perform a 10 CFR 50.59 Safety Evaluation C For Use of a Temporary Fire Pump -

50-335/96-1'2-0 eel Failure to Perform 'a 10 CFR 50.59 Safety Evaluation '

For Change in Setpoints Listed in UFSAR - i 50-389/96-12-04 eel Unreviewed Safety Question involving EDG 2B

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i - 50-335,389/96-12-05 eel Failure to Ensure Configuration Control P

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LIST OF ACRONYMS USED

. ATTN Attention CCW Component Cooling Water CEDMCS Control Element Drive Mechanism Control System CFR Code of Federal Regulations CR Condition Report CW Circulatory Water DFOST .

Diesel Fuel Oil Storage Tank DPR Demonstration Power Reactor (A type of operating license)

DWG = Drawing EA Enforcement Action EDG- Emergency Diesel Generator eel Escalated Enforcement item FIS Flow Indicator / Switch FO Fuel Oil FPL The Florida Power & Light Company FRG Facility Review Group gpm Gallon (s) Per Minute (flow rate)

HPSI High Pressure Safety injection (system)

ICW Intake Cooling Water IR [NRC) Inspection Report -

JPN (Juno Beach) Nuclear Engineering LIS LevelIndicating Switch MV Motorized Valve NL Non-Licensed Operator N Number NPF Nuclear Production Facility (a type of operating license)

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NRC Nuclear Regulatory Commission

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NUREG Nuclear Regulatory (NRC Headquarters Publication)

ONOP Off Normal Operating Procedure OP Operating Procedure PACB Plant Auxiliary Control Board PC/M Plant Change / Modification PDR NRC Public Document Room PM Preventive Maintenance PRA Probabilistic Risk Assessment PSL Plant St. Lucie QA Quality Assurance QI- Quality instruction QSL - Quality Surveillance Letter SAR Safety Analysis Report SE Safety Evaluation SFP Spent Fuel Pool SlAS Safety injection Actuation System SIT Safety Irijection Tank

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I- TOR Topical Quality Requirement UFSAR Updated Final Safety Analysis Report URI [NRC) Unresolved item

- USNRC Unite States Nuclear Regulatory Commission

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PAGE 27

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R 1216320e RUN DATE: 11/13/96 5AFETY ISSUE MANAGEMENT SY$ TEM GENERIC !$$UES WITH Tl GUIDANCE AVAILABLE - (OPEN) _

NS$$ THERMAL OL OL PLANT DOCKET TYPE SUPPLIER CAPACITY LICENSEE STATE REGION ISSUE EXPIRATION

....................... ........ .... .......... ........ ................................... ..... ...... ..... ..........

ST LUCIE 1- 05000335 PWR COMs 0802 MWT FLORIDA POWER & LIGHT C FL 2 03/76C 03/16 LIC ACT LICENSEE VE;i!FY TIREF VERIF" A RLICA8LE ISSUES TAC # CMP DATE IMPL DATE COMPLETE NUMBER INSPECTION REPORT NUMBERS ACCESSION

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'A 46 B105 sW9483 04/% 04/% 2515/124 SEISMIC QUALIFICAfl0N OF EQUIPMENT IN OPERATING PLANTS M f4T 6tCT* b C l'u M m e i4 @

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RUN DATE: 11/13/9

$AFETY ISSUE MANACEMENT SYSTEM GENERIC !$5UES WITH TI CUIDANCE AVAILABLE . (READY)

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. NS$$ THERMAL PLANT DOCKET TYPE SUPPLIER CAPACITT LICENSEE STATE REGION ISSUE EXPIRAT10t

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ST LUCIE 1 05000335 PWR COM8 0802 MWT FLORIDA POWER & LIGHT C FL 2 03/76C 03/16 1

.I VERIFY ;

Llc ACT LICENSEE VERIFY TIREF

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TAC # CMP OATE IMPL DATE COMPLETE NUM8ER INSPECTION REPORT NUM8ERS ACCESSION f.PPLICA8LE' ISSUES ^

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GL-89 10 8110 M75721- 06/90C 02/95C 03/97 2515/109 91 18 94-11 SAFETY RELATED MOTOR OPERATED VALVE TESTING AND SURVEILLANCE MPA 8041 8041 M43853 05/84C 02/85C 2515/062 ,

