IR 05000483/1997005: Difference between revisions

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{{Adams
{{Adams
| number = ML20141E479
| number = ML20149G857
| issue date = 06/26/1997
| issue date = 07/16/1997
| title = Insp Rept 50-483/97-05 on 970210-14 & 24-28.Violations Noted.Major Areas Inspected:Engineering
| title = Notification of Licensee Meeting W/Util on 970815 to Discuss Failure to Perform SE IAW 10CFR50.59 Requirements Identified in Insp Rept 50-483/97-05
| author name =  
| author name = Howell A
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| docket = 05000483
| docket = 05000483
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-483-97-05, 50-483-97-5, NUDOCS 9707010071
| document report number = 50-483-97-05, 50-483-97-5, EA-97-168, NUDOCS 9707240099
| package number = ML20141E437
| document type = INTERNAL OR EXTERNAL MEMORANDUM, MEMORANDUMS-CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 4
| page count = 39
}}
}}


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,1 ENCLOSURE i
l  U.S. NUCLEAR REGULATORY COMMISSION i    REGION IV a
! Docket No.:  50-483 License No.:  NPF-30 l' Report No.:  50-483/97-05 i Licensee:  Unicn Electric Company
 
, Facility:  Callaway Plant i
Location:  Junction Hwy. CC and Hwy. O Fulton, Missouri Dates:  February 10-14 and 24-28,1997 Inspectors:  T. Stetka, Team Leader, 8: .gineering Branch P. Goldberg, Reactor lnspector, Engineering Branch W. Wagner, Reactor inspector, Engineering Branch K. Thomas, Project Manager, Office of Nuclear Reactor Regulation Approved By:  C. V::nDenburgh, Chief, Engineering Branch Division of Reactor Safety ATTACHMENT:  Supplemental Information
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9707010071 970626 PDR ADOCK 05000483 G  ,
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TABLE OF CONTENTS EXECUTIVE SUMM ARY . . . . . . . . . . . . .... .... ............. ........ iii Report Details . . . . ............. ...................... ...... . 1 1. Engineering . . . . . . . ..... ........... .......................... 1 El Conduct of Engineering ............................. ..... 1 E1.1 System Review .............................. .... 1 E1.2 Temporary Plant Modification Review . . . . . . . . . . . . . . . . . . . 3 E1.3 Suggestion-Occurrence-Solution Report Review . . . . . . . . . . . . . 4 E2 Engineering Support of Facilities and Equipment . . . ............. 7 E2.1 Review of Facility and Equipment Conformance to the Final  l Safety Analysis Report Description . . . . . . . . . . . . . . . . . .... 7 l E2.2 Validation and Control of Design Basis Documents . . . . . . . . . . . 8  I E2.3 Engineering Backlog . . . ............................. 9 E CFR 50.59 Implementation ........................ 10 E2.5 System Walkdowns . . . . . . . . . . . . ................... 16 i
E4 Engineering Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 16 E5 Engineering Staff Training and Qualification . . . . . . . . . . . . . . . . . . . . 17 E6 Engineering Organization and Administration ................... 18 E7 Quality Assurance in Engineering Activities ............. ...... 20 V. Management Meetings . . . . .
    .................................. 21 X1 Exit Meeting Summary . . . . . . . . . . . ...................... 21 ATTACHMENT: Supplemental Information    j
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UNITED STATES
 
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EXECUTIVE SUMMARY Callaway Plant l'
y+1 *g NUCLEAR REGULATORY COMMISSION'
NRC Inspection Report 50-483/97-05 This team inspection evaluated the current effectiveness of the licensee's plant and design
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q$  REGloN IV 0 '
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811 RYAN PLAZA DRIVE, SUITE 400   I
engineering organizations to respond to routine and reactive site activities, which included I the identification and resolution of technical issues and problems. This inspection asse:: sed engineering and technical support by focusing on the functional aspects of the component i' cooling water system. The inspect!on also reviewed 10 CFR 50.59 safety evaluations and screenings, engineering evaluations for design modifications, and general engineering performance. The inspection covered a 4-week period with 2 of these weeks conducted onsite,
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  * The conduct of engineering activities was considered to be generally good. Aspects of good engineering practices included strong system engineers, a minimum
AR LINGTON, TEXAS 76011-8064 4'+4 * * * * * #
  , engineering backlog, effective control of plant modifications, good interfaces
JUL I 61997 -
 
l   NOTICE OF LICENSEE MEETING
between engineering and other plant disciplines, a good design basis information i process, and a very effective independent safety engineering group. However, the inspection identified two examples involving Technical Specification interpretations for the diesel generator building supply fans and the refueling machine, wherein the 10 CFR 50.59 safety evaluation screening process was ineffective. The inspection also identified one instance wherein a safety evaluation was not performed for a change to the method of operation of the post-accident sampling system. The inspection also identified two instances where required reports were not made to the NR * Modification packages for the component cooling water system were found to have appropriate safety evaluations and post-modification testing reauirements to assure component operability (Section E1.1).
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* The majority of the request for resolutions reviewed were of gor,d quality. The j request for resolutions had proper engineering justification and proposed corrective ;
Name of Licensee: Union Electric Company    i N vie of Facility: Callaway Plant E, .cket:
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  . 50-483 EA No.: 97-168 Date and Time August 15,1997 at 10:00 a.m. (CDT)
actions (Section E1.1).
of Meeting:
 
* Work packages were found to be performed in accordance with their instructions and no recurrent problems were noted. The team also concluded that no work requests resulted in a modification to the system (Section E1.1).
 
* The temporary plant modification program was found to be 'in conformance with procedures and properly managed (Section E1.2),  j l
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* The majority of the suggestion-occurrence-solution reports had resolutions with proper engineering justifications and proposed corrective actions. One violation was identified for the failure to issue a licensee event report when main steam safety valves had as-found setpoints in excess of the Technical Specification setpoint tolerances (Section E1.3.b.1).
 
* While some discrepancies between the actual plant con' guration and procedures and the Final Safety Analysis Report were noted, an at.oon plan existed to correct such deficiencies (Section E2.1). l
 
* Effective controls were implemented to ensure that c;esign basis documents were available, were being adequately maintained, and were easily retrievable (Section E2.2).
 
