IR 05000373/1997003: Difference between revisions

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{{Adams
{{Adams
| number = ML20140B685
| number = ML20199K127
| issue date = 05/20/1997
| issue date = 11/25/1997
| title = Insp Repts 50-373/97-03 & 50-374/97-03 on 970207-0321. Violations Noted.Major Areas Inspected:Licensee Operations, Maint,Engineering & Plant Support
| title = Ack Receipt of 970613,0827 & 0929 Ltrs Informing NRC of Steps Taken to Correct Violation of Insp Repts 50-373/97-03, 50-374/97-03,50-373/97-06,50-374/97-06,50-373/97-07 & 50-374/97-07 Issued on 970520,0728 & 0829
| author name =  
| author name = Vegel A
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| addressee name =  
| addressee name = Subalusky W
| addressee affiliation =  
| addressee affiliation = COMMONWEALTH EDISON CO.
| docket = 05000373, 05000374
| docket = 05000373, 05000374
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-373-97-03, 50-373-97-3, 50-374-97-03, 50-374-97-3, NUDOCS 9706060291
| document report number = 50-373-97-03, 50-373-97-06, 50-373-97-07, 50-373-97-3, 50-373-97-6, 50-373-97-7, 50-374-97-03, 50-374-97-06, 50-374-97-07, 50-374-97-3, 50-374-97-6, 50-374-97-7, NUDOCS 9712010021
| package number = ML20140B670
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 2
| page count = 80
}}
}}


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Mr. W. T. Subalusky, Jr.
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U.S. NUCLEAR REGULATORY COMMISSION


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i Site Vlos President LaSalle County Station Commonwealth Edison Company 2001 North 21st Road Marseilles,IL 61341 SUSJECT: . NOTICES OF VIOLATION (NRC INSPECTION MEPORT NOS, 50-373/97003(DRP), 50 374/97003(DRP); 80 373/97006(DRP),
REGION lli
50 374/97006(DRP); AND S0 373/97007(DRP),60 374/97007(DRP))
:
  - Dear Mr. Subalusky.
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,  Docket Nos: 50-373,'50-374 License Nos: NPF-11, NPF-18  '
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4-Report Nos: 50-373/97-03, 50-374/97-03  l
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Licensee: Commonwealth Edison Company
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Facility: LaSalle County Station, Units 1 and 2 Location: 2601 N. 21st Road   !
Marseilles, IL 61341 r
Dates: February 7 - March 21,1997  )
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Inspectors: M. Huber, Senior Resident inspector J. Hansen, Resident inspector l
C. Mathews, Illinois Department of Nuclear Safety <
Approved by: Marc Dapas, Chief, Projects Branch 2 Division of Reactor Projects
 
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9706060291 970520    l PDR ADOCK 05000373    i O  PDR
 
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    . EXECUTIVE SUMMARY
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LaSalle County Station, Units 1 and 2
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NRC inspection Report 50-373/97-03(DRP): 50-374/97-03(DRP)  {
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^  This inspection included aspects of licensee operations, maintenance, engineering, and  !
plant support. The report covers a six week period of inspection activities conducted by  l l  the resident staf I (        ;
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Licensee performance during this inspection period was characterized by human  j j- parformance errors, inadequate procedures, missed Technical Specification surveillances,  l
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and additional examples of previously identified problems with performing design changes  ;
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outside of the modification process. In one instance, licensee management identified  i maintenance process problems in the General Electric SBM (switchboard, miniature) switch  '
, . replacement project and decided to stop the work to correct the problem i
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t i  Plant Operations j
l        l l- *  The inspectors identified a violation for failing to notify the NRC of the permanent ,
reassignment of a licensed individual to a position which did not require a licens j
;  The inspectors also identified that the status of licenses could not be readily verified j
;  by the shift manager, informal communications were used to inform site personnel i
:  of changes in license status, and lists used to control licenses were inaccurate. No ;
;  instances were identified where an operator inappropriately assumed a licensed  l j-  position. (Section 01.2)    '
 
e' ~ While operators generally followed procedures, the inspectors identified two I examples of procedural violations while evaluating emergency diesel generator testing. One instance invalidated a completed surveillance and required an
;  additional run of an EDG. In addition, poor procedures caused delays in the EDG j-  testing evolutions observed by the inspectors and unnecessarily challenged 4  operators. (Section'02.1)    !
  *  The inspectors observed the licensee's response to a high lake level which was
,  above the flood level analyzed in the Updated Final Safety Analysis Report
_ (UFSAR). Plant personnel responded adequately to the event once the problem was -
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identified. Howeveriseveral conditions contributed to the event including poor
,  material condition of the lake'make-up and blowdown systems, operator i  complacency regarding operation outside of normal procedure bands, and incorrect
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acceptance criteria in the operator rounds procedure. The problems leading up to
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the Unusual Event, which resulted from the high lake level, indicated that operators
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were still not identifying plant problems and demanding their resolution. Incorrect
. acceptance criteria in the operator rounds procedure was considered a violatio (Section 02.2)
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A Technical Specification violation was identified by the inspectors for the failure to perform required testing of the 1 A residual heat removal (RHR) pump shutdown ,
cooling suction valve,1E12-FOO6A, within the required test interval. Operations '
personnel lacked attention to detail when reviewing the surveillance procedure l (Section O2.3)    I
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The inspectors observed an operator removing a breaker from service without having the required procedure available at the work location. This was an additional example of the violation issued in NRC Inspection Report 96018 for failing to follow procedures (50-373/97018-02; 50-374/97018-02). The inspection also revealed that operators did not know the expectation regarding the use of
" reference use" procedures. (Section 03.1)
Maintenance i
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The inspectors observed post-maintenance testing for the 1 A EDG. Two equipment )
deficiencies were identified by the inspectors and the licensee during the diesel testing. Rework of the Jacket water cooling heat exchanger, service water flange gasket which had been replaced while the diesel generator was out-of-service, ,
resulted from inadequate maintenance practices. In addition, the licensee identified I that inappropriately sized fuses had been installed in the EDG ventilation exhaust :
damper control circuitry. (Section M1.1)    !
* Licensee management stopped work associated with the General Electric SBM (switchboard, miniature) switch replacement project. Although this was considered an appropriate action, the deficiencies that resulted in the stop work order indicated that previously identified weaknesses within the licensee's maintenance processes continued to exist. (Section M1.2)
* Instances of poor housekeeping were identified by the inspectors during tours of high radiation and high contamination areas. Most of the poor housekeeping was related to maintenance work practices and not cleaning the area after completing work. (Section M2.1)
* The inspectors identified a violation regarding work instructions for inspecting the 1 steam tunnel check dampers that did not contain appropriate qualitative or quantitative acceptance criteria. (Section M3.1)
* Required surveillanca testing of the Unit 1 RHR pump 1 A discharge high/ low pressure switch was not performed within the required time interval because a work control scheduler failed to follow the scheduling procecure. The inspectors determined that the failure to perform the surveillance testing was a Technical Specification violation. (Section M3.2)
Enaineerina
* The licensee reported to the NRC in Licensee Event Report number 96019 that two RHR system containment spray isolation valves had not been tested according to
 
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ASME Section XI requirements due to personnel error. The inspectors concluded  i that the actions taken by the licenses following identification of the missed
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surveillance test were appropriate. The inspectors determined that failure to test  ;
the valves was a Technical Specification violation. (Section E1.1)
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Two violations were identified by the inspectors regarding the licensee's failure to  f use the design change process to replace the cooling lake blowdown flow
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l instrumentation. In addition, subsequent calibrations of the instrumentation were not performed by the licensee using approved contractors or procedures. The fact  !
that the replacement and calibration of the flow instrumentation should have been performed using appropriate procedures was not recognized by the license !
  (Section E2.1)      !
 
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The licensee identified that a modification to the actuation logic for the main control l
room atmospheric control system (MCRACS) in 1993 created an unreviewed safety  l question. The licensee also identified that the modification to the MCRACS actuation logic created a condition where the system would not meet single failure criteria per the design basis described in the UFSAR. Two apparent violations were  ,
identified by the inspectors for the failure to perform an adequate design chang (Section E2.2)-
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* The inspectors toured several high radiation and high contamination areas with the  !
support of radiation protection technicians. The technicians were knowledgeable of the plant and radiation protection practices. (Section R4.1)
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Report Details  !
i Summarv of Plant Status    '
Unit 1 was in a forced outage for the entire inspection period and Unit 2 remained shut down for a refueling outage. The licensee continued to keep both units shut down to perform design basis configuration reviews and to address equipment and human performance problem Exercise of Discretion Four violations (or individual examples of violations) described in Sections 01.2, M3.1, E1.1, and E2.1 of this report are based upon licensee activities which were identified after but occurred prior to the licensee announcing, in December 1996, an extended shutdown of the LaSalle County Station. These violations satisfy the appropriate criteria in Section Vll.B.2, " Violations identified During Extended Shutdowns or Work Stoppages" of the " General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG 1600, and Notices of Violation (NOV) are not being issued for these particular violations. The violations described in Sections 01.2 and M3.1 of this report were not identified by the licensee and, while the violations described in Sections E1.1 and E2.1 were identified by the licensee, the identification was not a result of a i comprehensive program for problem identification and correction that was developed in response to the shutdown. However, the other criteria in Section Vll.B.2 of NUREG 1600 were mat, which allows enforcement discretion to be applied. Specifically, in reference to the four violations, enforcement action was not considered necessary to achieve remedial j  action, the violations would not be categorized at Severity Level 1, and the violations were j  not willful. In addition, actions specified in Confirmatory Action Letter Rlli-96-008B
;~ effectively prevent the licensee from starting up LaSalle County Station without implicit {
;  NRC approva j l        1
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:    1. Operatio.ng
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01 Conduct of Operations
 