FIRE PROTECTION * FINAL TECH SPECS (INCLUDES SER SUPPLEMENTS)

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'RUN DATE: 11/13/96 SAFETY lSSUE MANAGEMENT SYSTEM GENERIC ISSUES WITN Tl GUIDANCE AVAILABLE - (READT) j NSSS THERMAL CL - OL PLAN DOCKET TYPE SUPPLIER CAPACITT LICENSEE STATE REGION ISSUE EXP!RAfl0N

....................... ........ .... ...,...... ........ ................................... ..... ...... ..... .......... i ST LUCIE 2 05000389 PWR COMB 0000 MWT FLORIDA POWER & LIGNT C FL '2 G3/83C 04/23 )

LIC ACT LICENSEE ' VERIFY TIREF . .

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- APPLICA8LE ISSUES - TAC 8 CMP DATE 'IMPL DATE COMPLETE NUM8ER INSPECTION REPORT NUM8ERS ACCESSION

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GL 89-10 8110 M75722 D6/90C 03/96C 03/97 2515/109 91 18 94 11 SAFETY.RELATED MOTDR DPERATED VALVE TESTING AND SURVEILLANCE l

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f- DATE: 11/14/9( ,

' c ITEM DETAIL REPORTS IFS . INSPECTION FOLLOW.UP SYSTEM REPORT BY $1TE - TIME: 9:19:3'

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• T RPT/ IFS / SEO.NO ITEM REF NBR / SEVERITY SALP REPORT / STS' CREATE' CLOSE0lli . CL50lli CLS0lff

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A EA/ NBR NOV.!D TYPE EA.NBR ~ SUPLMNT AREA EVENT DT- DATE PRJ/ACT* ORG NO EMP INSPECTION REPORTS

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STLl I 94 008 -03 'URI ENG 04/08/94 0 04/13/94-TITLE: QUALITY LEVEL OF PORY AND SRV DISCHARGE PIPING PROC NUMBER: 37700

' STLA I 96 001 02 UR MAINT 03/18/% 0 03/22/ % 2320 96 001 c t c d 4 D '

TITLE:' IMPROPER HEALTH PHYSICS PRACTICE PROC NUMBER: 62703 STL1 I 96 0041 05 URI -

OPS 04/29/ % 0 05/08/ % -

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TITLE: CONFIGURATION CONTROL MANAGEMENT

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PROC NUPEER: 71707- l STL2 .I 96 004 05 URI OPS 04/29/ % . 0- 05/08/ % 2230 -

NOTE: DEFINITION FOR CHARACE R PRECEEDING REPORT NO: I = INSPECTION REPORT NUMBER. E = EA NUPEER (ENFORCEMEN N = IFS NUMBER. NUMBER USED TO IDENTIFY NON. INSPECTION ITEMS

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TOTAL OPEN ITEMS' n 4 *IF ITEM IS OPEN, PROJECTED CLOSEOUT DATE IS REPORTED TOTAL OPEN REPORT SEQUENCES -n IF ITEM IS CLOSED. ACTUAL CLOSE0VT DATE 15 REPORTED

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i: REPORT BY SITE TIME: 8:56:!'

PAGE: 1 i SITE: STL ST LUCIE I

~ UNIT' RPT/lFS/ SE0 NO ITEM REF NBR / SEVERITY SALP REPORT / STS CREATE- CLOSEOUT CLSOUT. CLSOUT UPDATING ABBR EA/ NBR NOV.10 TYPE EA NBR SUPLMNT AREA EVENT OT DATE PRJ/ACT* ORG NO EMP - INSPECTION REPORTS

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STL2 1 93 025'01 IFI 12/01/93 0 12/07/93 2350-TITLE: REVIEW OPERABILITY OF UNIT 2 MOV HV.08 13 OURING THE PROC NUMBER: 37700 STL1 ! 94 008 03 ' URI -

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ENG 04/08/94 0 04/13/94 . 2320 JVL

. TITLE: QUALITY LEVEL OF PORY AND SRY DISCHARGE PIPING PROC NUMBER: 37700 STL1 I 94 011 01 VIO 4/1 ENG 06/13/94 0 06/08/94 .

