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UNfTED STATES NUCLEAR REGULATORY COMMISSION
' REGION 11 3- S 101 MARIETTA STREET N.W., SUITE 2300 j ATLANTA, GEORGIA 303234190
%*****/ July 1, 1996 MEMORANDUM FOR: Oscar DeMiranda, Senior Allegation Coordinator Enforcement and Investigation Coordination Staff FROM: Kerry D. Landis, Chief Reactor Projects Branch 3 f Division of Reactor Projects *
SUBJECT:
RII-95-A-0065 - CONCERNS REGARDING MANAGEMENT COMMUNICATION AND CONFIGURATION CONTROL The Division of Reactor Projects performed a review and independent inspection of the following four concerns:
- 1. The public was not notified of events as required
- 2. Technical manuals were outdated
- 3. Vendor darwings were not up to date
- 4. Poor communication between organizations.
Our inspections regarding these concerns have been completed and our findings are documented in the enclosures to this memorandum.
Based on the information provided and the inspection results, we were not' able to substantiate concern number 1 and were able to substantiate concerns 2, 3, and 4 of the allegation.
This concludes the staff's activities regarding this matter. This allegation is considered closed. If you have any questions, please contact me.
Enclosure:
- 1. Allegation Evaluation Report 9703130188 970306 3' ,
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ALLEGATION EVALUATION REPORT l
l ALLEGATION RII-95-A-0065 CONCERNS REGARDING NANAGENENT CONNUNICATION AND CONFIGURATION CONTROL ST LUCIE NUCLEAR PLANT DOCKET N05. 50-335 Alm 50-389 CONCERN:
Region II received information from a concerned individual (CI) who expressed concerns related to personnel practices at the St. Lucie Point Nuclear , Plant.
The CI provided the following information:
- 1. The public was not notified of events as required .
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- 2. Technical manuals were outdated
- 3. Vendor darwings were not up to date l
- 4. Poor communication between organizations -
DISCUSSION:
CONCERN 1: Inspection reports for the previous two years were reviewed and there were no indications that the licensee failed to notify the NRC of events as required by 10 CFR 50.72 or 50.73.
CONCLUSION 1: i l
Based on the information provided, this allegation is not substantiated. ,
CONCERN 2: The CI stated that the licensee utilizated outdated tech manuals.
The problems associated with technical manuals were identified in inspection reports 94-13, 94-24 and 96-04.
IR 94-13 identified the licensee control of vendor tech manuals (VTM) and l indicated that the licensee's control of VTM updates appeared weak. The l issues associated with vendor tech manuals were identified as URI 94-13-01, VTM Control Weaknesses. Concerns associated with VTMs were as follows:
FRG Approval of VTMs ~- Because the licensee employed VTMs as procedures for maintenance activities, FRG approval of VTM was required. However, VTM revisions originated and were approved in the Juno Beach engineering offices.
Thus, the FRG was tasked with approving (for the site) a previously approved (in engineering) document. FRG, reviews have resulted in VTMs being rejected for use at the site. The result of such rejections was that the active revision for such a VTM differed between the site and engineering. The licensee provided data which indicated that 28 VTMs were currently rejected by
'. FRG and that 28 were still under FRG review.
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Engineering Control of VTM Revision - The inspector was provided with information indicating that multiple VTM revisions have been issued under the same revision number. That is, for a given VTM, the site has received multiple transmittals, identified by the same revision number, which contain substantively different information.
An example of such a condition was revision 6 to VTM 8770-6703. The record of revisions, transmitted to the site, stated that the revision was made per DCR-SLM-93-124. Following FRG approval of this revision, the site received a different transmittal of revision 6, with a record of revisions stating that the revision was made per PCM 138-191. Revision 7 to this same VTM was then received, with its record of revision sheet indicating that the later of the two revisions 6 was the basis for the VTH.
The inspector concluded, based upon the information above, that the licensee's program for controlling VTMs appeared weak and that continued inspection was required to conclude whether or not regulatory requirements were being violated. As the current inspection period concluded during this inspection, the issue is identified as URI 94-13-01, VTM Control Weaknesses.
IR 94-24 identified a violation of NRC requirements associated with vendor
' manuals. The violation, VIO 335,389/94-24-02, Inadequate Pr6 cess for Changes to vendor Technical Manuals, and its details are described below:
During an NRC inspection conducted on November 6 - December 3, 1994, a violation of NRC requirements was identified. In accordance with the " General.
Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below:
' Units 1 and 2 Technical Specification (TS) 6.8.2 requires that changes to procedures of TS 6.8.1.a shall be reviewed by the Facility Review Group (FRG) and approved by the Plant General Manager prior to use.
Units 1 and 2 TS 6.8.3 states that temporary changes to procedures of TS 6.8.1.a may be made provided that the intent of the original procedure is not altered and that the change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license. TS 6.8.3 also requires that the temporary change is documented, reviewed by the FRG, and approved by the Plant General Manager within 14 days of implementation. 1 Units 1 and 2 TS 6.8.1.a requires that written procedures shall be l established, implemented, and maintained ' covering the activities i recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February i 1978. Appendix A, paragraph 9.a. includes maintenance that can affect I the performance of safety-related equipment. I Procedure QI 5-PR/PSL-1; Preparation, Revision, Review / Approval of Procedures; Revision 58; par.agraph 5.12; states that approved vendor technical manuals that .contain sufficient detail and acceptance criteria may be used as procedures. It requir'es that technical manuals shall be treated as plant procedures for the purpose of proce' dure adherence. l Also, it requires that changes to technical manuals received from the
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3 vendor shall be reviewed by the FRG and approved by the Plant, General ,
Manager. ;
4 Contrary to the above, on Ap~ril 12 - November 16, 1994, the licensee's process for making'and using changes"to vendor technical manuals was.
inadequate in that it failed to assure that, prior to implementation or use, changes to technical manuals were reviewed and approved as required by TS 6.8.2 and TS 6.8.3. Examples include:
a 1. Procedure QI 3-PR/PSL-1, Design Control, Revision 32, was !
inadequate in that it stated that a Documentation Change Request 1
, (DCR) for an administrative change to a technical manual, that
. does not require any physical work in the plant, does not require plant review and approval. As a result, on October 11, 1994, the Nuclear Engineering Department issued DCR #DCR-SLM-94-043 approving an April 12, 1994, vendor-recommended change to 3
technical manual 8770-6251, Velan Valves. The DCR approved the 4' change for use by the maintenance department without the prior
- plant review and approval that is required by TS 6.8.2 and 6.8.3.