* The backlog of engineering work was properly managed (Section E2.3).
 
* Overall, procedural guidance for implementation of 10 CFR 50.59 was appropriat However, the inspection identified two examples in which the guidance was not in accordance with the requirements of 10 CFR 50.59. Specifically, the licensee did l not report safety evaluations for temporary modifications and did not require safety l evaluations when a change was considered to be a plant improvement. The team identified the first example as a violation of 10 CFR 50.59(b)(2) (Sections E2.4.1 and E2.4.1.b.3(1)).
* The implementation of the 10 CFR 60.59 program was adequate; however, the licensee failed to perform a safety evaluation for a modification to the post-accident sampling system. This modification changed the method of operation of the system described in the Final Safety Analysis. Report. This was considered to be an apparent violation (Section E2.4.1b.3(2)).
* The implementation of the Technical Specification interpretation program was adequate; however, the team identified two interpretations that provided guidance that was contrary to the Technical Specification requirements and the Final Safety Analysis Report. in the first example, the interpretation effectively changed the setting of the trip setpoints for the refueling machine without performing a 10 CFR 50.59 safety evaluation. In the second example, the interpretation changed the operation of the diesel generator building supply fans from automatic to manual operation. This change may have increased the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report and increased the possibility for a malfunction of a different type than any evaluated previously in the Final Safety Analysis Report. This change was considered to potentially constitute an unreviewed safety question. Both examples were cited as apparent violations of 10 CFR 50.59 (Sections E2.4.2 and E2.4.2b2).
 
* The plant material condition and housekeeping were good and some improvements in material condition was noted. The boron control program was considered to be effective in improving the plant's material condition (Section E2.5).
 
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Engineering management expectations were clear and understood by engineering personnel. Communications between system engineering and other plant departments were effective. System engineers were knowledgeable of their  l assigned systems (Section E4). I
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The training program for the engineering staff was effectively supporting the role of the system and design engineers (Section ES).
 
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The independent safety engineering group was effective in providing an independent l assessment of plant operations, providing an independent assessment of the effect l of internal and external events on plant operations, and in providing recommendations to improve plant safety. The use of an independent safety engineering group engineer as a shift technical advisor was considered to be an effective and notable application of the group's experience base (Section E6). l
* Self assessments of engineering activities were conducted through the use of quality assurance audits and surveillances. The results of these audits and surveillances were generally consistent with the team's findings (Section E7).
 
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Report Details I. Enaineerina E1 Conduct of Engineering (37550)
E System Review The team reviewed the component cooling water system to verify the licensee's ability to maintain this system in an operable status. The team reviewed the adequacy of the licensee's plant modification process, engineering calculations, problem evaluation requests, and technical evaluation requests. In addition, the team interviewed the system engineer to determine the engineer's knowledge of the syste E1.1.1 Permanent Plant Modification Review Inspection Scone The team reviewed two safety-related plant modification records to verify conformance with applicable installation and testing requirements as prescribed by procedures. Specific attributes reviewed and/or verified by the team included:
10 CFR 50.59 safety evaluations, post-modification testing requirements, safety-related drawing updates, Final Safety Analysis Report updates, training requirements, and field installation, Observations and Findinas The team reviewed Plant Modification Records RFR 16981 A, " Replacement Gages and Transmitters," and RFR 17402A, " Material Change for Component Cooling Water Heat Exchanger End Cover Gasket." These modifications had a proper 10 CFR 50.59 screening or safety evaluation performed and neither represented an unreviewed safety question. The team also found that the post-modification testing requirements were adequate to assure component operability. The team verified that affected drawings and procedures were updated for the plant modification records. In addition, the team verified, through walkdowns, that the physical installations of these plant design changes were consistent with the descriptions in the modification package E1.1.2 Reauests for Resolution Review Inspection Scoce Requests for resolution were used to request technical evaluations, document the evaluation, recommend action, and obtain management concurrence. The team reviewed ten requests for resolution associated with the component cooling water
 
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! system and other selected areas to determine whether proper engineering resolution was performed and that issues requiring the use of the plant modification process i were properly identifie _ Observations and Findinas The team reviewed Request for Resolution 16444, " Operability of Component Cooling Water Pumps with Safety-Related Room Coolers inoperable," which  i requested an engineering evaluation to justify the component cooling water pump 1 operability while the pump room coolers were inoperable. This operation was permitted by Technical Specification Interpretation 35. This request for resolution was originally written for Technical Specification 3.7.12. This Technical  i Specification was subsequently deleted and the requirements incorporated into Final Safety Analysis Report, Section 16.7.4.1. Technical Specification interpretation 35, which was revised so that it applied to the new Final Safety Analysis Report section, specified that the component cooling water pumps could withstand ambient room temperatures of up to 119 F without pump room coolers. The team also noted, however, that the licensee had calculations indicating that the pump room temperature could increase to 128 F during accident conditions without the pump room coolers. The team noted that the request for resolution only stated that the pumps were qualified to 121 F and did not address the effect of the higher 128 F temperature on pump operation. However, following discussions with engineering personnel, the team determined that the pumps were capable of withstanding temperatures in excess of 128 F. T he team concluded that this request for resolution was inadequate because it lacked the pertinent information needed to determine that all aspects of the issue were addressed. The team considered this to be an isolated occurrenc E1.1.3 Review of Enaineerina Calculations inspection Scope The team reviewed the adequacy of two design engineering calculations associated with the component cooling water system to determine whether the calculation assumptions were technically reasonable and properly supporte Observations and Findinas The team found that the licensee's calculations were satisfactory. The calculations reviewed provided sufficient information and assumptions to reach the conclusion stated. The team concluded that the hcensee's calculations were acceptabl .
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E1.1.4 Review of Work Reouests    , Insoection Scope The team reviewed 18 work requests associated with the component cooling water system to determine if repetitive problems existed and to deterrnine the present ;
material condition of the system. This information was compared with the results ,
of the system walkdown. In addition, the work requests were reviewed to j determine if any unauthorized modifications were being performed using work l request Observations and Findinas The team found that the work packages were performed in accordance with their instructions and that the engineering staff was knowledgeable of the work performed. No recurrent problems were noted. The team's walkdown results indicated that the licensee was maintaining the system in good condition and that a very low threshold for deficiency identification had been established. The team did not find any recent work reqwsts that resulted in a system modification, System Review Conclusions
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The team found that the modification packages reviewed included appropriate !
safety evaluations and that post-modification testing was appropriate to assure component operabilit ,
The team concluded that in all but one isolated instance, the request for resolutions reviewed were of good quality. The request for resolutions had proper engineering justifications and proposed corrective action The team concluded that the work packages reviewed were found to be performed in accordance with their instructions and no recurrent problems were noted. The team also concluded that no work requests resulted in a modification to the syste E1.2 Temocrarv Plant Modification Review Inspection Scope The team reviewed temporary plant modifications to verify conformance with applicable installation and testing requirements as prescribed by licensee procedures. Specific attributes reviewed by the team included: 10 CFR 50.59 safety evaluation, license impact review, post-modification testing requirements, plant installation, and the process for periodically reviewing the status of the modification _
 