2350- ,

TITLE: INADEQUATE CORRECTIVE ACTION FOR HOVS WHICH STALLED PROC NUNCER: 2515/109 *

STL1 1 94 011 "02 IFI ENG 06/13/94 0 06/08/94 - 2350 TITLE: .INADCOUATE RECOGNITION OF MOV TEST PRESSURE AND FLOW PROC NUPBER: 2515/10 .

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STL1- ~ ! 94 011 03 IF! ENG 06/13/94' ' O 06/08/94 2350 TITLE: LACK OF INSTRUCTIONS OR GUIDANCE FOR TREN0!NG PROC NUMBER: .2515/109 STL2 - N 94 332 LER 94 006 01 07/14/94 0 10/11/94 2230

/p

. TITLE: TRIP CIRCUIT BREAKER FAILURE DUE TO A BROKEN PIECE OF PROC NUMBER:

STL1' N 95 005 LER- 94 009 00 11/22/94 0 01/04/95 . 2230 ' f Aq WhG W-

-TITLE: INADVERTENT SAFETY INJECTION ACTUATION SIGNAL /CONTAINME PROC NUMBER:

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~ NOTE: DEFINITION FOR CHARACTER PRECEEDING REPORT NO: I = INSPECTION REPORT NUMBER. E = EA NUMBER (ENFORCEMENT / NOV ITEM) *

N = IFS NUMBER. NUMBER USED TO IDENTIFY NON INSPECTION ITEMS

......................... . ................................. ............ ....,..... .......................... ....... ..... ....

TOTAL OPEN ITEMS n 7 *IF ITEM IS OPEN. PROJECTED CLOSE00T DATE IS REPORTED TOTAL.0 PEN REPORT SEQUENCES -n 7 IF ITEH 15 CLOSED, ACTUAL CLOSE0VT DATE 15 REPORTED

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INFORMATION ON THIS PAGE IS FOR 0 F F.I C I A L USE ONLY

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, e a8800 UNITED STATES

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/p NUCLMR REGULATORY COMMISSION 3 9'o$ REGION H A NTA,G dlA IN November 14, 1996

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MEMORANDUM TO: Roy P. Zimmerman. Associate Director Associated Director for Projects FROM: Ellis W. Merschoff. Director [ /////

Division Reactor Projects // 7W SUBJECT: INCONSISTENCIES BETWEEN NRC POSITIONS IN R PONSE TO TIA 95-013 AND NRC INSPECTION MANUAL CHAPTER. PART 9900 INTERIM

~

GUIDANCE ON 10 CFR 50.59. ISSUED IN APRIL 199 Attached is a copy of Florida Power and Light (FPL) response (L-96-254 dated ,

10/21/96) to a violation involving the 2B Emergency Diesel Generator oil line '

unreviewed safety question (NRC Special Inspection Report Nos. 50-33 /96-12 (EA 96-236 and 96-249)). While FPL agreed to the violation. FPL ;

identified inconsistencies between NRC positions in response TIA 95-013 and !

NRC Inspection Manual Chapter. Part 9900 Interim Guidance on 10 CFR 50.59 l 1ssued in April 1996. NRC Inspection Manual Chapter. Part 9900 (Page Paragraph 4) specifically states that " .. the staff has found compensating !

effects such as changes in administrative controls acceptable in offsetting l uncertainties and increases in the probability of occurrence or consequences I of an accident previously evaluated in the safety. analysis report or I reductions in margin of safety, provided the potential increase or reductions i in margins are negligible." On the other hand. NRC positions in response to j TIA 95-013 suggests that compensatory measures can no longer be credited to 1 offset small potential increases in probabilit Therefore the licensee requests that NRC resolve the differences in interpretation and apparent inconsistencie Since this issue is related to the current efforts to review the implementation of 10 CFR 50.59. it is provided to you for appropriate actio Docket Nos. 50-335 and 50-389 Attachment: As stated CONTACT: Kerry D. Landis. DRP/ Branch 3 404 331-5509 cc: S. Ebneter L. Reyes. RII J. Johnson. RII F. Hebdon. NRR F. Gillemie. NRR L. Wien',. NRR K. Lar. dis. RII C. Julian. RII A. Gibson. RII C. Casto. RII M. Miller. RII j Ncol30a9 4