i  01.1 General Comments (71707)
. The inspectors conducted frequent reviews of ongoing plant operations using
[  . Inspection Procedure 71707. Walkdowns were performed in the main controi j  room, emergency diesel generator (EDG) rooms, the auxiliary electrical equipment :
!  room (AEER), safety-related ' pump rooms, the reactor building including the drywell, l
. the turbine building, and the radwaste facility. The inspectors also observed and j  discussed plant status and pending evolutions with shift personnelin the control room.
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01.2 Ooerations Shift Staffino and Operator Licensina Proaram Weaknesses insoection Scone (71707)    l
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i (  The inspectors interviewed operators, operation's shift management, and operations !
j  ? and training staff personnel to assess the adequacy of administrative controls for I licensed operator staffing of shift crews. The inspectors also reviewed procedures j delineating the controls in place for operator licenses. The inspectors reviewed !
documents including:    l
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  * LaSalle Administrative Procedure (LAP) 200-3, " Shift Change," Revision 28
  * LAP-2OO-10, "NRC Operator License Active Status Maintenance and Reactivation," Revision 3
  * Administration and Course Management instructions (ACMI) for Licensed Operator Requalification Program, Revision 11 Observations and Findinos  I i
The inspectors identified several weaknesses during the review of shift staffing and j controls for operator licenses which included:    -
  * Shift managers could not read lly verify that operators were meeting administrative and license requirements when the operators assumed shift duties. While shift management ensured that required shift positions were manned, the status of the licenses of the operators in those positions could not readily be verified by shift management. The operations department timekeepers maintained records of the amount of time that operations employees worked and were also responsible for determining which operators met the requirements for maintaining an active license. However, the timekeepers were not ' required to provide shift management a copy of the list of active licenses. Without a list of operators with active licenses, operations shift management could inappropriately place an individual with an inactive license in a position requiring an active licens * Informal electronic communication (E-mail) was used to inform the timekeepers that an operator could not be scheduled for work in a licensed position due to medical problems or training issues. However, operations ,
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management was not notified when an operator was no longer qualified to work in a licensed position. The informal'means of communicating that an operator could not w'ork in a licensed position added to the possibility that an individual could be inappropriately assigned to an active positio ,
  * One senior reactor operator (SRO) who was licensed on January 22,1997, was not added to the list of operators who needed to be enrollod in the l requalification training progra * Tl e inspectors identified two problems with the licensee's list of licensed operators. In one instance, an individual was still listed as a licensed s
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operator but his license had been terminated by the NRC following receipt of a letter dated June 26,1996, requesting license termination. In the second instance, on May 23,1996, licensee management decided to relinquish the
,  license for an individual and the operator's name was removed from the list of licensed operators. However, the licensee did not request termination of the license from the NRC. The operator failed the written and operational
, annual requalification examinations in May 1996 and the licensee's  j Operations Training Review Board, in conjunction with the individual,  '
decided the license would no longer be required. The individual did not attend requalification training or perform any licensed operator duties following the decision. Also, his name was removed from the licensee's license tracking system. However, due to new personnel assigned in the training department and inadequate guidance in the ACMI for the Licensed Operator Requalification Program, the NRC was not notified by the licensee within 30 days that his license needed to be terminated, as required by 10 CFR Part 50 74(a). The failure of the licensee to notify the NRC within 30 days that a licensed individual had been permanently reassigned from a position where his license was required was considered a violation of 10 CFR Part 50.74(a).
 
The licensee initiated the fe" uing corrective actions after the problems were identified by the inspector * In a letter dated March 5,1997, LaSalle County Nuclear Station personnel )
notified the NRC that the individual who failed the written and operational requalification examinations would no longer require his license and, therefore, the license may be terminate *
The operations training department was planning to revise the Licensed Operator Requalification Program description to address the specific steps that need to be taken when a license is voluntarily relinquished or when it is terminated for cause. This revision would also address specific responsibilities of the licensed operator training staff and contain a check list for verifying compliance with operator licensing requirements. The licensee plats to complete the revision and conduct an orientation for the department administrative assistant and operations training staff by April 21,199 * The licensee planned to combine the requirements for terminating a license
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due to training issues and the requirements for license termination or  !
restriction due to medical reasons into one training department instructio l The instruction is intended to provide appropriate guidance on required activities and responsibilities. The licensee planned to develop the  !
department training instruction and conduct an orientation for the  l department administrative assistant and operations training staff by June 23, 199 I
* Training Department instruction 204, " Trainee Performance Review  l Process," was being revised to assure that the department administrative  !
 
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assistant was informed of pending review boards and to provide direction ;
regarding the administrative assistant's responsibilities following each board '
meeting. This revision was scheduled to be completed by April 1,199 l
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The inspectors verified that the licensee was following the progress of these items through their established tracking syste , Conclusions While weaknesses were identified with the licensee's administrative controls for  l licensed operator staffing of shift crews, the inspectors did not identify instances  ;
where an individual with an inactive license inappropriately assumed a shift position requiring an active license. The inspectors reviewed the licensee's corrective actions in response to this issue and consider them adequat Operational Status of Facilities and Equipment 02.1 Procedural and Ocarator Performance Deficiencies durina Emeroenev Dieue!
Generator (EDG) Testina Insoection Scone (71707)
The inspectors observed post-maintenance and surveillance testing for the 1 A and 2A EDGs and reviewed documentation from the most recent surveillance test performed on the O EDG Procedures reviewed by the inspectors included:
  *
LaSalle Operating Procedure (LOP) DG-02, ." Diesel Generator Startup and Operation," Revision 22
  * LOP-DG-04, " Diesel Generator Special Instructions," Revision 21
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LaSalle Operating Surveillance (LOS) DG-M1, "O Diesel Generator Operability Test," Revision 32
  *. LOS-DG-M2, "1 A (2A) Diesel Generator Operability Test," Revision 34
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LaSalle Maintenance Surveillance (LMS) DG-01, " Main Emergency Diesel Unit Surveillances," Revision 16 Observations and Findinas Planned testing of the 1 A EDG was delayed several times during this inspection period due to procedural problems. One of the planned tests, LMS-DG-01, had not been reviewed by engineering or operations personnal until just before the  j
  , scheduled run time. As people were gathering in the control room for the  i
  " heightened level of awareness" (HLA) briefing for the control room operators and involved engineering and maintenance personnel to discuss the testing activities, control room operators determined that LMS-DG-01 would need to be revised to address minor changes which had been initiated by the system engineer in the fall of 1996. Other planned tests, including LOS-DG-M2 and LOP-DG-02, had been previously updated to address the changes initiated by the system engineer, but LMS-DG-01 had not been revised and could not be performed as writte l j
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Operators encountered a delay in performing another planned test of the 1 A EDG per LOS-DG-M2 after completing the pre-start checklist for this procedure. A non-licensed operator identified that the switch nomenclature in LaSalle Special Test (LST) 96 283, " Diesel Generator 1 A Generator / Engine Control Switch (1HS-DG013)
Replacement Logic and Functional Test," Revision 0, did not accurately reflect the labeling on installed switches. The system engineer had used the nomenclature from plant electrical schematic diagrams rather than the nomenclature that existed in the plant during the development of the LST. Many of the labels in the plant had been changed due to a recent label upgrade program which had been completed for the 1 A EDG. Operations personnel performing portions of the test in the EDG room correctly identified the discrepancy and initiated a. procedure change. The label discrepancies increased the chance for error and placed an additional burden on the operators to screen the procedures for adequac In addition, the inspectors identified that the operator performing surveillance testing of the 2A EDG on March 3,1997, incorrectly recorded the time that the EDG reached its rated load of 2400kw rather than the time that the output breaker was closed. Step E.13.12 of LOP-DG-02 requires the operator to make appropriate entries in Attachment E, " Diesel Generator Start and Run Log," of LOP-DG 02.
 