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. Observations and Findinas The team found that there were eight open temporary plant modifications. Seven of ,
these modifications were nonsafety related and one was safety related. The team !
reviewed Temporary Modification TM-960E010, " Removal of the SR Power Supply i Source from D/P Gauges GKPDIS50028,39,100, and 103 on SGK04A, SGK048, l      )
l SGK05A, and SKGOSB." and found that the modification had the proper safety 1 l
evaluations, license impact review, and that the post-modification testing
; requirements were properly specified. The team also verified that the control room had a copy of the temporary modification and that the affected equipment in the plant was properly tagged. The team found that the temporary plant modification was being tracked for closur Conclusions    l Based on the review of this one temporary modification and the low number of open temporary modifications, the team concluded that the temporary plant modification program was in conformance with procedures and being properly manage E1.3 Suaaestion-Occurrence-Solution Reoort Review 1 Insoection Scope    l l
The licensee issued suggestion-occurrence-solution reports as a means to identify i problems with components and systems and to place these problems in their corrective action system for resolution. The team reviewed 37 suggestion- i occurrence-solution reports to determine the adequacy of the resolution, whether I the component / system operability was properly determined, and that the proposed corrective actions were adequate to preclude recurrence, in addition, the team interviewed the applicable licensee personnel to discuss the resolution of the suggestion-occurrence-solution reports, _ Observations and Findinas During the review of these suggestion-occurrence-solution reports, the team identified instances where main steam safety valves and a pressurizer safety valve were found to have lift setpoints that were out of tolerance. The team found that these findings were not reported to the NR Main Steam Safety Valves
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The team reviewed Suggestion-Occurrence-Solution Report 95-0508, which
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reported that 14 out of the 20 main steam safety valves exceeded their Technical Specification setpoint tolerance of i1 percent when surveillance tests were performed during Refueling Outage 7. The team reviewed the test data and noted that 3 of the 14 valves exceeded the setpoint tolerance by more than 3 percent (+ 3.6, + 3.01, and + 3.4 percentL One of the four steam lines had all five valves outside of the setpoint tolerance and included the two valves with tolerances
 
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l greater than + 3 percent of the setpoint. The licensee's corrective actions were to I
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adjust and re-test tha valves. The re-test indicated that the valves were set l properly. In addition, the licensee directed their nuclear steam system supplier,
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Westinghouse, to perform an analysis of the safety related impacts of having the main steam safety valve tolerance at + 3.6 percent and to provide information to support a Technical Specification amendment submittal to the NRC to increase the valves' setpoint tolerance to + 3, -1 percen When questioned, the licensee indicated that a licensee event report had not been issued because they considered that exceeding the setpoint tolerance occurred at time of discovery and not during the operating cycle. In addition, the licensee stated that the three main steam safety valves that had opening setpoints greater than the +3 percent tolerance, exceeded their safety analysis assumptions and the one valve that had a setpoint less than -1 percent of the tolerance exceeded the ]
component fatigue analysis assumptions. However, the licensee also stated that i preliminary evaluations indicated that the excessive setpoints were enveloped by the existing safety analyse The team concluded that due to the number of failures, it was unlikely that the main i steam safety valves failed at the time of discovery. The team also concluded that during Refueling Outage 7, two independent trains became inoperable in the main steam system that were designed to mitigate the consequences of an acciden l The licensee disagreed that these valve failures needed to be reported. Their basis for this disagreement was that they complied with the requirements of 10 CFR 50.73 and the guidance provided in NUREG 1022, Revision 0, " Event Reporting Guidelines for 10 CFR 50.72 and 50.73," which specified that when  l f ailures occur, the failures are assumed to have occurred at the time of discovery and not during the operating cycle, in a November 2,1993, memorandum issued by the Office of Nuclear Reactor Regulation, which was sent to the existing eight Region IV power reactc,r licensees, the staff stated that the guidance in NUREG 1022, Supplement No.1, " Licensee Event Reporting System," was clear that if conditions were discovered during an outage, but were believed to have existed during operation, they were reportable, as long an applicable threshold for reporting was reached. Although the licensee was not a Region IV licensee at the time the letter was sent, the licensee obtained a copy of the letter from another plant and was aware of the NRC positio CFR 50.73(a)(2)(vii) requires that an event be reported when a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems designed to mitigate the consequences of an accident. The failure to issue a licensee event report for this occurrence was considered to be a violation of 10 CFR 50.73 (50-483/9705-01).
 