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1 now. p , a upt c . e.o. nos ussa.J sw n.n.zussesas l

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l OCT 211995 J I

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L 96-254

< 10 CFR 2.201 l

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'96 0CT 23 f.!1 58

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U. S. Nucicar Regulatory Comnussion i

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i Atta: narn-e Contte! Desk

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Wa Mamaa. D. C. 20555 1 i

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Re: St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Reply to a Notice of Violanon NRC Snecial Is.E+:de- Recon 96-12 (EA 96 236 and 96-249)

Florida Power and Light Company has reviewed the subject Notice of Violadon and. p m

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to 10 CFR 2.201. the responses to the violations are anache Very truly your ,

) f l T. F. Plunkett President Nuclear Division l

f TTPHASEJW f i

Attachment cc: Stewart D. Ebneter. Regional Admmistrator. USNRC Region II Senior Resident Inspecter. USNRC. St. Lucie Plant l

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Attachment j

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STATE OF FLORIDA )

) s COUNTY OF PALM BEACH )

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J. A' Stall being first duly sworn. dp< and says:  !

Tha' L is Vice President. St. Lucie Plant, of Florida Fower & Light Company, the Licensee

That he has executed the foregoing document: that the statements made in this riarament are I true and correct to the best of his knowledge, information and belief, and that he is - "5---iz+i to execute the document on behalf of said Licensee. . i I

l i

J. A. Stall .

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STATE OF FLORIDA .

COUNTY OF 4/n e Swom to and subscribed before me  ;

this d day of M 19 by J. A. Stall, who is personally known to m A bstD % w '

Name of Notary Public - State of MA S.Eggaw w amamammaaammsersus Junt amans==se tear .e (Print. type or stamp Comrrusioned Name of Notary Public)

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L 96-254 Annch.w.,

Renly to a Newien of Violeiaa  ;

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1 VIOLATION I: -

10 CFR 50.59. " Changes. Tests and Ewm ;s." provides, in part, that the licamme may make Th =g= in the facility as described in the Safety Analysis Rapost (SAR)-

l without pnar e amm..=ma approval, unless the proposed change involves sa unremewed safety queenon. A proposed change sha!! be deemed to involve an

.

- a d safety quesnon if the probability of occurrence of a malfunction of eqmpe=at i-y--d to safety previously evaluated in the SAR may be increased, if a possibility for an accidear or malfuncuan of a different type than any evainneed previously in the SAR may be created, or if the margm of safety as dermed in the basis for any technical specificanon is #=d l

Contrary to the above in July 1995, the licensee made a change to the facuity which i involved an unreviewed safe:y question without prior Commit < ion approva Sa--huy, the 2B E%ay Diesel Generator (EDG) fuel oil line was rnann Il isolated to secure a through-wall fuel oil leak. In taking this acnon, the tiemaw introduced two new failure modes for the 2B EDG, which both is d the probability of occurrence of a malfuncuan of the EDG above that previously evaluated in the SAR and the possibility for malfunction of a diNerent type than any evaluated i pseviously in the SAR. resulting in an unreviewed safety question. (01013) l This is a Severity L: vel III violation (Supplement I)

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l RESPONSE I-i-

' FPL concurs with the violation.

t REASON FOR ' IRE VIOLATION

The cause of the violation was that FPL procedural guidance for performmg 10 CFR

550.59 evaluations in place at the tine of the violation was not consistent with the NRC's ir%y..i. des of the regulatio De subject evaluation was prepared in accordance with the " Nuclear N= i-g -

C.+ uwent Giiidmac for Perfonmng 10 CFR 50.59 Safety Evaluations." %e yds-y

basis for this engmeenng procedure was NSAC-125. "Gaidaliaan for 10 CFR 50.59 l Safety Evaluations." which is the generally accepted industry standard on the subjec ;

!

In July 1995. both the FPL procedure and NSAC.125 allowed a conclusion of no

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L-96-254

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Arrach tnent Realv to a Notice of Violation .

increase in probability if the increase was deternuned to be insignifican .