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Section 3 of Attachment E to LOP-DG-02 requires the operator to log the time that the output breaker is closed, not the time that the EDG reaches rated load. The operator recorded the time that the EDG reached its rated load of 2400kw to ensure that Technical Specification requirements for testing the EDG at rated load were met. The operator's failure to record the time that the output breaker was closed during the 2A EDG testing as required by procedure is a violation of Technical Specification 6.2.A.a, as described in the attached Notice of Violation (50-373/97003-01a: 50 374/97003-01a). On a different occasion, the inspectors observed a different operator using LOP-DG-02 to test the 1 A EDG. The operator recorded the time that the output breaker was closed as required by the procedure, which indicated that the inappropriate logging practice exhibited by the one operator was not a common practice for all operator The inspectors also identified one instance where the O EDG was not operated at rated load for greater than or equal to 60 minutes as required by LOS-DG-M1. The -
stated purpose of LOS-DG-M1 is to demonstrate that the O EDG can be started add operated at rated load for at least 60 minutes, although there are no procedural steps in LOS-DG-M1 to record the run time at rated load. Step 3.12 of Attachment A to LOS-DG-M1 requires the operator to record engine data on Attachment C2 of LOS-DG-M1 after the EDG has been loaded for at least (1) one hour .QB reached thermal equilibrium (which ever time is greater). The inspectors reviewed the results of the surveillance test performed per LOS-DG-M1 on February 27,1997, and determined that the EDG was not operated at rated load for greater than or equal to 60 minutes as required by the procedure. The EDG was operated at rated load for 56 minutes, which invalidated the surveillance test. The failure to operate the O EDG for at least one hour at rated load as required by test procedure LOS-DG-M1 is considered a violation of Technical Specification 6.2.A.a, as described in the attached Notice of Violation (50-373/97003-01b; 50-374/97003-01b). The licensee reviewed test data for all EDG surveillance tests
 
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conducted since January 1996 and did not identify any other test where the EDG .
  .was not run at rated load for the required time period,  j Conclusions f
Several problems were encountered by operators when attempting to conduct EDG l testing. The operators were unnecessarily challenged with procedures which other !
work group: had not recently reviewed for deficiencias prior to the scheduled ,
activities. While the operators normally followed procedures, in two separate ;
instances identified by the inspectors, the operators did not follow procedural requirements and performed steps incorrectly. This resulted in the invalidation of a
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l completed surveillance test and an additional run of an ED i
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02.2 - Inadeauste Accentance Criteria Results in an Unusual Event for Hiah Coolina Lake l 5      ,
        , inspection Scone (71707)
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The inspectors observed the licensee's actions following the Unusual Event that resulted from a cooling lake level which was above the level analyzed for flooding in !
the Updated Final Safety Analysis Report (UFSAR). The inspectors reviewed lake level trends, shiftly surveillance data, and discussed the event with plant personne i Documents reviewed included:    ,
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  * LOP-WL-04, " Lake Level and Blowdown Flow Control," Revision 11  ;
LOS-AA-S1, "Shiftly Logs," Revision 56
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  * Observations and Findinas
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On March 1,1997, a maintenance worker initiated a Problem identification Form (PlF) due to his concerns with cooling lake level being nearly 2 feet greater than the ,
top of the normal level band of 699.5 to 700 feet. On March 11,1997, a second l PIF initiated by engineering personnel identified that the LaSalle County Station Final Safety Analysis Report (FSAR), Amendment 24, question 010.10, states that the lake level will not exceed 701 feet except during a once in one-thousand-year flood. The PlF also stated that flood protection was provided for elevations up to 701 feet to prevent internal plant flooding from the cooling lake should a break occur in water system piping located in the plant, as specified in UFSAR Section 3.11.1.4.2 (interior floods). An Unusual Event was declared by operations management on March 11 due to the lake level being at 701.8 feet. The high lake level placed the units in an unanalyzed conditio Following declaration of the Unusual Event, several compensatory actions were taken by the licensee. The technical support center was staffed by licensee management, operations, and engineering personnel to assist the operations department during the event, the lake discharge valve was repaired and opened to initiate approximately 38,000 gpm blowdown flow, and a root cause investigation was initiate ~, ,, -  -
 
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Several problems which preceded the Unusual Event were identified by the inspectors and the licensee including:
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Incorrect acceptance criteria for maximum lake level existed in LOS-AA-S Step F.2 of LOS-AA-S1, required that Attachment C and C-1 to LOS-AA-S1 be completed when the plant was in Operational Condition 4, which was applicable at the time of the event. Attachment C-1, "Shiftly Control Room Back Panel Check For Operational Condition 4," Step 2.c, inappropriately used 701.7 feet for the high lake level acceptance criteria. Also, LOP-WL-04 did not specify a maximum high level, but indicated that the lake should be maintained between 699.5 and 700 feet and would overflow the spillway at  !
702.5 feet. Although the lake levellimit of 701 feet was discussed in the  i UFSAR and in operational evaluations for flooding completed as recently as l 1995, the 701 foot limit was not incorporated into either of these operations l
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procedures. The inspectors considered the failure to incorporate appropriate acceptance criteria into the shift log procedure LOS-AA-S1 an example of a j violation of 10 CFR Part 50, Appendix B, Criterion V, as described in the  ;
attached Notice of Violation (50-373/97003-02a: 50-374/97003-02a).
 
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Operations personnel had become complacent with operating outside of the
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specified operating band in LOP-WL-04 and did not generate a PlF when the  i level could not be maintained within the band. The inspectors reviewed  l records of the lake level that existed between January 1,1996, and March 1,1997, and determined that operators maintained the lake level above the normal operating level band limit of 700 feet approximately 75 percent of the time. The operators had not requested a change to the operating band specified in the procedure or that the cause of the problem be fixed when level could not be maintained in the normal operating ban * Operations personnel accepted the poor material condition of the cooling lake blowdown line discharge valve and the lake makeup system. Operators were reluctant to secure the cooling lake makeup pumps due to concerns with the structuralintegrity of the makeup piping and the knowledge of the amount of work required to restart the pumps. In addition, the blowdown valve frequently required repairs and no one demanded a permanent fix for the valve. Due to hesitancy by operators to secure the running makeup pumps, the lake level continued to increase for 14 days with no expediency being placed on repair of the discharge valve. The valve was repaired and opened within two days following the declaration of the Unusual Even * Trending of the lake level was not performed by the operations or -
engineering department personnel. Trending would have indicated that continuing to run two makeup pumps with the lake discharge valve closed would result in a rapid filling of the lake to the upper limit (701.7 feet)
specified in the shift logs procedure. Trending would have provided additional information and may have prompted a decision to take earlier action to prevent the lake level from rising furthe _ _ _ _ _ _ _ _ _ _ _ .
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- Conclusions The compensatory measures and immediate corrective actions taken after the licensee identified that the station was in an unanalyzed condition were goo However, several opportunities existed to rectify the situation prior to the event, indicating that the operations department personnel were not identifying problems
. and demanding their resolutio .3 Missed Technical Snecifiestion Surveillance due to Failure to Follow Procedures Insoection Scone (71707)
The inspectors reviewed an instance in which required testing of the 1 A residual heat removal (RHR) pump shutdown cooling suction valve,1E12-FOO6A, was not performed within the required time interval. The inspectors discussed the issue with operations and engineering personnel and reviewed the results of the -
licensee's investigation and the following procedures:
* LOS-RH-Q3, "RHR (LPCI) and.RHR Service Water Valve inservice Test for Cold Shutdown or Refueling Condition," Revision 26
* LAP 100-11, "LaSalle County Station General Surveillance Program,"
Revision 14 r LAP-100-29, " Conduct and Review of Station Surve'illances," Revision 6 b. Qbservations and Findinas On February 28,1997, the licensee identified that required testing of the 1 A RHR pump shutdown cooling suction valve,1E12-FOO6A, had not been performed within the required frequency. Operators satisfactorily tested the valve on October 6, 1996, and another test was scheduled for December 23,1996. However, plant conditions prevented operators from testing the valve during a partial completion of LOS-RH-Q3 on December 23. In addition, operators did not test the valve during a second partial performance of LOS-RH-03 on December 29 because the procedure step for testing the valve was incorrectly rnarked as "not applicable" (N/A) by the
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work control center (WCC) SRO. The WCC SRO also incorrectly marked the test cover-sheet as a "whole test" completion. The Unit Supervisor reviewing the test did not correct the cover sheet to indicate the test was only partially complet The two partially complete surveillance tests performed on December 23 and 29, 1996, were combined and inappropriately annotated by the operations surveillance coordinator to indicate that both tests constituted satisfactory completion of the entire test. He did not ensure the untested valve was recorded in the degraded equipment log (DEL) as required by LAP-100-29, Section F.9 and LAP-100-11, Section 3.a. Recording the valve in the DEL would provide a mechanism to monitor the status of the valve and ensure that the valve would be appropriately tested at a later dat On February 28, the inservice inspection (ISI) trend analyst identified that the 1E12-FOO6A valve had been excluded from the tests and that the maximum allowed
 