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b.2 Pressurizer Safety Valves ;
I l  The team reviewed Suggestion-Occurrence-Solution Report 96-1273, which  1 l  reported that, while performing surveillance testing during Refueling Outage 8, one l  pressurizer safety valve opened at -2.31 percent. This opening pressure exceeded the Technical Specification setpoint tolerance of t1 percent. The licensee's corrective action was to adjust and re-test the valve. The adjustment was I completed and the valve had a satisfactory re-tes While reviewing the pressurizer safety valve data, the team also determined that, prior to Refueling Outage 8, the licensee upgraded their inservice testing program to incorporate the 1989 edition of the ASME Code, Section XI. This ASME Code edition does not require increasing the sample size unless the as-found valve setpoint exceeds the setpoint criteria by 3 percent or greater, even though the Technical Specification setpoint tolerance is il percent. Due to the change in code years for inservice testing, the licensee did not test the other two pressurizer safety valves. The team discussed this with the licensee and determined that the Final Safety Analysis Report, Chapter 15, safety analysis was based on the Technical Specification tolerance of 1 percent and not the ASME Code allowed tolerance of i3 percen While reviewing Suggestion-Occurrence-Solution Report 96-1273 regarding the failed pressurizer safety valve, the team noted that the plant could have exceeded the Chapter 15 safety analysis. The licensee received a preliminary analysis from their vendor (Westinghouse) in letter SCP-97-105, dated February 26, 1997. This letter concluded that the out-of-tolerance valve was enveloped by the accident analysis. However, the letter also stated that a change in the  i pressurizer safety valve setpoint from -1 to -3 percent would require a change to one of calculations before permanent implementation of an increased setpoint tolerance. The final Westinghouse analysis to support an increased setpoint tolerance for the pressurizer safety va!ve and the licensee's plant-specific calculations will be reviewed when available. to verify that the increase to a -3 percent setpoint would still be enveloped by the accident analysis. This is considered to be an inspection followup item (50-458/9705-02). Conclusions The majority of the suggestion-occurrence-solution reports had resolutions with proper engineering justification and the proposed corrective actions were adequat A 10 CFR 50.73 violation was identified for the failure to issue a licensee event report when the as-found rnain steam safety valve setpoints exceeded the Technical Specification setpoint tolerance . .- .. - -- . - - . .. .~ - . - - _ . .
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E2 Engineering Support of Facilities and Equipment (37550)
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E2.1 Review of Facility and Eauioment Conformance to the Final Safety Analysis Report
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Descriotion 4 Inspection Scope
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A recent discovery of a licensee operating its facility in a manner contrary to the Final Safety Analysis Report description highlighted the need for a special focused 1 review that compares plant practices, procedures and/or parameters to the Final
, Safety Analysis Report descriptions. As the result of this discovery, the inspectors reviewed selected sections of the Final Safety Analysis Report.
 
4 Observations and Findinas    ,
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The team identified three discrepancies between the Final Safety Analysis Report and the actual plant configuration. These discrepancies are discussed in
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Sections E2.4.1.b.3(2), E2.4.2.b.1 and E2.4.2.b.2 of this report.
 
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, The team also noted that there was a previous NRC finding that identified that the component cooling water system temperature was below the lower limit
Location of Meeting: Region IV Office, Arlington, Texas  j 4th Floor, Training Conference Room  I Purpose of Meeting: Pre-decisional Enforcement Conference to Discuss the Failure to Perform Safety Evaluations in Accordance with 10 CFR Part 50.59 Requirements identified in Inspection Geport 50-483/97-
{ specified in the Final Safety Analysis Report (as described in NRC Inspection
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Report 50-483/96-11). As a result of this finding, the licensee reviewed the
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component cooling water system safety system functional assessment that was
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previcusly performed. This review was initiated because the assessment was limited to a design basis review instead of a review of all areas of the Final Safety 1 j Analysis Report that pertained to component cooling water. This review was  )
j conducted prior to this inspection to determine if cdditional problems existed. As a result of the review, the licensee identified several discrepancies between the Final
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Safety Analysis Report and the actual configuration and operation of the component I
cooling water system.
 
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Since the licensee had conducted seven safety system functional assessments, the licensee further expanded their review to include the safety system functional assessments that were previously conducted on the essential service water system and the auxiliary feedwater system. These reviews also identified several  )
discrepancies between the Final Safety Analysis Report and the actual configuration l and operation of these systems. Subsequent to this inspection, the licensee  '
decided to review the remaining four safety system functional assessments as a part of a task team formed in March 1996 to review all sections of the Final Safety Analysis Repor The purpose of this task team was to identify and prioritize sections of the Final
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Safety Analysis Report for a compliance review against plant hardware and procedures. The task team completed its review in July 1996 and identified actions and prioritized sections for further review. These additional reviews were scheduled for completion prior to the end of 1998. This licensee effort was documented in their letter ULNRC-03530 dated February 5,1997, to the NRC regarding an NRC  j
 
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enforcement policy revision. Enforcement Guidance Memorandum EGM-96-005, dated October 21,1996, set forth a revised enforcement policy applicable to voluntt.ry licensee efforts to correct inconsistencies in licensing documents, including programs for licensee reviews of the Final Safety Analysis Repor The licensee's efforts to determine the extent of Final Safety Analysis Report discrepancies were ongoing during this inspection. Further review of this effort will be conducted during future inspection efforts. This is considered to be an inspection followup item (50-483/9705-03), Conclusions l
While some discrepancies between the actual plant configuration and procedures l and the Final Safety Analysis Report were noted, the team concluded that the licensee had a process underway to identify and correct such deficiencie E2.2 Validation and Control of Desion Basis Documents inspection Scope The team reviewed the licensee's controls of design basis documents to determine if the documents were available, maintained, validated, and were easily retrievable, Observations and Findinas The team found that the licensee's program for identification and control of design basis documents was described in Procedure EDP-ZZ-04055, " Design Basis Control." This procedure also described the sources of design basis information and how this information was located, validated, and maintained for future use. The team found that system design basis validation was accomplished through safety system functional assessments performed by the quality assurance organizatio The licensee informed the team that seven safety system functional assessments were performed between 1988 and 1995, which validated 17 of 45 safety-related system The team reviewed the license's response to the NRC request for information pursuant to 10 CFR 50.54(f) regarding adequacy and availability of design basis information. The licensee response was documented in letter ULNRC-3531, dated February 6,1997. The team found this letter contained two licensee commitments for future work intended to verify the adequacy and availability of design basis information. Those commitments were: (1) initiate a review of the Callaway Plant in accordance with the Nuclear Energy Institute initiative as described in Nuclear Energy Institute 96-05; and (2) perform two safety system functional assessments I
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, by December 31,1998. The team found that the Nuclear Energy Institute review focused on licensing basis information, which included a sample review of Final
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Safety Analysis Report information, whereas, the safety system functional assessments focused on design basis information to support the as-huilt configuration of the plant.
: The licensee informed the team that they would evaluate the results of their reviews
- to determine if they had reasonable assurance that the original design basis would be maintained for future use. The licensee stated that the need for additional validation of design basis information would also be based on this evaluatio The team observed implementation of the licensee's program for maintaining, updating, and retrieving design basis information for the emergency service water
; system, the residual heat removal system, and the auxiliary feedwater system. The documentation was adequately maintained and easily retrieved.
 