The change, in part, revised the required torque for the bonnet-to-body. bolts for a safety-related valve such as 1-V-3660 from 130 ft. lbs. to 150 ft. lbs. On November 9, 1994, the licensee's
! Document Control distributed that change to holders of controlled-
- copies of the technical manual, including the maintenance department. On November 16, the copy of the change from Document Control was put into the maintenance department controlled copy of the technical manual. Throughout this process, the change was not '
reviewed and approved as required by TS 6.8.2 and TS 6.8.3.
IR 96-04 discussed the SGCS valve problems in which it was concluded that the licensee failled to incorporate all vendor PM guidance in their PM program for Details are as follows:
SGCS valves. 4 While checking the Unit.1 SBCS valves for problems similar to that four.d for PCV-8801, I&C personnel found that PCV-8804 had a closed input air signal to the positioner but indicated a pressure increase on the output air sigr.al line to the valve actuator. I&C personnel removed the positioner to deterr.ine the cause of the increased output signal and noted the following:
I&C found a black powdery substance in the tubing connection after removing the positioner. This substance was also found to a much lesser degrea in valves PCV-8802, PCV-8803, and PCV-8805. PCV-8801 had not yet been examined.
Finding the black powdery substance. caused the licensee to perform a check of the instrument air system.
An attempt'to calibrate the positioner on the bench.was unsuccessful due to excessive wear on an internal spool valve. 'The wear appeared to be from normal in service usage. The positioner was disassembled and the licensee found the black powdery substance and a ferrous and rubber-like material
' inside the positioner. However, laboratory analysis was not possible due to the small quantity (less than 1 gram) of the substance that was available.
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i The licensee indicated during discussions that they believed the source of the rubber-like material and the black powdery substance was the actuator diaphragm. This was based on a microscopic comparison of a sample of the diaphragm with the substance collected from the positioner.
The licensee concluded that these conditions would not cause a failure and believed that PCV-8804 would have opened upon demand. This was later l confirmed when a similar output demand condition (discussed below) was found i on a Unit 2 SBCS valve. .
The inspectors examined the material collected from the Unit 1 SBCS valves, reviewed the chemistry reports for the monthly routine sampling performed on the IA system, reviewed several annual particulate' reports and concluded that the IA system was not the source of the black powdery substance or the ferrous and rubber-like material.
The inspectors questioned the licensee as to whether the Unit 2 SBCS valves had been examined for conditions similar to those'found in PCV-8804. The licensee indicated that the Unit 2 valves had not been examined because they believed that, since all the SBCS actuators.and positioners had been replaced (under PCM 047-295, Steam Bypass Control System Actuator Modification) during the Unit 2 refueling outage in the fall of 1995, there was no need to perform I this examination. ,
1 The inspectors reviewed the PCM package, completed work implementing !
documents, performed field inspections, and verified that the actuators and I positioners for the Unit 2 SBCS valves had been replaced. Dcring the field '
verification, the inspectors observed that the Unit 2 PCV-8803 output demand signal read 85 psig, which was similar to Unit 1 PCV-8804 condition.
The licensee initiated STAR 960359 to evaluate the condition for Unit 2 valve PCV-8803. The evaluation concluded that a build up of positioner output ;
pressure while the valve was in standby had no effect on valve operation. The licensee verified this conclusion by isolating PCV-8803 and capturing as-found positioner settings and valve movement, inspecting the internal positioner !
filters for debris, and performing a calibration check on the positioner. I PCV-8803 was stroked on February 28. All pressures and strokes were normal l and no movement of the valve was produced. The filters internal to the '
positioner were inspected and found to be clean. The licensee concluded that the cause of the positive output pressure from the positioner while the valve :
was in standby was an indication of leakage of the spool valve inside the positioner. Spool valve leakage could be caused by failure of the spool valve to seat or miscalibration of the spool valve signal. Spool miscalibration may ;
be caused by calibration drift or a very small original setting discrepancy. I The licensee indicated that a review of the vendor recommended steam bypass )
valve actuator and positioner PM guidance was in progress for both units and I the results of the review would be documented in In-House Event Report No. 96- l 020. . l The inspectors concluded that the licensee's investigation into the SBCS valve
. problems lacked thoroughness in that the extent of condit. ion did not include
, examining the Unit 2 SBCS valves for similar conditions. Also, not all of the ;
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! vendor PM guidance for the positioner was incorporated into the licensee's PM
- program for the'SBCS valves, and these PMs had not been performed. ,
. l CONCLUSION 2: l Based on the information provided,- this allegation is substantiated.
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CONCERN 3: CI stated that vendor drawings were not up to date , !
IR 95-01 identified an instance in which valve components were incorrectly installed. Report details indicated that the licensee discovered MV-08-03's emergency trip spring holder had been installed backwards. Maintenance !
personnel _ worked with the vender to obtain accurate drawings to install the :
holder properly. !
IR 95-04 described deficiencies that were identified during the walkdown gf !
. engineering safety systems. Many of the deficiencies described differences '
between actual plant conditions and vendor documentation. Section of the .
report which describe the deficiencies are provided below: l i
During the week of March 4, the inspector. performed a walkdown of the Unit 1 Containment Spray System. This consisted of a review of the l procedures and engineering drawings and verification of current system i alignment.The following discrepancies were noted: '
Sensing. element was identified as PT-07-3A, rather than PIS-07-3A appearing on CWD 362 i
Sensing element was identified as PT-07-3B rather than PIS-07-3B l appearing on CWD'362 Setpoint of 10 seconds after OPEN signal was not shown on CWD 289.