The subject 10 CFR 550.59 safety evaluation allowed plant operation with the 2B diesel fuel oil transfer pump discharge isolation valve (normally locked open) in the 1 closed position in order to isolate a leak in the underground portion of tbs pipin r'+y;- =;--y actions were required by the evaluation to enswe the valve would be opened in the event of an EDG start. 'Ibese : ;= =-y actions were ca==imat with the guidelines of NRC Genenc Letter 91-18. "Information to r t nyng d Two NRC Inspection Manual Sections on Resolution of Degraded and N=-- '- - -5

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Conditions and on Operability" for the use of ==an=3 action in place of anunnanc action. As noted in the FPL evaluation, two new failure modes and a slight increase in the probability of a component failure were created. Pursuant to the FPL procedure. the evaluation concluded that this slight increase in probability was

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insignificant and that no unreviewed safety question existed as a result of the ywyewd

! plant configuration.

CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED

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The unreviewed safety question identified in NRC Inspection Report 96-12 was associated with the closing of the desel fuel oil transfer pump discharge isolation valve. Compliance was re-established upon restoring the valve to its normally open

position following replacement of the leaking uncbyeund piping downstream of the valve. De replacement of the leaking piping, restoration of the valve to the open position and return of the 2B EDG to OPERABLE stams were completed by November 25.199 .

.t . CORRECTTVE SEPS TO AVOID FURTHER VIOLATIONS FPL Engineering issued a Ta hair =1 Alen to r.ng neering personnel on March i

6.1996, informmg them that when perfo: ming 10 CFR 550.59 evaluanons, any quantified increase in the probability of occurrence of accidents or any qtutified increase in the probability of occurrence of a malfunction of

, eqmpment important to safety must be considered an unreviewed safety question.

, Revision I to the " Nuclear Engineering Guidance for Performing 10 CFR 50.59

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Safety Evaluations" was issued on May 17. 1996. This revision procedurahzed the requirement stated in the Technical Alert discussed in 4.A. abov . . - . - . - - -- - - . . - -.---..- -_ . -

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! L-96-254 i Aetnehntaar ,

i Reely to a Notice of Violation i

! FPL will evaluate the need for furtherem.ed. l revisions taking into account ,

the NRC's position dc== =4 in TTA 95-013 and the latest industry guidance J

on y.Jo..Ji.g 10 CFR 550.59 evalumnons. This acuan will be completed within three months following the issuance of NRC and li.J y guidanos on -

l i narfar-a= 10 CFR 550.59 evainmaa"=

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j Full compli= ace was achieved by November 25,1995, with the c _-- ;@"= ofItem 3 i above.

! ADDITIONAL INFORMATION i

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While FPL concurs with the violation as cited, it was not until August 19,1996, that t

FPL had the benefit of reviewmg the NRC memvi.iidum dated July 30,1996, which responded to a Technical Assistance Request (TIA 95 013). The response to TIA 95-l 013 illustrates the existing confusion surrounding io y.m.iion of 10 CPR 550.59.

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The issue of concem deals with the Staff's position on the introduction of now failme

! modes as they relate to permined cc ;===- y actions. The NRR respones to TIA

95 013 contams a narrower io.my.m. tion of the peranssible use of =danaiar='in

controls, sp=A&=lly c9-- ;-

==g actions, when - - -- ; =.J to a previous NRC

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i position. Specifically, NRC faea-ion Manual. Part 9900 interun (= on 10 CPR

150.59, issued in April 1996. Part 9900 (pg. 3, paragraph 4) states that ". ths staff has e found compensa mg effects such as changes in =dmimeranve conuols Whle in

- oMummag uncertamties and increases in the probability of occurrence or consequences

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of an accident previously evaluated in the SAR or reductions in margm of safety, provided the potential increase or reductions in margin are negligible."

i On the other hand, the response to TIA 95 013 suggests that co-g =---y measures

can no longer be credited to offset small potential increases in probability. In the case

of the FPL 10 CPR 550.59 evaluation, a co-aaamary operator action was used in place of an automatic function. Specifically, the response to the TTA asserts that "an unreviewed safety question exists hae=naa the pmposed change intmduces a now

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pacedme and =*aar* malfunction of a different type (operssor error) " The response to the TIA further asserts that "[I]n general the introduction of eaaTaaa'ary r measures suggests that there is an unreviewed safety question for wtuch eaaTaa*=aa" is naadad hane* a 50.90 subminal should be y..y .J by the licensee and evaluated by the staff to determine whether the compensation is adequate." 'This posttion conflicts with the position set forth in the April 1996 Part 9900 guidance.