' Conclusions The team concluded that the licensee implemented effective controls to ensure that design basis documents were available, were being adequately maintained, and were easily retrievable.


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NRC Attendees:  E. Merschoff, Regional Administrator A. Howell, Director, DRS K. Brockman, Deputy Director, DRP C. VanDenburgh, Chief, Engineering Branch, DRS W. Johnson, Chief, Projects Branch B, DRP T. Stetka, Senior Reactor inspector, Engineering Branch, DRS K Thomas, Project Manager G. Sanborn, Enforcement Officer Licensee Attendees: G. Randolph, Vice President and Chief Nuclear Officer C. Naslund, Manager, Nuclear Engineering R. Affolter, Manager, Callaway Plant A. Passwater, Manager, Licensing and Fuels J. Laux, Manager, Quality Assurance K. Kucchenmeister,' Superintendent, Design Engineering T. Sharkey, Supervising Engineer, Safety Related Mechanical Systems f
E2.3 Enaineerina Backloa
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, Insoection Scope The team evaluated the extent of backlogged engineering work to determine the size of the backlog and to determine whether it was being properly managed, a Observations and Findinas The team found that the engineering backlog consisted of 238 suggestion-occurrence-solution reports,138 request for resolutions,153 central action tracking items, and 151 modification packages ror a total of 680 items. The team found that the licensee had a process to set priorities, such that, work and resources were allocated first to the most significant items. This process assigned a weighing factor to set the priority within a category. For example, a weighing factor of five was assigned for category items involving nuclear, industrial, or radiological safety. The lowest weighing factor was a two, which was assigned to items involving management discretion. The team did not identify any safety-significant issues that were not being properly resolved. The team reviewed the trend of the backlog items and found that the number of backlog open items had remained relatively constant over the last 12 months. The team noted that there were no old open items that were safety significan . - . .
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! Conclusions


The team concluded the licensee was effectively managing the backlog of engineering wor E CFR 50.59 Imolementation (37001)
E2.4.1 10 CFR 50.59 Program    ;
1 Insoection Scooe l
The team reviewed the licensee's 10 CFR 50.59 program guidance, 20 screenings that concluded that a safety evaluation was not required, and 16 safety evaluation l The screenings and safety evaluations were associated with permanent and temporary rnodifications to the plant and procedures, request for resolutions, and Final Safety Analysis Report change notice l Observations and Findinas Administrative Reauirements The licensee's safety evaluation process for changes to the facility was controlled by Procedure APA-ZZ-00140, " Safety, Environmental and Other Licensing Evaluations." This procedure delineated the methods and responsibilities to  l determine and document whether procedure and facility changes could be made  I without prior NRC approval. The licensee's safety evaluation process began with a safety evaluation screening that utilized specific screening criteria. This screening was performed to determine whether the proposed activity needed additional review to determine if an unreviewed safety question existe Procedure APA-ZZ-00140 provided this screening criteria in the form of questions that were answered by a reviewer. Specifically, these questions were: (1) the activity did not change the facility or a procedure as described in the Final Safety Analysis Report; (2) the activity was not a test or experiment not described in the Final Safety Analysis Report; and (3) the activity did not involve a change to the Technical Specifications. If the results of this screening concluded that one or more of the screening criteria was not satisfied, the process then required that a formal safety evaluation be performed to assess the merits of the activity and to determine whether an unreviewed safety question existed. The unreviewed safety question determination was documented in this safety evaluation. If it was determined that an unreviewed safety question existed, then NRC approval was required prior to implernenting the proposed chang m
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4 The team found the documentation contained in the safety evaluations that were performed to be sufficiently detailed and the conclusions logically supported. While the team determined that overall, the guidance contained in Procedure APA-ZZ-00140 was adequate, the team identified problems with this guidance and with the implementation of the 10 CFR 50.59 proces b.2 Reoortina of Plant Chances During a review of Procedure APA-ZZ-00140, the team noted that the procedure stated that short-term modifications (e.g., temporary modifications) did not fall within the periodic reporting requirements. A review by the team of the  i 10 CFR 50.59 safety evaluation reports submitted to the NRC, confirmed that the  l safety evaluations for temporary modifications were not reported. As the result of discussions with licensee personnel, the team determined that the licensee had not reported these safety evaluations since June 14,1988. The licensee further stated that the decision to not report safety evaluations for the temporary modifications was an error. The team identified that Temporary Modification 95-MOO 2,
  " Temporary Filter and Piping for BTRS Chill Water Loop," was an example of a temporary modification that was not reporte CFR 50.59(b)(2) requires licensees to submit a report containing a brief description of any changes, tests, and experiments, including a summary of the  I safety evaluation of each. This report was required to be submitted annually or along with the Final Safety Analysis Report updates. Since 10 CFR 50.59(b)(2)
does not differentiate between long- and short-term modifications, all safety  i evaluations for modifications are required to be reported. The f ailure to report temporary modification safety evaluations is considered to be a violation of 10 CFR 50.59 (50-483/9705-04).
b.3 Performance of Safety Evaluations During the review of documentation involving screening to deterr.iine if safety evaluations were required, the team identified the following two examples where safety evaluations were not performed:
  (1) The team noted that Procedure APA-ZZ-00140 stated that, if the design, function, or method of performing the function of an associated system, structure, or component was either unaffected or improved, there was no change in the f acility as described in the Final Safety Analysis Report. Since 10 CFR 50.59 requires a safety evaluation for all changes to the facility irrespective of whether or not the change is believed to be an improvement, the team considered this procedure guidance to be inconsistent with the 10 CFR 50.59 requiremen The team identified one instance where this guidance was used as justification for not performing a safety evaluation. Modification CMP 95-1027A upgraded the source of power for instrument cabinet fans from nonsafety related to safety related. The licensee documented that this
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modification was an improvement; therefore, a safety evaluation was not performed. However, a subsequent review of this modification by the team determined that the modification did not change the racility as described in the Final Safety Analysis Report. Therefore, a safety evaluation was not required. Nevertheless, the team considered the guidance in Procedure APA-
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ZZ-00140, regarding improvements, to be misleadin :
(2) Modification RMP 94-2005A was implemented to redesign the post-accident  l sampling system by replacing the computer control of the sample panel with j manual control. Before the modification, the system was manually initiated '
by selecting a particular analysis. The computer would then automatically
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position the valves, as necessary, to accomplish the analysis. Due to problems with softwear, the computer controls did not work properly. After :
q Union Electric C Meeting August 15,1997
the modification, the computer was removed and the necessary lineups to perform specific analyses were performed by personnel following specific analysis procedure The safety evaluation screening, performed on November 1,1995,
. concluded that a safety evaluation was not required. The conclusion stated that Final Safety Analysis Report Change Notice 94-05, issued in February i 1994, and Change Notice 94-23, issued in July 1994, were already approved and incorporated in the Final Safety Analysis Report and that the modification involved nonsafety-related equipment. Therefore, the modification did not require any additional Final Safety Analysis Report change The team noted that Change Notices 94-05 and 94-23 involved a change to the chemical analyses that were being performed by the post-accident
; sampling system. The analyses for atmosphereic oxygen, dissolved oxygen,
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pH, conductivity and in-line chloride analysis were eliminated. However, the team also noted that the change from a computer controlled analysis to a manually controlled analysis was not identified in these change notice When this finding was discussed with the licensee, the licensee stated that the change was not included because the description regarding the computer controlled operation of the post-accident sampling system was only described in a letter to the NRC, dated November 4,1983. Since this letter was only referenced by the Final Safety Analysis Report, they did not consider the operation described in this referenced letter to be a part of the Final Safety Analysis Repor The team did not agree with the licensee's position. The February 4,1983, letter referenced by Section 18.2.3 of the Final Safety Analysis Report provided additional details on the post-accident sampling system operatio This information was used by the NRC to determine acceptability of the design of the post-accident sampling system. Specifically, the information contained in this letter was referenced in Supplement 3 to the NRC's Safety
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Evaluation Report. Therefore, the inspectors concluded that the method of operation of the system, as described in the Final Safety Analysis Report,
 