CWD showed alarm after 15 seconds.
Setpoint of 10 seconds after OPEN signal was not shown on CWD 289.
CWD showed alarm after.15 seconds.
V07223, V07230 and V07233 were not shown as Locked Closed.
V07152 IB SDC HX Outlet vent was installed at the bottom of the pipe, thus appearing unable to vent the pipe. In contrast, V07148 1A SDC HX Outlet vent was' installed on the top of the pipe.
During the week of March.11, the inspector ' performed a walkdown of the Unit 2 Containment Spray System. This consisted of a review'of procedures and' engineering drawings and verification of current system alignment. The foll.owing discrepancies were noted:
2998-G-088, " Flow Diagram Containment Spray and Refueling Water' Systems," Rev 23, Sh 1
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V07101 (86) and V07106 (D4) were shown as Closed instead of Locked j Closed.
V07334 (F5) and V07335 (G5) were shown as Open instead of Locked Open.
" LOW LEVEL STOPS PUMP & CLOSES VALVE" (F1 and F3) was incorrect.
The note should have read " LOW-LOW LEVEL STOPS PUMP & CLOSES VALVE."
2998-G-088, " Flow Diagram Containment Spray and Refueling Water
. Systems", Rev 23, Sh 2 V07390 through V07393, V07191 and V07163 (C2-3, D2-3) were shown ;
as Closed instead of Locked Closed. '
OP 2-0010123, Rev 69, " Administrative Control of Valves, Locks and Switches" ,
The licensee documented the inspector's findings on STAR 951515 and PMAI 03-402 and 403. The number of findings suggested that a potent.ial configuration control weakness existed. In an effort to established whether or not programmatic weaknesses rise to a level requiring enforcement, the inspectors. expanded the scope of the detailed walkdowns to include the Intake Cooling Water System of both units. At the close of the inspection period the ,
reviews were not complete. Accordingly, the issue will be tracked as an l unresolved item (URI 96-04-05, " Configuration Control Management").
CONCLUSION 3:
Based'on the information provided, this allegation is substantiated.
CONCERN 4: CI stated that there was poor communication between organizations Poor communication between organizations were identified in several inspection reports. The reports which identified problems in communications were inspection report (IR) 95-01, 95-07, 95-10, 95-18, and 95-99.
IR 95-01 identified weak communication between operators and other plant organizations. Weak communication between operators and other plant organizations was-noted with respect to Unit I hot 1eg temperation ;
stratification. The inspector concluded that the resolution to this issue was impeded by a number of communications failures between operators, .
Operations management, I&C, and JPN Nuclear Fuels.
IR 95-07 identified poor communications between electrical and mechanical ~
maintenance organizations and Operations. The poor communication led to confusion on.the status of both the mechanical and electrical NPW0s status.
1 IR 95-18 identified an event in which a technician from the safety department '
' reported to the control room, and due to poor communications, entered the H&V room through the blocked access door. The technician entered the H&V room
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alone using a handheld oxygen meter sampling at neck level.- The Operations Manager ordered the technician out the H&V room and posted a sign on the access door stating, "No entrance without prior NPS approval". A followup survey of the H&V atmosphere using a freon detector confirmed that freon , ;
levels were within an acceptable range. Maintenance reported that the 3-C i freon safety valve had apparently lifted based on low system pressure and that both the 3-A and 3-B units were available. Unit 2 then exited TS 3.7.7. A -
i followup discussion by the inspector with the Operations Manager confirmed that the safety department technician's actions were inconsistent with plant '
procedures and that the technician had been counseled on this item. The .
licensee is evaluating a procedural enhancement which will either fully open the ICW discharge valves.when flow is reestablished during the test or increase the number of turns throttled open to preclude recurrence of this .
. problem. This issue was compounded by identified weaknesses in communications across organizational interfaces, in that failures in informal communications were not compensated for by programmatic methods.
IR 95-10 identified poor communications between engineering and maintenance.
The poor communication resulted in the improper .instulation of the Woodward Governor on the'1B Emergency Diesel Generator. PWO #65-0984 was, issued to replace the Woodward Governor on the 182 Diesel Engine per MP 1-EMP-59.02, "1B '
Emergency Diesel Electric Periodic Maintena'nce and Inspection."
A replacement governor was procured from central stores and bench tested prior to installation. Part of the bench test involved an inspection of the condition of the limit switches for the governor's electric drive motor. All bench inspections were satisfactory. The replacement governor was installed late in the evening of May .17,1995, and EDG 1B started. At this point, due to difficulty in adjusting speed, an inspection of the installed replacement 4 governor discovered that the limit switches were defective, i.e., the internal
! spring in the microswitch did not extend the plunger when off-cam. -The defective limit switches were replaced and a second run performed the morning of May 18, 1995. Instead of starting at the idle speed of approximately 450 rpm, EDG 1B increased speed to approximately 900 rpm. The cognizant engineer quickly determined that the indexing of IB2 replacement governor was not
! correct. In a discussion with the inspectors, the cognizant engineer identified the cause as poor communication to the electricians who installed the replacement governor. He had instructed the electricians that the splined
, shaft could only be installed in one position based on a machined flat on the
! spline shaft, when in fact it could be installed in any position.
- IR 95-99, SALP, identifed weaknesses in communication. The report ' stated "The ability of Operations to identify and correct problems in a manner sufficient
- _to prevent recurrence was also of concern. -This issue was compounded by
- identified weaknesses in communications across organizational interfaces, in
, , that failures in informal communications were not compensated for by
- programmatic methods".
1 CONCLUSION 4:
Based on the information provided, this allegation is substantiated.
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.'St . Lucie' Inspection' Report 50-335/96-09 Dat'e: 06/15-24'/96 .
A. Inspector: _M.. Miller Branch-Chief Concurrence: ///' '~
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B.. 1.0 Persons' Contacted:
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1 J. Barbieri,. Design Engineering. '
W.' Bladow,. Quality Manager D. Denver, Engineering Manager
.:L.'Hiegel, I&C Supervisor J K. Mohindroo, Chief Site Engineer ;
R. Olson, I&C Department-Head !
J.-Scarola, Plant General Manager '
E. Weinkam, Licensing Manager . .;
D. Wolf, Production Engineering Group Manager III Enaineerina l El ~ Conduct of Engineering ,
E1.1 Nuclear Instrumentation Modification ,
'a . Insoection Scone (37550).