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1 1<96-254

, Attachment  !

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Reniv to a Notice of Violation l

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The new position also has implications for the NRC's operability guid- in Genenc Latter 91-18. "Information to I i-- Regarding Two NRC T=i- -:d= Manual tacriana on Resolution of Degraded and Nonconformmg raatiiriaan and on Operability." Under this g"id~. NRC mangnizes that ==harnarvaa of mammal aconn )

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for =anara=rie action may be acceptable under certain l - -- Adadanauy,in )

i i an NRC letter to Northeast Nuclear Energy c.-mpany dated h 21,1994 (John '

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Stolz to Richard M. Kacach), it is stated that. "[T)f an operability canch=ian is made based upon implementing campen..rary actions r=8ria- in a change to the facility or

procedures as described in the FSAR, an evaluation pursuant to 550.59 must be

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performed ." However, as stated above, the new position taken in the response to the i TIA. with regard to ce r ==~y rocasures, would appear to foreclose the possibility

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i that such a change could be made pursuant to 10 CFR 150.59.

! 'Ite inconsistencies between the positions set forth in the TIA response. NRC

la=a~ Hon Manual Part 9900 guidance, and docketed courg-

-="= illustrare the cunent state of confusion regardmg the i-y.. Mon of 10 CFR 550.59. In the response to the TIA. the Staff states that it "...is. in the process of beoer defining what

, constitutes ayy vydage use of compensatory measures in 10 CFR 50.59 safety evaluations." Further anesung to the cunently evolving state of 10 CFR 550.59 i-y.. Mon is NRC's " Action Plan for Improvements to 10 CFR 550.59 i

' l Implementation and Oversight." from James M. Taylor to Chairman Jackson, dated Apnl 15.1996. The action plan recogmzes several issues in need of clardicarina and I

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that a final paper to the Commmion on the action plan is not =charialari for issuance until February 1997.

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FFL respectfully requests that. in light of the posidon on the introduction of new 4 faihut modes through manual operator acuon expressed in the response to the TI I the Staff resolve, in a timely rnmaner, the differences in interpretation and apparent inconsistencies that exist. The Staff's resolution of these differences will penet FPL and other licensees to properly implement the' requirements of 10 CFR 150.59 and 10 CFR 150.90 in day-to-: fay plant operations.

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L-96-254 Attachmant

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Rsolv to a Notice of Violanon  !

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VIOLATION II A- l l

c 1 1 10 CFR 50, A=;-adiv B " Quality Assurance Criteria for Nuclear Power Plams and '

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Fuel br--=-3 Plants." Critanon III requuss, in part, that maammes he ====Mahad to asmee that applicable .=_ 4= y requirements and the design basis for safety-telated

and components are co .dy translated into s;+ N- i =

saucanes, ey&

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drawags, Fh and instruction j

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! Plonda Power and Light Company Topical Quality Assurance Repon. TQR 3.0,

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Revision 11 implements these .q.uwts. Section 3.2, " Design &==== Control,"

r J e in part, that design ch= u shall be reviewed to ensure their * - ; '= =:=

l is in each case, coordinated with any necessary changes to operatmg y..ci4 In !

l- addition. Seenon 3.2.4, " Design Verification." provides. in part. that design connel

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nessures shall be established to verify the design inputs, design process, and that the

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design inputs are correctly incorporated into the design outpu Contrary to the above, the licensee failed to coordinate design changes with the '

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nace==ry changes to operanng evce.Jm as evidenced by the followmg examples: j

! Plant n=a-/Madifiention (PC/M) 109-294, "Setpoint Change to the j l Hydrazme Low Level Alarm (LIS-07-9)," was completed on January 6,  ;

1995, without ensurmg that affected Procedure ONOP 2-0030121, i " Plant Annunct=rar Su ry," was revised. "Illis resulted in Annunciator S-10. "HYDRAZINE TIC LEVEL LO," showmg an  !

inconect setpoint of 35.5 inches in the ptocedur . PC/M 268-292, " Intake Cooling Water Lube Water Piping Removal and i

! Circulatory Water Lube Water Piping Renovation." was c---;' ^ 4 on l l " .,. e.y 14,1994, without ensunng that affected Prnaarhwe ONOP 2- l

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0020131. " Plant Ananamatar Summary," was revised. '11ds resnited in the insuuctions for Annunciator E 16. " CIRC WTR PP LUBE SPLY

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BACKUP IN SERVICE." inconectly requiring op .zers to vertfy the position of valves MV 21-4A and 4B followmg a safety injecnon

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=cmarion system signal to ensure they were de energized and had no

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control room position indiMon.