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extension of the specified quarterly test interval had been exceeded. The maximum allowed extension is defined in Technical Specification 4.0.2 as 25 percent of the specified surveillance interval. When the ISI trend coordinator determined that the valve was not tested as required, he informed the Shift Manager who took appropriate actions. The failure to test the 1 A RHR pump shutdown cooling suction valve,1E12-FOO6A, within the required surveillance interval is considered a violation of Technical Specification 4.0.5, as described in the attached Notice of Violation (50-373/97003-03). Valve 1E12-FOO6A was stoke timed satisfactorily on March 29,1997, during a realignment of the RHR syste c. Conclusions Operations personnel lacked attention to detail during test preparation and did not I fully review the partial surveillance tests to ensure that valve 1E12-FOO6A had been '
tested. The errors made by the shift manager and the operations surveillance coordinator reflect human performance deficiencies that the licensee has been attempting to correc Operator Knowledge and Performance l
03.1 Operator Racked Out Breaker Without the Operatina Procedure at the Work !
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Location Insoection Scone (71707)
The inspectors performed a plant tour on February 11,1997, and observed an operator removing the waste treatment facility electrical supply breaker from servic Observations and Findinas During the plant tour, the inspectors identified that an operator removed the waste treatment plant feed breaker No. 2 from service without having the procedure with him while performing the work. The operator was removing the breaker from '
service in accordance with out-of service checklist number 970001407 for a j planned maintenance activity at the waste treatment facility. The governing '
procedure, LOP-AP-10, " Racking out a 6900 Volt or 4160 Volt Manually Operated ;
Air Circuit Breaker to Test or Disconnect Position," Revision 7, was a " Reference !
Use" procedure required to be at the work location by LAP-100-40, " Procedure Use J and Adherence Expectations," Revision 8, Section B.3.2. When questioned by the l inspectors, the operator stated he was not required to have the procedure available at the job site and knew how to perform the tas The inspectors interviewed other licensed and non-licensed operators regarding expectations for reference use procedures. The operators revealed that they did not feel they needed to have reference use procedures at the job site. The I operators stated that reference use procedures were available if needed and were
 
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not intended to be at the job site. As identified in NRC Inspection Report 96018, the licensee management's expectation is that reference use procedures be available for operators to use if needed when performing work. The failure of the operator to have the procedure available when removing the breaker from service is an additional example of a previous violation (50-373/96018-02; 50-374/96018- 1 02) issued in NRC Inspection Report 9601 )
l Conclusions The operators' definition of reference use procedures and their applicability to work l performed in the field demonstrates that some operators lacked an understanding of i licensee management's procedure use expectations. This issue was an example of l a previous violation and although the licensce's corrective actions for procedure I adherence and procedure quality have genera!!y been effective, this example indicates that additional management attention is needed in the area of " reference use" procedure :
l ll. Maintenance  l l
M1 Conduct of Maintenance    l M1.1 Eauioment Problems identified Durina EDG Post-Maintenance Testina Insoection Scoos (62707)
The inspectors observed post-maintenance and surveillance testing for the 1 A ED Procedures reviewed as part of the inspection included:
* LOP-DG-02, " Diesel Generator Startup and Operation," Revision 22
* LOS-DG-M2, "1 A(2A) Diesel Generator Operability Test," Revision 34
* LMS-DG-01, " Main Emergency Diesel Unit Surveillances," Revision 16 Observations and Findinas On February 26,1997, the inspectors observed post-maintenance testing of the 1 A EDG. Also, a member of plant management and a member of the licensee's Independent Safety Engineering Group were observing the EDG testing. The  ,
inspectors and the licensee identified two equipment deficiencies during the diesel 1 testing. The first deficiency involved a leak on the flange gasket for the service ,
water portion of the jacket water cooling heat exchanger. Maintenance personnel j repaired the heat exchanger while the EDG was out-of-service and the jacket water :
cooler flange was reinstalled without adequate preparation of the sealing surfac l The lack of an adequate surface preparation resulted in rework to repair the lea I The second deficiency consisted of a failure of the EDG room ventilation exhaust damper which failed closed while the EDG was operating. The EDG room ventilation exhaust damper was designed to fail closed and failed to its required 14    l i
 
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position. On February 28,1997, instrument maintenance technicians determined that the ventilation damper failure was due to a short in the wiring to the controlle The technicians also identified that a 1/16 amp fuse was inappropriately installed in l
; a controller for the ventilation system where a 1/10 amp fuse was required. The fuse with the lower amp rating did not fail in service, however, it did indicate that the controls for ensuring the ventilation circuitry remains properly configured was in question. As of March 20,1997, the licensee's investigation had not been
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completed and the size of the fuses installed in the remaining EDG ventilation system's control circuitry were being evaluated. The inspectors will evaluate the j impact of the incorrect fuse size on the EDG ventilation system operability and the-
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status of fuses in the other EDG ventilation control systems.' This is considered an Unresolved item (50-373/97003-05; 50-374/97003-05) pending further review by  !
;  the inspectors, Conclusions
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In general the test was well controlled by the test director and good communication was used within the diesel generator room. Oversight by management and site  ;
quality verification personnel was adequate during the diesel testing. However,
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poor maintenance practices resulted in rework on the jacket water cooling heat i  exchanger. The incorrect fuse in the EDG room ventilation system indicates that a j  potential configuration control problem exits. The inspectors will followup on the licensee's efforts to evaluate this issu ,
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M1.2 Stoo Work Order for SBM (switchboard, miniature) Switch Reofacement Project  j i Insoection Scone (62707)
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The inspectors reviewed the licensee's activities following licensee management's !
decision to stop work on the General Electric (GE) SBM switch replacement projec The inspection included a review of the licensee's plan to correct the problems which precipitated the stop work order and discussions with management and engineering personne >
. Qbservations and Findinas
. On February 8,1997, the Unit 2 Plant Manager suspended the SBM switch replacement project because of deficiencies identified in the procurement of SBM switches, the procedures used to install the switches, and the switch testing procedures. The licensee was replacing approximately 1157 switches located throughout the plant to correct potentially degraded switches and switches that
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have reached the end of their service life. The licensee performed investigations to determine what factors contributed to the problems and to identify improvements >
needed in the procurement, maintenance, and testing of the switches.
 
l After the licensee identified the root cause for the problems, a plan to address the identified problems was initiated. The plan included guidelines for developing
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maintenance and test procedures, performing maintenance, and conducting testing
 
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of the new SBM switches. The licensea enhanced post-maintenance testing by including additional test guidelines, revised maintenance procedures by adding clarifying procedural steps, and revised the switch procurement procedures by adding additional receipt inspection guidance. The corrective actions implemented by the licensee to address deficiencies in the testing procedure were generally effectiv Conclusions The inspectors concluded that the license's decision to stop the SBM switch replacement project to correct the identified deficiencies in the switch procurement process, the maintenance process, and the testing procedures was good. However,
' the deficiencies leading to the stop work order indicated that problems continued with processes which support , maintenance, including poor procedures and workers not meeting or understanding performance objective M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Poor Housekesoina identified in Hioh Radiation Areas Insoection Scone (71707)
Over the course of the inspection period, the inspectors toured several high radiation and high contamination areas with the support of radiation protection technicians. . Areas toured included the upper and lower radioactive waste tunnels, Unit 1 and 2 reactor water cleanup heat exchanger rooms and filter /demineralizer valve rooms, and the high level radioactive waste storage area, Observations and Findinas During the tours, the inspectors identified problems which primarily involved poor housekeeping following maintenance, items left in selected work areas included  ,
tools, unused or abandoned parts, insulation, catch basins, tubing, and drain hose .
j Equipment problems identified by the inspectors included leaking valves and a  '
damaged pipe hanger. The inspectors also identified some personnel safety hazards  j such as the lack of a warning sign or barrier in the upper radioactive waste tunnel  l where the tunnel ends with an elevation change of approximately 75 feet down to  !
the lower pipe tunnel. The inspectors informed the licensee of their observation .!
I Conclusions      j While housekeeping in areas normally toured by plant management was adequate,  j
. the inspectors identified that several high radiation and high contamination areas  -
were in poor condition. This was often due to maintenance personnel leaving tools, parts, and discarded materialin the work area following completion of wor :
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1 - M3 Maintenance Procedures and Documentation M3.1- Inadeounte Accentance Criteria for Steam Tunnel Check Damner Testina
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' Insoection Scone (62707)
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The inspectors reviewed the test methodology and acceptance criteria for steam
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tunnel check damper inspections performed by the licensee using Work Request (WR) 960040017, " Inspect 2VRO1YA/B/C, 02YA/B/C, 04YA/B/C, 05YA/B, 08Y thru 14Y DMPR," and discussed the work request and valve testing with engineering personne b.' ~ Observations and Findinos t
The inspectcra identified that the licensee was performing inspections of the steam tunnel check dampers, VR08Y, VR09Y, VR10Y, VR11Y, VR12Y, VR13Y, and VR14Y for both Unit 1 and Unit 2, without having established appropriate'  ;
acceptance criteria for determining that the valves were operating properly. A system engineer was conducting the inspections during refueling outages, but he
  - did not evaluate the condition of the valves against any qualitative or quantitative acceptance criteri t The steam tunnel check dampers are designed to isolate the main steam tunnel from the reactor building in the event of a high energy line break (HELB) in the steam tunnel. The dampers are designed to close as the steam tunnel pressure increases, since the dampers do not have any type of actuator to reposition the valves. The failure of the check dampers during a HELB would result in steam migrating to portions of the plant for which no analysis has been performed to ensure safety equipment would operate in a steam environment. The steam could potentially affect safety system operabilit The system engineer used his judgement when assessing the condition of the dampers and recorded the as-found conditions during the routine check damper inspections. Comments regarding the condition of the valves during the latest inspection, documented by the system engineer in the WR on November 8,1996, indicated that the valves moved freely and could close. However, the WR did not contain specific criteria for evaluating the capability of the valves to clos Criterion V, " Instructions, Procedures, and Drawings," to Appendix B of 10 CFR Part 50, requires that activities affecting quality be prescribed by documented procedures or instructions and that, these procedures or instructions include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.- The inspectors identified that on November 8,1996, WR No. 96004001, " Inspect 2VRO1YA/B/C, 02YA/B/C,04YA/B/C,05YA/B,08Y thru 14Y DMPR," used by the system engineer to inspect the steam tunnel check dampers, VR08Y, VR09Y, VR10Y, VR11Y, VR12Y, VR13Y, and VR14Y for both units, did not contain acceptance criteria for determining that the dampers would operate satisfactorily. The absence of appropriate quantitative or qualitative acceptance criteria in the subject WR is
 