The inspector reviewed the concerns. listed in the six \
Quality Assurance reports (CR) related to the ,
i- implementation of plant change / modification PC/M 009- i
, 195 for the Nuclear Instrumentation System.'
Backaround.-One specific report, CR 96-1358, "Possible
- Loss'of Design Control", for the WI System, dated June 12,~1996 indicated there were significant problems.- CR 96-1358. stated there have been problems encountered during implementation which have required design '
changes,. numerous deviations-from the approved test
- procedure,-and key individuals are no longer on the
- project. .In addition, the work package had become voluminous and unwieldy with trouble shooting activities. There were 13 work package scope changes i and approximately 40 deviations to.the test procedure.
Trouble shooting was further complicated by not having vendor; engineering support on-site. CR 96-1358 is l- discussed below in the section Conditions Recorts.
The inspector's work scope. included an examination of.
. 4 all aspects of the.NI System modification-' including a review of the design package, design changes, deviations, . testing procedures, test procedure changes, logs, drawings,-work orders, equipment specifications, memorandums,. design procedures, administrative y '
procedures,FSAR Chapter 7.2.1, and Technical Specification'3.3.1.1. and 3.9.2. The inspector
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) '. initiated.walkdowns and observed ongoing work.- In-addition,~ discussions were held with vendor engineering ;
personnel concerning the ongoing problems. The items ;
in the-work: scope were performed to verify 1that the j modification was being performed within the 1- requirements of the licensee's. program and NRC .
I requirements.
L 'b. Observations and Findinas
, Plant chanae Modification (PC/M)009-195 l Nuclear Engineering completed the des'ign as REA/ Project
- SLN-94-025-11, "RPS NI Drawer Replacement", dated *
. February.27, 1996. It.was. released as plant change / modification PC/M 009-195, Revision 0,
" Replacement Of The Neutron Flux Monitoring And Protective System (NI Drawers) For The RPS System".
t The PC/M's purpose was to upgrade the Unit 1 NI System L with similar instrumentation that was' installed.in Unit 2 NI System during the last Unit 2 outage. The design ;
change included the.following modification for the four -
NI channels-A, B, C, and D: -
- 1) Replace the four existing Gamma-Metric. wide range excore detectors and cables with improved dual fission chad er assemblies.
- 2) Replace the four existing amplifiers located in ;
containment with new assemblies. [
- 3) Replace the eight existing RPS NI drawers located in the Control Room with the new NI instrumentation.
PC/M'009-195 was being implemented under work order WO No. 96007751. . The work order package consisted of PL\M ,
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09-1950 and.the Pre-Operation (Pre-Op) Test Procedure
-Number 1-1400280, " Functional Testing of PC/M 009-195
. Safety Channel". The inspector verified that the WO was being' implemented under administrative procedure ADM-0010432, Revision 3, " Control of Plant Work Orders". In addition, PC/M 009-195 met the requirements in the following Quality Instructions (procedures) for Nuclear Engineering and the equipment specification:-
1)- ENG-QI 1.0, Revision 3, " Design Control"
- 2) -ENG-QI 1.1,' Revision 0,." Engineering Packages" l
. :3)- .Eng-QI'1.2, Revision 1, " Minor Engineering
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- 4) ENG-Q1 1.3, Revision 1, " Drawing Change Requests"
, 5) SPEC-IC-004, Revision 0, " Equipment Specification For The RPS Nuclear Instrumentation System" The inspecto.r reviewed and verified that the factory.
acceptance test procedure, GAMMA-METRICS Test Procedure RMSP Assy No. 201663 agreed with Pre-Op 1-1400280 Test Procedure. Drawings JPN-009-195-001,to 017', "Out-of-Core Neutron Detectors" and " Nuclear' Instrumentation &
Reactor Protection System, were reviewed and verified that'the modification was being implemented according to design drawings.
Scocina Chances
. Thirteen scoping changes were reviewed to determine if the changes were controlled and within the requirements of the modification program and procedures. The scoping. changes [ Change Request Notice (CRN)] reviewed are listed below:
- 1) CRN-6100 - Enhanced install'ation instructions and i details'for using existing c l
- 2) CRN-6155 - Corrected a vendor drawing error.
- 3) CRN-6170 - Vendor approved voltage adjustment change f, rom 15 VDC to 15.5 VDC for 18 gauge cable. l l
4)' CRN-6193 - Changed relay operating position. I 1
- 5) CRN-6104 -
Minor installation change to eliminate interference between new drawers.
- 6) CRN-6094 -
Changed length of rope used for pulling j detector cables. 1
- 7) CRN-6132 - Revised connector clamp for triaxial cable. Superseded CRN-6130
- 8) CRN-6130 - Provided tolerance for connector. clamp.
- 9) CRN-6091 - Modified mounting tabs for amplifier boxes. (Note - The inspector verified welder qualification for this change).
- 10) CRN-6254 - Provided additional wide range -
calibration data.
- 11) CRN-6196 - Provided new data to change recorder's scale and FSAR Table 7.2-1.
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- 12) CRN-6078 - Removed hold points to allow .
implementation of work order. !
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- 13) CRN-6342 - Provided instruccions to install "C"- i channel drawer connector J6.
. The inspector did not have any safety concerns with the j scoping changes and did not consider the number excessive.
Deviations From Pre-OD Test Procedure The inspector reviewed 52 " deviations" from the pre- j operational test procedure. The inspector verified all 52 deviations were approved by the Facility Review Group (FRG) which was the licensee's independent safety evaluation board. Most of the deviations were minor in ,
nature and had no safety significance. All the i deviations concerning design had been approved'by Gamma-Metrics or Engineering. More than several deviations resulted as the consequence of a minor modification where "B and D" ' detectors were connected to "A and C" NI drawers to facilitate fuel loading.