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' PC/M 275-290. " Flow Indicator / Switch Low Flow Alarm and Manual Annunciator Deletions." was completed on October 28,1992. without

, ensuring that affected Procedure ONOP 2-0030131. " Plant Annunciator i l

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i I-96 254 Anmeh- nt

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Renly to a Notice of Violation -

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} Summary," was revised. This resulted in the mstructions for safety-

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related Aaaaaei=mrs LA-12. "ATM STM DUMP MV.08-18N18B

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OVERLOAD /SS ISOL." and LB-12 "ATM STM DMP MV-08-19A/19B OVERLOAD /SS ISOL." l-.widy requiring operssors to i check Ausn/ Manual switch or switches for the snannal position. @2014)

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This is a Severity Level IV violation (Supplement I).

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RESPONSE II A- - '

, FPL concurs with the violation.

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, REASON FOR THE VTOLATION

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The cause of the violation was an madequate con 5guration control process which failed to ensure that pwce.Luws and processes affected by plant mMiMemnaa* were

' identified and updated in a timely manner as required to accurately reflect the

modificanons made in the plant.

A contributing factor to this violation was that, at the time of implementarian of the

. plant modifications in the examples to the violation, there was a general acceptance by 1 plant management of routine backlogs for open items related to plant matinemnons.

l CORRECTTVE STEPS TAKEN AND THE RESULTS ACHIEVED

'Ilie St. Lucie Plant Annunciator Summary Procedure. ONOP 2-0030131, was revised to correct the tivee discrepancies identified in this violation.11is action was completed on July 5,1996.

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4 CORRECTIVE STEPS TO AVOID FURTHER VTOLATIONS

A self.=== amen =ar of the plant mMinearian fmnt-end review process was conducted. which included benchmarking with FPL's Turkey Point Plant. In addition, a Quailty Assurance audit of the design control yim was caad"~i which supported the results of the self-=====<mear St. Lucie Design Control Procedure QI 3-PR/PSL-1 was revised to incorporate i a positive check for completion of procedure updates prior to system turnover l 6 l l

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L-96 254

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! and to ensure updatmg and tracking of affected ev - - pnor to seasonng a

) madinad sysum to service. "Ihe revised QI 3-PR/PSL-1 process was that

i-- ; ! =- --i at SL Luca Plant dunng the Sunnner 1996 St. Incie Unh I , refuehag outage. Procedme QI 3-PR/PSL-1 was funhar revised, post-Unit I !

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sufneling outage, to incorporaes the resnits of the self-assessamma discussed in i 4.A. abov l

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l The Configmation M - ----at Group at St. Lucie Plant was reorgantzed and reem-hi with the addidon of stafHng and sw. ion to suppost the new a plant modificanon review process and to conunne to opunuze process connel

] and design /rocedww integration.

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! Plant management expectadons and requuements for thorough review,

processing, and closecut of madiMr=aon-related action items, and the .

docismentatinn of these aedous prior to restorms a mndsfied sysmen to service,

!, were enmmumented to plant personnel via Items 4.B. and 4.C., above. The plant's =-- ;=== of backlogged madification-related action items was .=;!=d

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! with an unambiguous rq-w; to fully process mndifiennan daeamaa'=aan

to closure prior to declaring restoration of operability.

l Full compliance was achieved on July 5,1996, with the completion ofItem 3 above.

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VIOLATION II B ,

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, 10 CFR 50. Appendix B. " Quality Assurance Critena for Nuclear Power Plants and

! Fuel Reprocessmg Plants." Critenon III requues, in part, that measures be established to assure that applicable regulatory requirements and the design basis for safety-related

[ saucares, systems, and components are conectly trunciated into s;- I *- =

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l drawings, procedures, and instructions.