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considered an example of a violation of 10 CFR Part 50, Appendix B, Criterion I However, this violation is not being cited because it satisfies the criteria in Section Vll.B.2 of the NRC's enforcement policy (NUREG-1600). Conclusions
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Although the licensee was periodically inspecting the steam tunnel check dampers, the work instructions for performing these inspections did not contain acceptance criteria for determining that the dampers would operate satisfactorily. Engineering -)
personnel did not recognize that qualitative or quantitative acceptance criteria need to be considered for all activities affecting qualit M3.2 Missed Technical Soecification Surveillances due to Proaram Weaknesses and  ,
Failure to Follow Procedure Inspection Scone (61726)
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The inspectors reviewed the circumstances surrounding the licensee's failure to !
conduct required testing of the RHR pump 1 A discharge high/ low pressure switch for Unit I low pressure coolant injection (LPCI) train "A" within the required surveillance interval. Documentation reviewed included the licensee's investigation results and the following procedures:
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LAP-100-11, "LaSalle County Station General Surveillance Program,"
Revision 14
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LAP-100-29, " Conduct and Review of Station Surveillances," Revision 6
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LAP-300-6, "LaSalle County Station Instrument Surveillance Program,"
Revision 6
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LaSalle Instrument Surveillance (LIS) RH 316A, " Unit 1 RHR Pump 1 A Dischsige High/ Low Pressure Functional Test," Revision 3
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; b. Observations and Findinas On February 25,1997, while reviewing scheduled surveillance procedures, the ;
WCC SRO identified that surveillance procedure LIS-RH-316A had not been completed within the 31-day periodicity required by Technical Specification ,
Surveillance Requirement 4.5.1'.a.2.a plus the 25 percent maximum allowable ;
extension specified by Technical Specification Surveillance Requirement 4.0.2. The surveillance test had last been conducted on December 28,1996. Subsequently, l the Shift Manager was informed by the WCC SRO that the allowable time extension i of the 31-day surveillance interval had expired on February 4,1997. Operators i satisfactorily completed the test on February 2 Technical _ Specifications require that the RHR pump 1 A, discharge pressure functional test, performed per LIS RH-316A, be completed when the unit is in Operational Condition 1,2,3,4, or 5. In addition, the cover sheet for LIS-RH-316A states that the test be conducted when the unit is in Operational Condition 1,2,3, 4, or 5. However, the mode applicability identified in the General Surveil.ance
 
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Instrument (GSIN) and Electronic Work Control System (EWCS) programs was .
incorrect. The GSIN and the EWCS programs are used to schedule surveillance
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testing and neither program listed the subject surveillance test as being applicable in j
  - Operationa' Condition 4. The incorrect mode applicability on the scheduling sheets
)  originating from the GSIN and EWCS programs, contributed to the incorrect  i i  decision'that the surveillance test was not require j
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;  Administrative procedure, LAP-300-6, requires that an Instrument Maintenance Degraded Equipment Log (IM DEL) be maintained for equipment surveillance tests I
which have exceeded the required surveillance interval which includes the
!  25 percent maximum allowable extension, in order to ensure that the surveillance
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tests are performed before declaring the associated equipment operable. The IM
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DEL is intended to identify any inoperable equipment ti.at requires a surveillance
!. test before changing the plant's operational condition. In addition, LAP-300-6-
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requires the IM DEL to be reviewed and approved by licensee management, and i  Section E.1 of LAP-100-11 requires that the Shift Engineer and responsible
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department supervisor be immediately notified when any Technical Specification surveillance test has not or will not be performed as required. However, in the case -
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;-  of the RHR 1 A discharge pressure functional test, the IM DEL was not completed
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by the scheduler until February 8, four days after the allowed surveillance test interval had expired, and the shift engineer or department supervisor were not notified by the scheduler that a Technical Specification required surveillance test had not been conducted. Consequently, the management reviews of the IM DEL which were required by the licensee's administrative procedures, were not completed. The failure to complete the channel functional test within the required surveillance test interval is considered a violation of Technical Specification 4.5.1.a.2.a, as described in the attached Notice of Violation (50-373/97003-04).
 
The subject test was satisfactorily conducted on February 25,199 c. Conclusions Actions taken by plant personnel following discovery of the missed surveillance test were appropriate. The failure to follow procedures and incorrect mode applicability information in the surveillance testing scheduling programs caused the licensee to exceed a required surveillance test interval. This is another example of the
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licensee's failure to cornplete required Technical Specification testin l M8 Miscellaneous Maintenance issues    I i
M8.1 (Closed) LER 50-373 97007: Missed Technical Specification Surveillance on the  3 High and Low Discharge Pressure Switches for the 1 A RHR Pump Due to  i Procedural and Human Performance Errors. This problem was discussed in Section M3.2 and a Notice of Violation was issued. This item is considered close j
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E1 Conduct of Engineering    +
E1.1 Massed Technical Soecification Surveillance Due to Personnel Error Insocction Scoos (37551. 92903)     i
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i On January 9,1997, the licensee reported to the NRC in Licensee Event Report  f (LER) 96019 that the Unit 1 RHR system containment spray isolation valves had  :
not been tested according to American Society of Mechanical Engineers (ASME)  )
Section XI requirements due to personnel error. The inspectors reviewed LER 96019, " Residual Heat Removal System Containment Spray isolation Valves Not Tested According to ASME Section XI Requirements Due to Personal Error,"
Technical Specification requirements for the associated valve, the requirements of the Section XI portion of the ASME Code, and applicable operating procedure Observations and Findinas During a review of the LaSalle Station Inservice Test (IST) Program as part of the corrective action for LER 374/96-006-00, " Unit 2A and 2B RHR Service Water Pumps not tested per ASME Section XI," the licensee identified on December 9, 1996, that required testing of the Unit 1 and Unit 2 motor-operated containment spray isolation valves 1(2) E12-F016A and 1(2) E12-F017A were not included in LOS-RH-02, "RHR (LPCI) and RHR Service Water Valve Inservice Test for Operating, Startup and Hot Shutdown Conditions." The inspectors reviewed completed tests for valves with similar functions and determined that Unit 1 valves, 1E12-F016B and 1E12-F0178, and Unit 2 valves,2E12-F016A/B and 2E12-F017A/B, were tested appropriatel Technical Specification 4.6.3.3 requires that "the isolation time of each primary containment power operated or automatic isolation valve be determined to be within its limit when tested pursuant to Technical Specification 4.0.5." Technical Specification Surveillance Requirement 4.0.5 requires implementation of the  .i applicable ASME Section XI code for inservice testing. The applicable code for the ;
current 10-year IST interval requires that selected valves be exercised to the  j
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position required to fulfill their function and that the corresponding valve stroke time be measured to the nearest second. The LaSalle County Station Pump and Valve inservice Testing Program requires that Unit 1 motor-operated containment spray isolation valves 1E12-F016A and 1E12-F017A be tested in the closed ~
direction on a quarterly frequenc !
A system engineer inappropriately revised LOS-RH-02 to test valves 1E12-F016A and 1E12-F017A in the open direction following a modification to the valves. The valves had been tested satisfactorily in the closed direction on March 5,1996, however, subsequent tests stroke timed the valves in the wrong direction due to the procedure revision. The valves were tested correctly on October 6,1996. 'The failure to stroke test valves 1E12-F016A and 1E12-F017A within the required
 