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was changed by the modification, and a safety evaluation was not performe CFR 50.59(b)(1) requires the performance of a safety evaluation when plant modifications change the plant as described in the Final Safety Analysis Report. The failure to perform this safety evaluation is considered to be an apparent violation of 10 CFR 50.59 (50-483/9705-05). Conclusions While the team concluded that the licensees procedural guidance for implementation
, of 10 CFR 50.59 was adequate, the team identified two areas where this guidance was not in accordance with 10 CFR 50.59. These involved the failure to report safety evaluations for temporary modifications and not performing a safety
, evaluation when a change was considered to be a plant improvement. The team also concluded that the implementation of the 10 CFR 50.59 requirements was
 
adequate; however, the team 'dentifica an apparent violation involving the f ailure to perform a safety evaluation for the change in the automatic operation of the post-J accident sampling system.
 
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E2.4.2 Technical Soecification Interpretations
: insoection Scone
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As a result of a concern identified in October 1996 at another facility regarding Technical Specification interpretations that were found to be in conflict with Technical Specification requirements, the team reviewed 41 of the licensee's Technical Specification interpretations to ensure that these interpretations did not conflict with Technical Specification requir9ment Observations and Findinas The tec.m fourid that the Technical Specification Interpretation Program was controlled by Procedure APA-ZZ-00104, " Technical Specification Interpretations and Notice of Enfcrcement Discretion." Procedure APA-ZZ-00104 defined a Technical Specification interpretation as a formalinterpretation that provided guidance for both the Technical Specifications and Section 16 of the Final Safety Analysis Report. This procedure also specified that all Technical Specification interpretations were reviewed by the onsite review committee and approved by the plant manage The licensee informed the team that, as a result of the concern identified in October 1996, the onsite review committee performed an additional review of the Technical Specification interpretations to ensure that the interpretations did not conflict wRh Technical Specification requirements. Nevertheless, the team identified an example in which an interpretation provided guidance that potentially violated the requirements of the Technical Specifications, in addition, the team also identified an example in which an interpretation p;ovided guidance that was inconsistent with the Final Safety Analysis Repor . - _ . - . _ _ . .  . _ . _ --
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. Technical Soecification Interpretation 25 Section 16.9.2 of the Final Safety Analysis Report described the limits for setting the overload and load reduction trip setpoints for the refueling machine-at 250 pounds above and below the weight of the suspended loads, respectively.
 
Technical Specification Interpretation 25 interpreted Section 16.9.2 to mean that these trip setpoints could be set to 250 pounds above the heaviest fuel assembly
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load for the overload trip and 250 pounds below the lightest fuel assembly load for the load reduction tri The team was concerned that this interpretation allowed these trip setpoints to be set in excess of 250 pounds by approximately 150 pounds (the estimated weight of a rodded assembly). This meant that, while the overload trip would be correct for a rodded fuel assembly, it would be excessive for the unrodded fuel assembly and would not trip until the weight of the suspended load was 400 pounds above the suspended load weight. It also meant that, while the load reduction trip would be correct for the unrodded fuel assembly, it would be excessive for the rodded fuel assembly and would not occur until the insertion force was 400 pounds less than the suspended load weigh Through discussions with licensee personnel, the team determined that these were the trip setpoints used during Refueling Outage 8. Additional review by the team indicated that on October 20,1995, Technical Specification 3.9.6, which provided the same trip setpoint setting requirements, was deleted and the requirements incorporated into Final Safety Analysis Report 16.9.2. Therefore, during the period !
of October 18,1984, through October 20,1995, Technical Specification 3.9.6 was I violated during seven refueling outages (Refueling Outages 1 through 7).
 