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Flanda Power and Light Company Topical Quality Assurance Report TQR 3.0, Revision 11 i==3- these requirements. Secuan 3.2, " Design Change Controi"

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provides, in part, that design changes shall be reviewed to ensure their : ;-- --- - -tation

is in each case, coordinated with any necessary changes to operaung procedores. In i

addition Section 3.2.4. " Design Versfication." provides, in part, that design control

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measures shall be established to verify the design inputs. design process, and that the l

design inputs are correctly incorporst-d into the design outpu !

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L-96 254 Annehment Reely to a Notice of Violation i

Contrary to the above, the licensee failed to assure that the design of the &=! ::-5 and Intake Cooling Water System was conecdy tr=n=larad into plant drawing Specifically, dunng imp!=== Mon of PCM 341-192. " Intake Cooling Water I.mbe

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Water Piping Removal and Circulatory Water Lube Water Piping Renovation," the as-built Drawag No. JPN-341-192-008 was not incorporated into Drawing No. 87704-0 2. " Flow Diagram Circulating and Intake Cooling Water System." Revision 11.

J Sheet 2 issued May 9,1995, for PCM 341-192.11Is resulted in Drawag No. 8770-

! GO82 enuncously showing valves 1-FCV-21-3A and 3B and associated piping as still inernlied. (03014)

This is a Severity level IV violadon (Supplement I).

RESPONSE U B:

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> FPL concurs with the violatio . REASON FOR THE VTOLATION The cause of the violation was cognitive personnel error by utility drawing. update

' personnel who failed to incorporate as-built drawing number JPN-341-192408 into drawing number 8770-G.082. Revision 11. Sheet 2, which resulted in an inaccurate documentation of design changes made to the plan Several additional factors contributed to the event

The drawing update discrepancy noted above was not identified during

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subsequent independent review by utility personnel pnor to releas For the drawing that was not updated. a discrepancy was observed in the computer based drawing update tracking program in place at the tune, regarding the date of transminal of the subject drawing to dmimant control.

i-This discrepancy in transmittal date provided a source of confusma regardmg actual drawing status at the time of transmittal. and contributed as a causal

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factor to the violation.

• The rmssed drawing update in this event was associated with a plant modificauon which was implemented over a long period of time, which also contnbuted to the even .

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L-96-254  !

Aarnehmane t Realv to a Notice of Violation ,

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. Accountabilities and responsibilides associated with the Drawing Update l checker and verifier roles were informally commanienta,i at the time drawag verificanan was mitially performe . CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED

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Drawing number 8770.G-082, Sheet 2, was revised to incorporate as-bmit drawmg

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n==iw JPN-341-192-008, which deleted valves 1-FC%21-3A and 3B. This revision  :

4 was issued on April 24,199 ; CORRECTIVE STEPS TO AVOID FURintx VIOLATTONS 1 A complete review of plant change / modification (PC/M) 341-192 was

+

conducted. Three additional drawing errors were discovered and all drawings '

, have been corrected and reissued.

! This event was reviewed with drawing update personnel to provide traming and

ensure '=L ==%- of the responsibilities. accountabilities, and expectanons of personnel involved in the rm of drawing update.

. To ensure proper updarmg, and to generically assess the potential for other ,

! errors resulting from other causes, a sample of updated drawings from previously implemanrari PC/Ms is being reviewed. This sample includes drawings exhibiting a potential discrepancy t, wacs status tr=a-1 date, as

. described above, and also drawings associated with PC/Ms which were

, implemented over an extended penod of time. " Itis anion will be complete by

December 30,1996.

'- The computer based system used to track drawing updates was converted to a new system in 1996. The new system allows g.0-si cWi--5 drawing updates to prmt a complete list of all drawmg updates requued for a given modtScanon. The accountabilides and ispossibilities ===arinrari with the

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Drawmg Update checker and wifws are bener riefineri in that drawag update personnel are required to use this list when verifying the drawing W=

performed by the drafter for a given modificanon package. The use of this list

.

aids personnel in verifying that all applicable revision requirements for a given drawing have been incorporated.

. Full compliance was achieved on April 24.1996. with the completion of Item 3 above.

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