This minor modification was implemented at the requect of Fuel Engineering. The inspector did not consider the problems encountered or the number of deviations to be excessive in this area as long as they were reviewed by FRG. However, the inspector determined the method i of implementing deviations was cumbersome. )
Wide Rance NI Temocrary System Alternation l
The inspector reviewed the minor modification package j and safety evaluation JPN-PSL-SEIS-96-028, " Wide Range j NI Temporary System Alteration" approved May 24, 1996. i The purpose of this minor modification was to provide for the temporary installation of coaxial jumpers cables to. allow the connections of wide range NI j detectors No. 2 (channel B) and No. 4 (channel D) to i the preamplifiers inputs for channels A and B respectivelyi This modification accommodated the fuel loading ,
analysis which required that the fuel locations adjacent to detectors "B" and *D" be loaded first for fuel shuffling and to meet Technical Specifications (TS).
TS 3.3.1.1. provides the requirement for. reactor protective instrumentation and TS 3.9.2 provides the requirement for refueling. This minor modification required several " deviations" to the Pre-Op testing.
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5 The inspector reviewed this minor modification to verify it met.the requirements in procedure ENG-QI 1.2,
" Minor Engineering Package" (modification) including the safety. evaluation. No concerns were identified.in
, this area.
Condition ReDorts Six Condition Reports (CR) were initiated by Quality Assurance identifying concerns, problens, and non-conforming conditions with the Unit 1 NI Instrumentation System modification. The seventh CR was initiated as the result of it being identified by the inspector during a walkdown of the control room NI
. cabinets and work observation. Each CR was reviewed by the inspector to determine if the concern or condition was properly evaluated and appropriate corrective action initiated. The seven CRs are listed as follows:
- 1) 'CR 96-1358 - This 10 discussed above in E1.1.a
" Work Scope". It is also discussed below in
. detail since it was considered significant.
- 2) CR,96-711 - The amplifiers boxes had.their mounting tabs welded without' removing the electronics. Gamma-Metrics stated no damage was expected due to the design of mounting the printed circuit board on standoffs.
- 3) CR 96-1443 - The 120 VAC power wire was damaged during removal of the "C" channel wide range drawer. Wire needs to be repaired.
4)- CR 9.6-1480 - The triaxial cable for connector J6 on'"C" channel dra~wer is very stiff and causes connector to break. The connector cable assembly was rebuilt using CRN 96-6342 twice. The work for the first rebuild not satisfactory.and the connector had to be redone.
5). CR 96-1434 - Several instances of non-compliance with requirements in specification SPEC-IC-004 have not been met. The preamplifier power cable (J3) was 18 gage instead of 16 ' gauge. This required the actions in CRN 96-6170 (listed below). The existing wide range signal cables between the amplifiers and containment
' penetrations was listed as coaxial cable on
' drawing 8770-6390. Instead it was found to be triaxial for the original installation. CRN 96-6342:was' initiated to install new connectors.
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- 6) CR 96-1464 -' Data from ECAD concerning "C" channel j source' range detector indicated low impedance during testing.- Gamma-Metrics was consulted for an evaluation and solution for corrective action. :
7). CR 96-1,481 - The. inspector identified that cable support clamps were missing from four connector on 1 each.NI drawer in the control room. A long term -
solution for corrective action'was being initiated j by the licensee. "
The-inspector reviewed each CR and discussed its y concern with engineering for their evaluation and the
' corrective action. The inspector verified that ,
problems did exit. However, the problems were not-considered significant and no safety concerns were ;
identified. .
CR 96-1358 l The conditions listed in CR 96-1358 (CR) and the !
inspector findings (IF) are listed below:
- 1) CR - There was possible loss of design control.
IF - The inspector did not identify any loss of )
design control. However, coordination between design engineering and I&C was initially weak.
The initial design engineer assigned to the NI system. project left the-site. He was later replaced with a design engineering manager.to coordinate the project. 3
- 2) CR - Key. individuals who had been involved with the design and implementation are no longer available.
IF - Two of the three individuals initially assigned to the 20t project were no longer available. One individual left the site and the other suffered a serious' illness. However, both engineering and I&C had a sufficient number of qualified' engineers capable of completing the modification. - The problem was,not considered a loss of: key personnel, it was more a lack of coordination;since the I&C supervisors were
. performing more than function. The I&C group in the Maintenance' Department performs several' -
'2unct. ions that are unique to it. First level- 4
- supervisor's duties include being a craft foreman,.
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l system engineer, installation l engineer,' test?
engineer,. trouble shooter and etc. In several
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( 1 instances, the IEC supervisors were overwhelmed I
- with work due to the refueling outage. ;
1
- 3) CR - There have been problems encountered during- )
implementation which have required changes to .
- . '" design" as well as numerous " deviations" from the FRG approved Pre-Op Test Procedure'and vendor
, technical specifications manual.
i
'IF - The inspector verified there: have been J
- problems encountered'during implementation and '
testing.. The problems encountered were mostly l with' cables ~and connectors during work. .
implementation. .The " deviations" were reviewed by
.! the inspector and approved by FRG (safety review group)'for the Pre-Op Testing.
l Many of the " deviation" and " scoping. changes" were written to incorporate corrective action and post maintenance testing after trouble shooting installation problems. This was id'entified by the
! inspector as a work control implementation
, weakness. This problem was corrected later by the i licensee when new work orders (WO) were written
- with the appropriate post maintenance testing.
For example, instead of initiating a " scope i- change" and " deviation" for trouble shooting,
- implementing corrective action, and testing, three
- ~
new WOs 96016391,-96016395, and 96016397, all dated June 21,.1996, were written.to " Repair MB NI
- Drawer". All.three WOs had one specific task with
{*- appropriate post maintenance testing.. All three
- WOs were approved by FRG. This method of using
. . additional WOs improved control and implementation j
! of the modification. I
-, i
[ 4) CR - The~ work package had become voluminous and l
- unwieldy.
- IF - The inspector agreed. The' licensee initiated corrective action by implementing an " engineering.