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10 CFR 50.59 (b)(1) requires the performance of a safety evaluation when plant l modifications change the plant as described in the Final Safety Analysis Repor The failure to perform this safety evaluation is considered to be an apparent violation of 10 CFR 50.59 (50-483/9705-06).
 
b.2 Technical Soecification Interpretation 18 Technical Specification Interpretation 18 provided an interpretation regarding l the operation of the diesel generator building supply fans. Technical i Specification 3.8.1.1b required the diesel generators to be operable. In l addition, Technical Specification 1.19 required that for a system to be operable, I all supporting subsystems must also be operable. The diesel generator building l supply fans are a subsystem of the diesel generators that the licensee determined are required to be operable when outside ambient temperature is greater than 65 F. ;
Final Safety Analysis Report, Section 9.4.7.2.3, stated that the diesel generator l building supply fans automatically start when the room temperature exceeds 90 F and automatically shut down when room temperature falls below 86 F. If the building temperature exceeded 90oF, Final Safety Analysis Report, Section 16.7.4, allowed the temperature to rise to a maximum of 119 F. At 119 F, Section 16. required the temperature to be lowered below 119oF within 8 hours or to perform
 
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an analysis to demonstrate that equipment was not affected by the elevated
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temperatures. In addition, Section 16.7.4 also required that if the temperature exceeded the 119 F limit by 30 F (149 F) for more than 4 hours, the affected 4 equipment (i.e., the diesel generators) were to be considered inoperable.
 
i The licensee developed Technical Specification Interpretation 18 to allow the
, diesel generator building supply fans to be placed in manual operation (i.e.,
defeating the automatic starting function by placing the supply fan control switches in a " pull-to-lock" position), without declaring the diesel generators inoperable when the outside ambient temperature was greater than 65 F. The purpose for this interpretation was to eliminate excessive cooling of the diesel generator jacket
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water system that was causing system alarms. In addition. the licensee revised Procedure OTN-NE-00002, " Standby Diesel Generator Auxiliary Systems," to add a precaution and limitation (Step 2.6), which stated that each diesel generator building supply fan was considered capable of performing its intended safety function (i.e., the ability to supply air to the building if temperatures rise
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above the fan start setpoint) if the fan was placed in pull-to-lock and was under
, the control of the operator. The procedure also directed the operator to assign
! the room temperature point to annunciate on Window 65F, " Optional Parameter Setpoint," at or below 110 F. However, the licensee concluded, through the 10 CFR 50.59 screening for the procedure change, that a safety evaluation for the change was not required. Therefore, a safety evaluation for the procedure change was not performe Since the diesel generator building supoly fans were a subsystem of the diesel generators and these fans were considered to be inoperable when they were in a
" pull to-lock" condition, the team concluded that the diesel generators were also inoperable. Therefore, in effect, Technical Specification Interpretation 18 changed
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the diesel generator Technical Specifications by allowing the diesel generators to be declared operable while the diesel generator buildiqq supply fans were inoperable.
 
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Based on this finding, the team requested that the licensee review the operations logs to determine when the Technical Specification interpretation / procedure guidance was implemented. Although licensee representatives stated that the interpretation was used to place the fans in manual operation during the Fall and Spring evenings from 1987 to 1990, the licensee's review of the operator logs from 1987 to 1990, revealed that there were no documented instances in the operator logs in which the f ans were placed in pull-to-lock. In addition, since a plant modification in 1990 eliminated the need to place the fans in manual operation, che licensee interviewed five operators to determine if they recalled any instances of placing the fans in manual operation since 1990. As a result of these interviews, the licensee informed the team that no operator interviewed recalled placing the f ans in manual operatio Since the system was described in the Final Safety Analysis Report as operating automatically, it appeared that a safety evaluation was required to substitute the manual operator action for this automatic function. This substitution may have increased the probability of occurrence of a malfunction of equipment importam to
 