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. surveillance test interval is considered an example of a violation of Technical Specification Surveillance Requirement 4.6.3.3. However, this violation is not being cited because it satisfies the criteria in Section Vll.B.2 of the NRC's enforcement policy (NUREG-1600). Conclusions      ;
I The system engineer did not verify that LOS-RH-02 contained the correct stroke l time requirements when revising the procedure. Consequently, the subject valves were tested in the wrong direction. The IST program was being reviewed by the l lST Engineer to verify program requirements were being satisfied. The actions I
taken by the licensee following identification of the failure to test the valves in the I required direction were appropriat I E2 Engineering Support of Facilities and Equipment  ;
E2.1 Uncualified Contractor Performed Installation and Calibration of Lake Blowdown Flow Instrumentation Insoection Scoce (62703)
The licensee identified that an unqualified vendor installed and calibrated the lake blowdown line flow instrumentation. The inspectors interviewed cognizant personnel and reviewed the licensee's investigation results documented in LAP 220-5, Attachment B, " Concern Screening Form," dated February 26,199 Observations and Findinns On February 12,1997, the licensee identified that an unqua!ified contractor calibrated the lake blowdown line instrumentation, OFE-WL-001, in September 1995 and December 1996. In addition, the original flow I instrumentation was replaced by an unqualified vendor. The blowdown line  '
instrumentation provides indication of the flow rate of water from the cooling lake to the river. The instrumentation is also used to monitor discharge flow rates to l ensure that radioactive waste discharges do not exceed 10 CFR Part 20 limit On August 31,1994, the licensee completed a review of the installation of new i flow instrumentation installed that was installed per LST-93-061, " Lake Blowdown j Ultrasonic Flowmeter Special Test Procedure," Revision 1, and controlled with I temporary alteration 1-1018-94. The licensee installed the new flow  l instrumentation because the original instrumentation system was degraded. The !
engineering evaluation for this temporary alteration incorrectly concluded that no controls were required for procuring the equipment or for its calibration. The original lake blowdown flow instrumentation was removed from service and replaced with different instrumentation without a design review. The installation of l
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the new flow instrumentation constituted a design change, however, it was not reviewed or approved by the design organization. Criterion 111, " Design Control," of
 
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Appendix B to 10 CFR Part 50 requires that measures be established to ensure that I l
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applicable regulatory requirements and the design basis are correctly translated into !
specifications, drawings, procedures, and instructions. Criterion ill further requires
:  that design changes be subject to design controls commensurate to those applied l  to the original design and that the changes be approved by the responsible design ,
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!  organization. On August 31,1994, the lake blowdown flow instrumentation was
!  removed from service and replaced with different instrumentation. This design ,
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change was not subject to design controls commensurate to those ap,nlied to the !
j  original design. Furthermore, the design change was not approved by the
'        l responsible design organization because it was performed during a testing activit !
l  This is considered a violation of 10 CFR Part 50, Appendix B, Criterion Ill.
 
I However, this violation is not being cited because it satisfies the criteria in l  Section Vll.B.2 of the NRC's enforcement policy (NUREG-1600).
 
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On February 12,1997, the licensee identified that on two different occasions, the blowdown flow instrumentation was not calibrated with approved plant procedures by contractors on the licensee's list of contractors approved for work on safety-related equipment. The instrumentation was listed in the licensee's Offsite Dose Calculation Manual (ODCM) and used to perform the radioactive waste discharge calculations. Because the instrumentation was listed in the ODCM and used to perform discharge calculations, the licensee's quality assurance program required ;
that approved contractors or vendors be used to procure and calibrate the I equipment. The licensee's failure to calibrate the blowdown flow instrumentation using approved procedures is considered a violation of 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," as described in the attached NOV (50-373/97003-02b; 50-374/97003-02b).
 
Upon identifying this issue, engineering personnel performed an operability evaluation of the blowdown flow instrumentation. The evaluation addressed the impact of the potentially improperly calibrated flow instrumentation on the radioactive waste discharges which occurred after the new equipment had been installed.~ The licensee concluded that the flow instrumentation was operable by verifying the original calibration using another qualified, approved vendor. In addition, the licensee's chemistry department reviewed calculations for 18 radioactive waste discharges ' performed between December 1996 and  I January 1997. The licensee's calculations verified that the discharges did not exceed allowable release limit Conclusions No radioactive waste discharges which exceeded 10 CFR Part 20 limits occurred while the unqualified flow instrumentation was installed. However, licensee ]
personnel did not recognize that the replacement and calibration of the flow !
instrumentation should have been performed using appropriate processes and !
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E2.2 Inadeouste Desian Chanae introduces Sinale Failure Vulnerability and Unreviewed Safety Question    i
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a. . Insoection Scone (62703)
i On January 13,1997, the licensee determined that the main control room  j ventilation system radiation monitors were susceptible to a single failure. The I licensee modified the radiation monitoring system in 1993 during which the I initiation logic was changed. The change in the wiring configuration for the radiation monitors' initiation logic introduced both a single failure vulnerability and an unreviewed safety question. The inspectors reviewed the radiation monitor logic drawings, the UFSAR, previous FSAR revisions, and the design change documentation used to conduct the modificatio Observations and Findmgs During a review of Technical Specification interpretations for the control room (CR)
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and auxiliary electric equipment room 1AEER) radiation monitoring system, engineering personnel discovered that the radiation monitors were susceptible to a single failure. A postulated single failure in the radiation monitoring circuitry could have caused the emergency ventilation system not to actuate. This failure, concurrent with a design basis accident, could have resulted in a radiation exposure to control room personnel greater than the limits specified in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 1 The CR and AEER ventilation system wrs designed to supply filtered air to the j respective rooms through a shared emergency ventilation system actuated upon )
detection of a high radiation condition at the system air intake. The CR and AEER ventilation system consists of two 100 percent capacity ventilation trains with an air intake for each train. Four radiation monitors are located at each air intake to monitor for a high radiation condition. The radiation monitoring system was originally designed to initiate the emergency ventilation system if one of the four ,
monitors detected a high radiation conditio The licensee had experienced spurious actuations of the radiation monitors and therefore decided to modify the system in July 1993. The modification was performed per modification M01-0-88-003 A, "MCR [ main control room] HVAC
[ heating, ventilation, and air conditioning] Intake Radiation Monitors," which was approved on May 14,1993. The radiation monitoring system was modified to require two monitors to initiate the emergency ventilation system. This  ;
modification introduced the potential for a single failure in the logic circuitry which j could have prevented starting the emergency ventilation system following a design basis acciden In addition to the single failure vulnerability, the licensee determined that the modification increased the probability of a failure of equipment important to safet The increased failure probability was due to the increase in the number of radiation
 
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monitor to tw i
! In describing the configuration of the radiation monitors, Section 6.4.4 of the j . UFSAR states that there are four monitors divided into two channels, with j actuation of two-out-of-four monitors required to start the emergency ventilation
; system. The UFSAR also states that the emergency ventilation system is designed to limit the occupational dose to less than the limits specified in GDC 19 of
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Appendix A to 10 CFR Part 50, Appendix A, GDC 19. However, before the
! radiation monitor initiation logic was modified in July 1993, the FSAR, Revision 0, 1- dated April 1984, stated that the four monitors were divided into two channels and
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that any one monitor sensing a high radiation condition would start the emergency
; makeup filter train for each air intake. Technical Specification 3.3.7.1, " Radiation j' Monitoring Instrumentation," requires that two channels of the main control room j'
radiation monitors be operable during Operational Conditions 1,2,3, and 5, and when irradiated fuel is being handled in the secondary containment.
 
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The licensee did not identify during the original design change review that the modification would introduce a single failure vulnerability into the syste .
:  Criterion ill, " Design Control," to Appendix B of 10 CFR Part 50 requires the
!  licensee to ensure that the design basis is correctly translated into specifications, i
drawings, procedures, and instructions. The failure to correctly translate the design j  basis for the CR and AEER ventilation system radiation monitors into procedures
:  and instructions associated with a modification to the system in 1993 is considered
;  an apparent violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design Control" j  (50-373/97003-06; 50-374/97003-06).
 
i j' in addition, the licensee did not identify that the 1993 modification would require i  NRC approval because the design change involved an unreviewed safety question.
 
i  The licensee's safety evaluation for the modification concluded that no unreviewed j  safety question existed and that the ability to meet single failure criteria described j  in the FSAR was not affected. The failure to identify that the design change to the j  radiation moniters involved an unreviewed safety question is considered an  ,
;  apparent violation of 10 CFR Part 50.59, " Changes, Tests and Experiments" !
  (50-373/97003-07; 50-374/97003-07).
 
; Conclusions l
1        i
-
The safety evaluation completed by the licensee for the 1993 modification of the I
:  CR and AEER ventilation system radiation monitors was inadequate. The safety i
        '
;
evaluation did not identify that the change to the radiation monitor initiation logic circuitry constituted an unreviewed safety question and would also introduce a single failure vulnerability into the syste l
'
~E8 Miscellaneous Engineering issues .
E (Ocen)LER 50-373-96019: Residual heat removal system containment spray isolation valves not tested according to ASME Section XI requirements due to :
 
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personnel error. This issue is discussed ir. Section E1.1 of the report. The licensee i  did not identify in the subject LER that Technical Specification Surveillance  ,
'
Requiremen'. 4.6.3.3 is applicable. The licensee is further evaluating this issue and i i  - therefore s.ER 96019 will remain open.
 