- . implementation / test log" and using additional WOs
- i. to control.the work as discussed above. .
5)- .CR --Trouble shooting-was'further complicated by
' I
- not having-vendor engineering support on-site. -
t ,
IF - A. Gamma-Metric's design engineer and a field engineer were on-site to support'the modification.
p The field engineer was scheduled to remain on-site -
F -
- until the. modification was completed- and turned '
~
over to Operations'. _, i i
1
- + , ,- , ~m.,.
i A
c., Conclusion The inspector concluded that there was justification for the comments stated in CR' 96-1358. IEC' personnel
. were sometimes overwhelmed:with problems and work.
However, there was no loss of design control. The IEC.
. supervisors and technicians were knowledgeable and technically capable of-implementing the NI-modification. The inspector considered the capabilities of I&C personnel to be a strength in-the
- Maintenance Department. l The inspector verified'that FRG (safety review group).
reviewed,all changes and deviations to ensure plant .
safety. Design engineering was made aware that I&C needed additional support and was in the process of :
providing.it. . Program weaknesses for implementing ;
changes and trouble shooting were identified and corrected'using new work. orders. The vendor, Gamma-Metrics, provided good engineering and field support with knowledgeable personnel. The inspector concluded the NI System modification was being completed in a satisfactory manner to ensure plant. safety. ,
D. Exit The Exit was conducted Monday 24, 1996. There were no :
dissenting comments. !
.E. 'NA F. Summary Statement The I&C group,.with e~ngineering support, was satisfactorily l implementing the Unit 1 NI modification after overcomming i initial. implementation and coordination problems due to the ;
outage' work load.
'G. Acronyms CR -
Condition Report
'CRN. -
' Change Request Notice FRG;
~
Facility Review Group. ,
I&C -: Instrumentation and' Control
.NI :- . Nuclear' Instrumentation PC\M - PlantLChange/ Modification Pre-Opy Preoperational test RPS -
- Reactor Protection. System.
. TS' .
Technical Specifications .
VAC -
Volts Alternating Current VDC -
-Volts Direct Current .
WO -
Work. Order 4 , w , -
=mev -- m - - e,, eve = "=-se ~ - - =d-it -r-.
- I w
e
^b .,
- 1 1
I i
- l 8'
1 ;
- H. NA i
i
' I.- NA' i
i i
I 9
9 8
I I
4 e
O 4
ST. LUCIE INSPECTION INPUT l .
A. Inspector: J. York Branch Chief I/ ^ - - f/6%
C.'C&sto Inspection Dates: May 13-June 28,1996 Report-Nos. 50-335, 389/96-12 B. Persons Contacted D. Denver, Engineering Manager D. Culpepper, Chief-Engineering Assurance, Nuclear Engineering C. Wasik, Nuclear Licensing Engineer, Nuclear Engineering C. O'Farrill, Supervisor of Nuclear Engineering C. Input for the Inspection Area 4.0 Engineering El Conduct of Engineering
- a. Insoection Scoc_e l The inspectors reviewed a sample of the licensee's safety evaluations (10 CFR 50.59) . The evaluations were reviewed for
- .hreshold for determining if an unreviewed safety question (USQ) existed because of an increase in the probability of a design basis accident occurr!ng, an increase in equipment malfunction, a reduction in the margin of safety, or an increase in radiation dose consequences. These evaluations were also reviewed for adequacy of screening and assumptions used for the safety evaluations,
- b. Observations and Findings The inspector.s reviewed eleven safety evaluations (50.59s) The eleven 50.59s were:
o Cracking of Westinghouse Alloy 600 Mechanical Steam Generator Plugs.
e Temporary Relocation of Class Break on Intake Cooling Water.
e Installation of Temporary Fire Penetration Seals in Pipe Barrier BN064.
e Temporary Installation of Strain Measuring Devices on the Pressurizer Relief Valve Discharge Piping.
- Safety Injection Tank (SIT) Discharge / Loop Check Valve Stroke Test-Unit 1.
e Freeze Seal Application for V3651 and V3652 on the 1B e
t k
Shutdown Cooling Return Line.
e Safety Evaluation For Boraflex Blackness Testing Results.
- Wide Range Nuclear Instrumentation Temporary. System '
Alteration.
e Temporary Configuration for CEDMCS (Control Element Drive Mechanism Control System) Cooling System and Enclosure, Unit 2.
e Safety Evaluation for Inoperable Fire Pump e St. Lucie Unit 1 Refueling Equipment Underload and Overload Settings.
Problems were identified with the last three items and the details are discussed in the following paragraphs.
- 1) Terporary Configuration for CEDMCS Cooling System and Enclosure On June 4, 1996, a control room annunciator indicated that an undervoltage condition existed on the Control Element Drive Mechanism Control System (CEDMCS). Operations responded to the CEDMCS equipment and noted that the CEDMCS enclosure was approximately 11 degrees warmer than normal. This enclosure is located in the cable spreading room on the 43 ' foot elevation of the reactor auxiliary building.
Following this event, an STA In-House Event Report and Condition Reports '96-1238, 96-1245 and 96-1325 were issued.
Some of the following items with appropriate plant corrective action tracking number were identified by these reports:
CEDMCS enclosure and air conditioning units did not appear on the plant's controlled drawings. (STAR 951320 )
CEDMCS enclosure air conditioning units were not seismic qualified. Final design was in process to provide seismic restraints f or the air condition units. (PM 96-06-208)
As part of the action for Condition Report 96-1325, a.10 CFR 50.59 safety evaluation was performed on the CEDMCS enclosure.
The evaluation found that this air conditioned enclosure was erected in the early 1980's during the pre-operational testing phase. This testing found that the CEDMCS enclosure required an air conditioned' environment to prevent overheating of the four CEDMCS cabinets. The licensee's review determined that the design of the enclosure was acceptable, except that the air conditioning units and one air conditioning duct presented a hazard to safety related equipment in a seismic event.
Therefore, seismic supports and restraints were provided for
0 l
the air conditioning units and duct prior to the unit's restart on June 13. !