,
IV. Plant Suncort I
4  R4 Staff Knowledge and Performance in Radiological Protection and Chemistry R4.2 ' Radiation Protection Technician Performance f
!
; Insoection Scone (71750)
!
l  The inspectors toured several high radiation and high contamination areas with the
;  aupport of radiation protection technicians. Areas toured included the upper and
!
'
lower radioactive waste tunnels, Unit 1 and 2 reactor water cleanup heat exchanger rooms and filter domineralizer valve rooms, the high level radioactive waste. storage area, and phase separator valve room Observations and Findinas Over the course of the inspection period, the inspectors toured several high radiation and high contamination areas. Radiation protection technicians were assigned to accompany the inspectors as these areas were not normally surveye ~ The technicians performed surveys, took samples for lose contamination, transported materials between rooms, and ensured that the requirements of the radiation work permits were met. The inspectors observations of plant conditions during these tours are described in Section M2.1 of this repor Conclusions The technicians accompanying the inspectors were knowledgeable of the plant and radiation protection practice V. Manaaement Meetinas  ,
 
        !
X1 Exit Meeting Summary    i The inspectors presented the results of their inspection activities to licensee management listed below at an exit meeting on March 20,1997. The licensee acknowledged the findings presente The inspectors asked the licensee if any materials examined during the inspection should be considered proprietary. No proprietary information was identifie I l
l 25  i i
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X3 Management Meeting Summary    :
.        !
On March 17,1997, the licensee and NRC representatives discussed various issues  ;
*
including the licensee's restart plans, the High Intensity Training Program for  :
operators, and plans to address human performance deficiencies in a meeting open  i to public observation. Attached to this report are the slides used by the licensee in .
        '
; its presentation.
 
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4  PARTIA.L LIST OF PERSONS CONTACTED Commonwealth Edison W. Subalusky, Site Vice President
  *F. Dacimo, F! ant General Manager L. Guthrie, Unit 1 F1 ant Manager S. Smith, Unit 2 Plant Manager
  'J. Mcdonald, Site Quality Verification / Safety Assessment Manager
!  'A. Javorik,' System Engineering Supervisor  j l  D. Soone, Health Physics Supervisor i  'P. Barnes, Regulatory Assurance Supervisor
  * Present at exit meeting on March 20,199 I INSPECTION PRDCEDURES USED IP 37551 Onsite Engineering '
iP 61726 Surveillance Observation IP 62707 Maintenance Observation IP 71707 Plant Operations IP 71750 Plant Support Activities IP 92903 Followup-Engineering
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ITEMS OPENED, CLOSED, AND DISCUSSED
,
Open I
50-373/374-97003-01 a VIO Operator failure to follow EDG test procedure 50-373/374-97003-01 b VIO Operator failure to follow EDG test procedure 50-373/374-97003-02a VIO Inadequate acceptance criteria in shift log procedure 50 373/374-97003-02b VIO Lake blowdown flow instrumentation calibrated without procedure 50-373-97003-03 VIO Failure to stroke time test RHR shutdown cooling valve 50-373-97003-04 VIO Failure to perform RHR dincharge pressure alarm instrumentation surveillarace test 50-373!374-97003-05 URI Review of licensee investigation of EDG ventilation fuse
,
sizing 50-373/S? v37003-06 eel Ventilation radiation monitor design basis incorrectly translated during modification review 50-373/374 97003-07 eel Design change implemented witnout required Commission approval
.
Discussed or Closed .
, LER 50-373-97007 Closed Missed Technical Specification surveillance on the high and low discharge pressure switches for the 1 A RHR
'
pump due to procedural and human performance errors
'
l LER 50-373-96019 Open RHR system containment spray inclation valves not j tested according to ASME Sectius Xi requirements due '
to personnel errors
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LIST OF ACRONYMS USED ,
ACMI Administration and Course Management instructions *
AEER Auxiliary Electric Equipment Room  i ASME American Society of Mechanical Engineers
] CR Control Room DEL Degraded Equipment Log  ]
DRP Division of Reactor Projects  ;
EDG Emergency Diesel Generator  '
EWCS Electronic Work Control System FSAR Final Safety Analysis Report 1 GDC General Design Criteria GE General Electric  ;
GPM Gallons Per Minute GSIN General Surveillance Instrument  i HELB High Energy Line Break  I i HLA Heightened Level of Awareness  !
HVAC Heating, Ventilation, and Air Conditioning i IM DEL Instrument Maintenance Degraded Equipment Log IR inspection Report ISI inservice Inspection
, IST Inservice Test  1
    '
>
LAP LaSalle Administrative Procedure LER Licensee Event Report
>
LIS LaSalle Instrument Surveillance LMS LaSalle Maintenance Surveillance
,
LPCI Low Pressure Coolant injection
'
LOP LaSalle Operating Procedure LOS LaSalle Operating Surveillance  ,
LST LaSalle Gpecial Test  i MCR Main Control Room
    '
MCRACS Main Control Room Atmospheric Control System i NRC Nuclear Regulatory Commission NSO Nuclear Station Operator NOV Notice of Violation  :
ODCM Offsite Dose Calculation Manual PIF Problem Identification Form PDR NRC Public Document Room PMT Post-Maintenance Test RHR Residual Heat Removal  l RP Radiation Protection SBM Switchboard, Miniature SOV Site Quality Verification SRO Senior Reactor Operator UFSAR Updated Final Safety Analysis Report
 
*
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*
URI Unresolved item VIO Violation WCC Work Control Center WR Work Request ;
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30
 
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LaSalle County Station Perfonnance Review Meeting i
March 17,1997
:
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j  Mel Leach, NRC  Opening Remarks i
l  Bill Subalusky   introduction
}  LaSalle County Site Vice President i
!  Fred Dacimo  Human Performance and Restart Plan Overview
.
;  Dave Farr  Restart Plan 1.18, l  Unit 2 Operations Manager  Operator High Intensity Training  :
i
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Bill Subalusky  NextMeeting/Critque Mel Leach  Closing Remarks
    -
    .
        - -
February 21,1997 I
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Simplified Restart Plan Flowchart  -
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LCNS Restart Plan
,___________--------~~>  '
j  - Safe Plant Operation '  'NPUTS
      '
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{ Emergent
; issues  -
!
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V Management Review Committee
.______________________
  - Work Scope 4 -----
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      ,l  l Internal Assessments ! -
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  -Une Organization -J l-SQV  l-Performance Indicators  l l  4
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External Assessments  l i
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  -Extemally Led Review-INPO, NRO. Other  V
    *
1  Long Term improvement Plan V
Implement Successful Restart and Resume Full Power Operation i
Page1
 
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  - - _ - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -  ---- . - ---------
:
i l  M        Introduction
,
j        e Recentlyjoined the LaSalle Team i
!        e Member of the ISA Team l        - Overall Corporate Perspective
}        - LaSalle Specific Perspective
!
:
e Endorse Findings of ISA
,
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e Commitment to Change t
        - Proper Safety Culture
;
!
 
e All Hands Meeting
              -
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March 17,1997
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,  M Our Expectations of Management i
!  . We will operate our plants on the basis of
!
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conservative decision making;
;
!  e We will listen to our employees and communicate I  openly and honestly with them; i
i
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e We will improve the materiel condition of our plants;
!
l  . We will continuously practice critical self-l  assessment (we will be our own worst critics); and
!
I  e We will take ownership and accountability for the
{  present and future of this organizatio .
              -
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              ;
March 17,1997
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        -  -
 
!
i t
l    lIl    Number One Priority i
!
!:
l e  We will operate LaSalle in a safe and
!        conservative manner, i.e. we will operate safely j      or not at all;
;
!
i
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e  Never will anyone doubt or be concerned about decisions made at "2:00 a.m. on a Sunday l      morning";
!
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l      e  if we fail, it will be on economics - not on I      performance....
!
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March 17,1997
. . _ - . . . _ , __....______... ____.... _ ,__..__. ,__-..--_~_ ...- .__... _ . _ ~.. _. ., . . . . ~ - _ _ _  , _ . . _ , - - _ _
 
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M Demand Effective Communication l
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e Up / Down / Laterally i
j e Fundamental Job Skill
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March 17,1997 l
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lll Accountability
!
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l e " Insufficient reinforcement of'
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; management expectations, follow-up and i
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use of day - to- day accountability
 
l contribute to short falls in LaSalle's-
:
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performance." - From ISA Team
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March 17,1997
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- --  ------- - - -  - -- - _ -
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M      Expectations
:
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!    e  Hold yourself and your peers accountable
:
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}
j      - I will hold each of you accountable e Individuals at all levels - Executive, Managers,
;      Supervisors, and Workers i
l
)      - Take responsibility for their actions and are
!
I committed to improving their performance
!
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March 17,1997
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l
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i M  We Are the Team
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j e We set the standards, in our communication, and in our actions, that we expect others to l
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follow i
!
e We are the team ofindividuals who will take
]
: this plant forward into world-class status I
!
l l h EEc5
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L    March 17,1997
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- - - - - - - - - - - - - - - - - - - - - - -
    - _
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i M  Self-Assessment l
e Critical Self-Assessment is expected of all; i
;
e When you stop learning from your i  mistakes or stop learning how to improve
;  - you stop adding value; i
i
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i e We need to self-identify on our own l  issues; if INPO or NRC finds these issues j  before we do, we've failed!
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March 17,1997
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      - - - - - - - - - -  --- ---  - .
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M    Self-Assessment i
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e Use the corrective action process to l
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l    ensure we are never cited for the same
!    thing twice.
:                  -
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;    e We will learn from our mistakes and
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l    implement effective corrective actions to          '
l    drive a stake through the heart of the
!    problem.
 