The inspector reviewed the 10 CFR 50.59 evaluation provided for the design and installation of the seismic restraints and justification of the installation of the CEDMCS enclosure.
This air conditioned enclosure was erected during the pre- 1 operational test phase in the early 1980's to provide cooling for the CEA system. However, a 10 CFR 50.59 review was apparently not performed when the enclosure was originally erected. The CEDMCS .was described in the UFSAR but the cooling system and enclosure for the CEDMCS .were not described in the UFSAR. This was identified as another example of URI 50-335, 389/96-04-09, " Failure to Update UFSAR".
The failure to perform an evaluation as required by lO' CFR 50.59 prior to making a change to the plant as described by the UFSAR is an apparent violation (EEI 50-389/96-12-XX, ;
" Failure to Perform a 10 CFR 50.59 Safet.y Evaluatioll for CEDMCS Enclosure)." i
- 2) Safety Evaluation for Inoperable Fire Pump During the Spring 1996 Unit i refueling outage, one of the two Unit 1 EDGs had been placed out of service to perform ,
maintenance and modification work activities. Only one EDG l was in service to provide. power in the event of a loss of l power event. To prevent a possible overload on the single EDG unit, a number of breakers to various components were opened and the units 480V electrical busses were crosstied in :
accordance with OP 1-0910024, Rev 6, "Crosstying/ Removal of I 480V Buses." One of the components removed from service was j Fire Pump 1B. The breaker to this fire pump was opened on May 21, and this pump was removed f rom. service and remained out of ;
service on June 8.
l AP 1800022, Rev 16, " Fire Protection Plan," Appendix A, Sections 2 .~2 and 2.3 required two fire pumps rated at a capacity of 2300 gpm to be operable at all times. Appendix A Section 4.1.A stated that with one of the two fire pumps inoperable, restore the inoperable equipment to service within seven days or provide an alternate backup pump within the next 30 days.
Fire Pump 1B had been out of service for 18 days. The compensatory measure established for this pump being out of l service was the installation of a portable gasoline engine -
l drive pump rated at 750 gpm. This pump had been connected to take suction from the fire protection water storage tank for Fire Pump 1A. This alternate pump was not of the same c'apacity as one of the two required pumps and a justification
.was not provided to demonstrate that this pump was of adequate ,
capacity to meet the maximum fire flow requirement for the safety related araas of the plant., The licensee initiated a 9
z .)
1 CR to review this item.
The licensee informed the inspector that the out of service pump could be restored to operability by restoring the existing open breaker to the closed position. Also, the 30 day time to provide an alternate backup pump had not been exceeded. This met the requirements of AP 1800022 for dne pump being inoperable.
Resolution of the Condition Report (CR 96-1356) indicated that the installation of the portable fire pump as the comp.ensatory measure with one of the permanently installed fire pumps out of service violated the fire protection configuration as described in the UFSAR'. An engineering evaluation should have been prepared to justify and document the temporary This is an apparent violation (EEI 50-335, configuration.
389/96-12-YY, . " Failure to Perform a 10 CFR 50.59 Safety Evaluation For Use of a Temporary Fire Pump".
- 3) Refueling Equipment Overload and Underload Settings Condition Report 96-812 was issue'd on the last safety evaluation (number SEFJ-96-020) by the licensee. The report stated that an engineering eval'uation had been written to modify the overload and underload setpoints described in the UFSAR without. performing a 50.59 safety analysis / evaluation.
These overload and underload load cell setpoints provide a margin to account for resistance encountered while lifting or lowering fuel assemblies and prevent exceeding the fuel assembly and refueling equipment design loads.
The licensee had obtained information from the vendor for use in this Unit i refueling outage which would allow an increase in hoist interrupt from 10 percent of the weight of a fuel assembly to 18 percent (approximately 200 pounds). The original engineering analysis did not take into account that these changes in setpoint values would affect the UFSAR and thus the CR was written.
St. Lucie Quality Instruction (QI) 2.0, " Engineering Evaluations , " Rev. 1 dated January 31, 1996, provides general requirements and guidance for the development and processing of engineering evaluations. This procedure references QI 2.1, "10 CFR 50.59 Screening / Evaluation," Rev. 1 dated March 30, 1996, which states in part that the screening process is designed to determine whether the activity requires a complete 10 CFR 50.59 by asking a series of four questions. One
. question, "Does the change represent a change to procedures as described in the SAR?" should have been answered "yes" in the
- case of the original engineering analysis. The procedure also states that, "A positive response to any of the first four
...... questions requires a 10 CFR 50.59 evaluation."
The FRG, the site safety commit' tee, noted that a safety
c I
.:e i
evaluation was not present with the requested procedure change and returned the procedure to the engineering group for i correction and the CR was written to identify the problem. - .
This failure to perform an evaluation as required by 10 CFR
.50.59 prior to making a change to plant procedures described in the FSAR is an apparent violation (EEI 50-335/96-12-ZZ,
" Failure to Perform a 10 CFR 50.59 Safety Evaluation For Change in Setpoints Listed in FSAR").
- c. Conclusions on Conduct of Encineerina Three of the eleven 50.59 safety evaluations'or conditions requiring safety evaluations examined were found - to have apparent violations of NRC requirements and were being considered for escalated enforcement.
G F
4 9
4 4
/
'Mv/s ST. LUCIE REPORT INPUT, REV. 1 A. Inspector: J. York Branch Chie CW C. Casto Inspection Dates: June 17-28,1996 Report Nos. 50-335, 389/96-09 B. Persons Contacted D. Denver, Engineering Manager C. Input for the Inspection Area 4.0 Engineering El Conduct of Engineering
- a. Insoection Scope The inspectors reviewed the functions of the current engineering organization and the anticipated changes after downsizing takes place on August 1, 1996.