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l  M    Immediate Challenge i
lr    e  Stay Focused on the Fundamentals i
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      - Conservative Decision Making
:
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      - Leadership
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      - Accountability and Ownership
 
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      - Self-Assessment
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      - Materiel Condition i
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l            12 March 17,1997
  + .
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          -
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Our Performance in j    Day-to-Day Operation
;
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l    e We follow our procedures
,
i b
!    e We utilize STAR {Stop, Think, Act, l    Review?
\.
l    e We have a Questioning Attitude
 
i e We demand resolution ofissues
:
:
        -      -
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March 17,1997
  *  .
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. _ _ - . _ - . - _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - . . _ _ _ _ _ _ _ . . . _ - . _ . - . _ _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ _ . . . _ _ _ . _ . . . , , , _ . . . _ _ . _ . , .
 
  -
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j Management Sets the      '
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i Standards in the Field i
i    e Worker performance in the field is a~ direct l    reflection of you and your standards and        .
i    workers take their cue from what they perceive j    management wants
!              -
e They reflectyour:
l    - Attitude towards procedure compliance
    - Philosophy of critical assessment
;    - Philosophy of standards on materiel condition l    - Beliefin ownership and doing thejob right the first time h E li c E
 
              '
Mach 17,1997
.
  . .
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- - - - . _ --
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li Station Direction:
 
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e Safe, Uneventful Start-up
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l  e Safe, Long Uneventful Run
 
      .
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i e World Class Performance i
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March 17,1997
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      .------,n,_--N
 
  -  _ -    - -
!
M Each Individual's        Contribution
\
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Must Be:
    .          .
I 1)  Strict Procedural Adherence l  2)  Strong Use of Self-Checking Program i    '
i, STAR)
!
.
l
 
3)  Questioning Attitude
:
j  4)  Demand Resolution of issues
;  c o u n miics            i
~
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:
M50-Day Performance Measures      .
;
End date: 4/2/97 1. Station Event Free Clock
      *
l      average time between events -increasing
      # of resets - decreasing
 
1l i
2. Human Performance PlFs -Trending Down
:
:  .
significant vs. non-significant I
3. Procedure Adh'e rence PlFs -Trending Down
!
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March 17,1997
  *
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.-_...__---._,.._.,__.._,._._.___._____._.,_____.-___________..____________.__._____,_.__.___,___..._,,,....4.,,---  - _ . , , _ , . .
 
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50-Day Performance Measures i
End date: 4/2/97
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l
!
j Procedure Quality PlFs -Increasing l Correlate Depart Clock Resets
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!
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!  Contribute i
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. Individuals
  .
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'
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l  M    50-Day Performance Measures l      End date: 4/2/97
 
j      e  All individuals can state Station Direction and j        how each can contribute
!                '
:
l        - Survey conducted March 13,1997
;
i
.
Station Direction
]        90% provided correct responses
;
!
Individuals Contribution
;        90% provided correct response
                '
i I
\-
,
              .
March 17,1997
                ,
.+  .
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Windows based format
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This will Mrc;id-;+ receipt of your letters dated June 13, August 27, and September 29,1997, in response to our letters dated May 20, July 28, and August 29,1997
.
- transmitting Notices 'of Violation associated with inspection Report Nos. 50 373/97003(DRP),-
.
  - 50 374/97003(DRP); 50 373/97006(DRP),50 374\97006(DRP); and 50 373/97007(DRP),
T ,     a t i
50 374/97007(DRP). We have reviewed your corrective actions and have no further questions at this time. These corrective actions will be examined during future inspections.
_
U A
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_      _,
Sincerely,
 
G
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.
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a i
I Anton Vogel, Acting Chief Reactor Projects tranch 2 Docket No. 50-373 Docket No. 50-374 See Attached Distribution
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  *
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    - General Topics l
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.
  *
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In-the-field performance significantly declining
    - Sixteen (16) of the 1996 Event Free Clock Resets j    attributed to Operations Activities i
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W. Subalusky  2-cc: O. Kingsley, Nuclear Generation Group President and Chief Nuclear Officer M. Wallace, Senior Vice President, Corporate Services E. Kraft. Yke President BWR Operations Liaison Officer, NOO BOD D. A. Sager Vice President,    ,
      *
Generation Support D. Farrar, Nuclear Regulatory Services Manager 1. Johnson, Licensing Operations Manager Document Control Desk-Licensing F. Decimo, Plant General Manager  .
P. Barnes, Regulatory Assurance Supervisor Richard Hubbard Nathan Schloss, Economist Office of the Attomey General State Liaison Officer Chairman, Illinois Commerce Commission Distribution:
Docket File    DRP OC/LFDCB    TSS PUBLIC IE-01    DRS (2)
A. Beach    Rlli PRP, Deputy RA    RAC1 (E Mail)
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        ,
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l  Crew 5 mid-term written exams successful
            '
Crew average score was 88%
l Fundamentals average was 82%
            '
l    Lowest score was 80%
.
i-
\ . .
. . - _ - - . . - . . . . - . - . . - - , . - - - . _ _ _ - _ . . - - . - . . - . - . - - . . - . _ . . . . - . - - , _ - - - . . . - . - . - . . .- ,.. .-,.- - . - --. .-
 
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  . . , i Results to Date, continued
  -
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i Mid-term simulator examinations were NOT i
i successfully completed for Crew 5.
 
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s
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l  - Log keeping      ,
j  - Adherence to procedures
:
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          *
 
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  - -
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:
!
 
l
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.
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!
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;
;
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      . Logkeeping l Licenses suspended
:
;
.
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-
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,
;  Crew Suspended from licensed-duties c  -
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  , .
i i    ..
'
  . , ,,
  . ,
    , _ . . _ . . . . . , _ .
 
i
;
.
. ..
l
>
; . . > Next Steps
;
i
!
!
Do extensive failures have generic implications?
l i  The Performance Review Committee recommended:
j  - Remaining 5 crews screened on ara accelerated basis j
Accelerated Screening to begin 3/21 and complete by 4/2 j  HIT will continue following the evaluation process
 
l
;
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' _ _ _ . . . _ _ . _ _ _ _ . - _ . . . , . . . _ . - - - - - - - -
}}
}}

Latest revision as of 05:09, 19 November 2020

Ack Receipt of 970613,0827 & 0929 Ltrs Informing NRC of Steps Taken to Correct Violation of Insp Repts 50-373/97-03, 50-374/97-03,50-373/97-06,50-374/97-06,50-373/97-07 & 50-374/97-07 Issued on 970520,0728 & 0829
ML20199K127
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 11/25/1997
From: Anton Vegel
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To: Subalusky W
COMMONWEALTH EDISON CO.
References
50-373-97-03, 50-373-97-06, 50-373-97-07, 50-373-97-3, 50-373-97-6, 50-373-97-7, 50-374-97-03, 50-374-97-06, 50-374-97-07, 50-374-97-3, 50-374-97-6, 50-374-97-7, NUDOCS 9712010021
Download: ML20199K127 (2)
v • d • e


Text

r

,. --

Mr. W. T. Subalusky, Jr.

i Site Vlos President LaSalle County Station Commonwealth Edison Company 2001 North 21st Road Marseilles,IL 61341 SUSJECT: . NOTICES OF VIOLATION (NRC INSPECTION MEPORT NOS, 50-373/97003(DRP), 50 374/97003(DRP); 80 373/97006(DRP),

50 374/97006(DRP); AND S0 373/97007(DRP),60 374/97007(DRP))

- Dear Mr. Subalusky.

This will Mrc;id-;+ receipt of your letters dated June 13, August 27, and September 29,1997, in response to our letters dated May 20, July 28, and August 29,1997

- transmitting Notices 'of Violation associated with inspection Report Nos. 50 373/97003(DRP),-

- 50 374/97003(DRP); 50 373/97006(DRP),50 374\97006(DRP); and 50 373/97007(DRP),

50 374/97007(DRP). We have reviewed your corrective actions and have no further questions at this time. These corrective actions will be examined during future inspections.

Sincerely,

.

I Anton Vogel, Acting Chief Reactor Projects tranch 2 Docket No. 50-373 Docket No. 50-374 See Attached Distribution

,

i o

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-

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

_ _ - . . - .. - _ _ _ _ - = _- - - - --

. .

W. Subalusky 2-cc: O. Kingsley, Nuclear Generation Group President and Chief Nuclear Officer M. Wallace, Senior Vice President, Corporate Services E. Kraft. Yke President BWR Operations Liaison Officer, NOO BOD D. A. Sager Vice President, ,

Generation Support D. Farrar, Nuclear Regulatory Services Manager 1. Johnson, Licensing Operations Manager Document Control Desk-Licensing F. Decimo, Plant General Manager .

P. Barnes, Regulatory Assurance Supervisor Richard Hubbard Nathan Schloss, Economist Office of the Attomey General State Liaison Officer Chairman, Illinois Commerce Commission Distribution:

Docket File DRP OC/LFDCB TSS PUBLIC IE-01 DRS (2)

A. Beach Rlli PRP, Deputy RA RAC1 (E Mail)

'

Rlll Enf. Coord. CAA1 (E-Mail)

l. SRI LaSalle DOCDESK (E-Mail)

Project Mgr., NRR o

i

, , _

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