- b. Observations and Findings The inspectors discussed with engineering management the l current engineering organization, the functions of each of the i groups, and the approximate number of people in each group. l The licensee then discussed the changes in groups, functions I and numbers that would take place on August 1, 1996. I Currently the St. Lucie site is supported by an additional engineering organization located within Corporate Engineering to Juno Beach, Florida. Most of these Corporate Engineering i functions are being transferred to the St. Lucie site (and to l I
the Turkey Point site) and Corporate Engineering is being downsized. To follow and evaluate any effect of the downsizing on the engineering function at St. Lucie on a periodic basis the inspectors will use inspection procedure IP 37550, Engineering. The objectives of this procedure are,
" Evaluate the licensee's engineering activities, particularly
, the effectiveness of the engineering organization to perform routine and reactive site activities including the identification and resolution of- technical issues and problems". As a basis for future monitoring, the inspectors discussed current engineering work load, i.e., open Modification Packages, Condition Reports,etc.
O u
w -
o INPUT TO INSPECTION REPORT 96-11 for ST.LUCIE -
Conducted July 8 - 12, 1996 Inspector: d / I N b - Branch Chief Concurrence:
" '" '/ d /r # EXECUTIVE suNNAny Enaineerina s
Engineering support and the maintenance program for the control element drive mechanisms and related controls and instrumentation were inspected. The inspection of this area was prompted by an increase in dropped control ' element events in 1996. The inspectors concluded that engineering support and maintenance for the system was good. The licensee has established a team to analyze the reliability of the system and propose upgrades or replacements if data trending indicates that aging is becoming a factor in system performance. Sections E2.1 and M2.1.
III ENGINEERING E2 Engineering Support of Facilities and Equipment E2.1 Control Element Drive Mechanisms and'Related Control and Instrumentation.
- a. Inspection Scope (37550) l There were four control element assembly (CEA) drop events in 1996, three on Unit I and one on Unit 2. The CEAs are one of the systems which are used to control core reac'.ivity through insertion or withdrawal of absorption rods. The scope of the inspection was to review the cause determinations and corrective actions for these probl ems ~.
Historical information on dropped CEA events and system upgrades were reviewed. Maintenance activities were also reviewed, and that portion of the inspection is covered in Section M2.x.
- b. Observations and Findinas On Unit'1, the control equipment for the control element drive mechanisms consisted of discrete electronic-components. On Unit 2; the control equipment was a later version consisting of integrated circuits, and was referred to as the Advanced Control Timing Mechanism. Unit 2 i control equipment incorporated a feedback (or checkback) circuit whi.ch had the capability of blocking all movement signals;if an error is ,
, detected. The CEAs have been operated-in manual mode only.
Dropped'CEA events that occurred between the beginning of 1993 and the time of this inspection are summarized below. '
y ,
j DROPPED CEA EVENTS-JANUARY 1993 TO JULY 1996 f
Unit' CEA No. QLle Cause
. i 2 12 5/24/96 Unknown, suspect bumped fuse.
'1- 1~ 3/4/96 Operator error, quick release of bypass switch.
1 47 2/23/96 Loose connection in interconnecting wiring.
1 20 2/22/96 Failed (i.e. shorted) silicon controlled rectifier (SCR) in A phase upper gripper switch-module cau4Dg fuse to open.
1 63 i 11/1/93 Timer card not seated. ~
^
1 3 8/26/93 Unknown.
2' -
5/21/93 Ground faults at containment penetration coupled with failed circuit breaker. Seven CEAs dropped.
In all but two cases, the root causes for the dropped CEAs were determined and corrected. In the two cases where the root cause was not definitely established, a momentary perturbation in the timing sequence was suspected,
,and those CEAs have operated correctly thereafter.
Review of ~ statistics on dropped CEA events and their causes would indicate that the Unit 2 system has been more reliable than Unit 1. Since at least 1990, with the exception of.one unknown cause, the Unit 2 dropped CEA events .!
were caused by problems outside the boundary of the control hardware. The '
inspector concluded that these. problems would not be expected to recur in the r future. -These facts support the' licensee's contention that the Unit 2 CEA-L System has been highly reliable. However, this conclusion based on St. Lucie 3 i
specific data was in conflict with data obtained from a broader base. Data ;
L gathered by Combustion Engineering on dropped CEA events at all plants where i CE was the NSSS' supplier would indicate that the Unit 2 class of equipment has not been as reliable as the Unit 1 class equipment. !
r The multiple dropped.CEA event on Unit 2 in May 1993 was caused by insulation E ;
' breakdown of multiple conductors at the containment penetration together with
~
a failed CEA circuit breaker. The penetration problem appeared to be an.
isolated case. The circuit breaker that failed was a four pole breaker. The j (failure mode of the circuit breaker was failure of two poles to open, which resulted.in tripping of the upstream sub-group circuit b,reaker. l
'Overall, the known causes of.the dropped CEA events listed'in the -above table
, are all different.- Therefore, the data does not indicate any negative '
component;nor personnel related trend. I The only modification implemented to upgrade the reliability of the CEA System '
has been the changeout of the CEA 15 VDC power supplies with power supplies built to a stricter specification. This modif,1 cation has been completed on-Uni _t 1 during.the current outage. Also;at Unit 1, cables which connect to the
h
) .
CEAs themselves (referred to as head cables) were replaced during the current ,
outage, because the prefabricated connectors were failing at an increasing rate.
The licensee has established a team to analyze the CEA System reliability.and propose upgrades if warranted. The team had a target date of mid-September for issuance of their report.
V. MANAGEMENT MEETINGS X1 Exit Meetino Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 12, 1996. There is no proprietary informationcontained in this report.
Partial List of Persons Contacted Licensee *
- W. Bladow, Quality Manager ,
I
- D. Denver, Engineering Manager
- D. Howard, I&C Maintenance Engineer
- S. Lavelle, Licensing Engineer D. Maley, I&C Instructor
- J. Marchego, Maintenance Manager T. Neuhouse, Unit 2 I&C Maintenance Supervisor
- R. Olson, I&C Maintenance Manager
- J. Scarola, Plant General Manager R. Sherman, Unit I I&C Maintenance Supervisor
- J. Stall, Site Vice President
' Other Oraanizations A. DeGrass.e, Site Representative for Combustion Engineering, Inc ;
Inspection Procedures' Us'ed 37550 Engineering 9
4 6
e
. , -