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l 10 CFR 50.59 (b)(1) requires tne performance of a safety evaluation when plant :
E-Mail To: FOR (EWM, ATH, KEB, CAV, WDJ, TFS, GFS)NRC Attendees PMNS Mtg Announcement Coordinator JKR E. L. Jordan, DEDO PAB  H. L. Thompson, DEDR FJM  F. J. Miraglia, A/D/NRR    l ACT A. C. Thadani, ADT/NRR RPZ R. P. Zimmerman, ADP/NRR (TWA) Project Manager, NRR OEMAll J. Lieberman, D/OE    ,
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LJC  L. J. Chandler, OGC    I (DP)  D. Passehl, Resident inspector CAH  C. A. Hackney, RSLO    j DMK1  D. M. Kunihiro, WCFO    I GFS  G. F. Sanborn, EO BWH  B. Henderson, PAO MFH2  M. Hammond, PAO
i modifications change the plant as described in the Final Safety Analysis Repor The failure to perform this safety evaluation is considered to be an apparent violation of 10 CFR 50.59 (50-483/9705-07),
, KEP  Ken Perkins, Director, WCFO JAC1, CMS, CJG RA Secretaries LAT, DLF, CAM 3 DRP Division      '
i j- Conclusions      !
CLG, LMB  DRS Division NLH, WSW  DNMS Division LSO  Linda Ousley cc:        )
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. DMB (IE 45)g
Overall, the team concluded that the implementation of the Technical Specification
* 'OEDO RIV Coordinator (17G21)
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Receptionist ORA File RIV File DOCUMENT NAME:
interpretation. program was adequate. However,' the team identified two examples in which interpretations provided guidance that was inconsistent with the Final
To receive copy of document, Indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy RIV: SRI:EB l E C:EB  l E D:DRP PAO gl RSLOLM{1 D:DRS l TFStetka/Imb" CAVanDenburgh" TPGwynrd BHendersonk CAHackney  ATHowell lip 07/11/97 07/11/97  07/11/97 07/l(/97 07/% 97 07/4/97 OFFICIAL RECO 3D COPY
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Safety Analysis Report and potentially violated the requirements of the Technical j Specifications. One example involved an apparent violation for the failure to l
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;- perform a safety evaluation for a change to the refueling machine load setpoint l
!  The second example involved an apparent violation for the substitution of manual l
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operator action for the automatic operation of the diesel generator building supply fan E2.5 Svetem Walkdowns a Insoection Scope At different times during the inspection, the team performed walkdowns of selecte plant areas to determine the overall material condition of equipment and the maintenance of housekeepin Observations and Findinos During the walkdowns, the team noted a number of tags on pumps and valves that were called " boric acid" tags. The licensee stated that they started a boron control program in January 1997. The purpose of the program was to hang such tags on components to identify that there was some leakage that required occasional cleaning, but the affected valves were not in need of immediate repair. The licensee stated that the purpose of the program was to differentia'.e between acceptable periodic residue removal on components and those ccmpoaents that needed to be repaired. For components that needed repair, work request tags were hung. The team noted that these boric acid tags were not limited to boric acid problems and extended to other leakage problems as well. This explained the existence of such tags on such systems as the component cooling water system, which was not a borated water system. Based on these observations, the team considered the boron control program to be innovative and effective toward maintaining the material condition of the plan .. . .  .. -
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c Conclusions The team's walkdown of the plant indicated that the material condition of the plant was good and that some improvements were noted. Housekeeping was also noted to be good. The boron control program was considered to be effective in maintaining the plant's material conditio E4 Engineering Staff Knowledge and Performance (37550)
a. inspection Scope The team interviewed the nuc' ear engineering department manager, one engineering supervisor, five system engineers, and one design engineer. Interview topics included management expectations for staff engineers, training regarding system interrelations, and interface with other plant organizations. The team also questioned staff engineers about knowledge of their assigned systems and conducted system walkdowns with the system engineers. In addition, a detailed walkdown of the component cooling water system was conducted with the associated system engineer to determine the level of knowledge of this engineer.
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b. Observations and Findinas The team found management expectations for staff engineers were clearly define System engineers were effectively coordinating with design engineering to evaluate and improve their specific systems. System engineering personnel indicated that communication and cooperation with operations, maintenance, and design engineers were effective in resolving work issues and also in assuring that modifications were properly installed. Team interviews and plant walkdowns indicated that system engineers were knowledgeable of their assigned system The detailed walkdown of the component cooling water system with the system engineer indicated that this system engineer spent approximately 25 percent of the time in the plant performing system walkdowns, witnessing surveillance tests, and witnessing maintenance activities. The system engineer trended flow versus differential pressure for the component cooling water and essential service water pumps to detect pump degradation. The system engineer indicated that their identification of a degrading trend in one of the essential service water pumps will result in the replacement of this pump during the next refueling outage. The team determined that the system engineer was also trending for fouling in the heat exchangers. This engineer also explained component deficiencies in detail and discussed specific problems with system operation. This walkdown further confirmed that system engineers were knowledgeable of their system _ . _ _ _ _ _ _ _ _ . _ _ _ . _ .. _ _ _ . - - _ _ . _ _ _ _ _ _ _ _m_
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. Conclusions The team concluded that engineering management expectations were clear and understood by enginee-ing personnel. Communications between system engineering and other p ant departments were affective. System engineers were knowledgeable of their assigned system E5 Engineering Staff Training and Qualification (37550) Inspection Scope      .
l The team reviewed the licensee's training and certification program requirements for the engineering staff. This review included a review of training records fo- -9 engineering staf ,
i b.~ Observations and Findinas    l The team found that all 35 design engineers had completed their qualification  j modules to ensure that they possessed sufficient knowledge and skills to  {
independently perform their assigned tasks. The team found that 16 out of  I 23 system engineers completed their qualification modules. The team noted  l that engineering management's goal was to have all system engineers qualified by l June 1998. The team's review of the training records and schedules indicated that engir eering management was on track for completing the system engineer qualification Conclusions The team concluded that the training program for the engineering staff was effectively supporting the role of the system and design engineer !
E6 Engineering Organization and Administration (37550) insoection Scoce The team evaluated the overall effectiveness of the independent safety engineering group by reviewing selected reports, interviewing independent safety engineering group personnei, and by determining if issues identified by the independent safety l engineering group were corrected or in the process of being correcte ! Observations and Findinas in addition to their Technical Specification required activities, the independent safety l engineering group provided an independent review of plant operating activities to ]
detect potential operational problems. To accomplish this operational experience ;
mission, the group performed operating experience reviews and analyzed industry
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events for applicability to the Callawev Plant. The indecendent safety engineering group also reviewed and analyzed internal plant events to determine the corrective actions needed to prevent recurrenc The independent safety engineering group consisted of eight engineers and a supervising engineer. Of these eight engineers., six were also qualified shif t technical advisors. Each week one of these six shift technical advisors stood a day shift watch in the control room. The team considered the practice of having an independent safety engineering group engineer performing shift technical advisor duties to be an effective method for the independent safety engineering group to keep abreast of ongoing shift operation As the result of their operating experience reviews, the independent safety engineering group issued periodic, " Operating Experience Journals," to provide information to plant personnel regarding events that have occurred at the Callaway Plant and in the industry. The independent safety engineering group also performed operating experience crew briefings to assure that plant personnel were aware of operating experience issues. The team reviewed four operating experience journals and six operating experience crew briefing sheets. As the result of these reviews, the team considered these reports to be informative and effective at keeping plant personnel informed of operating experience event As required by the Technical Specifications, the independent safety engineering group issued monthly reports to the quality assurance manager and the plant manager. The team reviewed the independent safety engineering group monthly reports for the period of August 29,1996, through February 3,1997. The team found these reports to meet the requirements of the Technical Specifications and to be indicative that the independent safety engineering group provided independent assessments of on-going plant activitie The team also reviewed a listing of suggestion-occurrence-solution reports written by independent safety engineering group engineers to determine the group's involvement in identifying plant problems. The licensee's suggestion-occurrence-solution reporting system is used to p ovide documentation of plant problems into their cotrective action system and to assure that these problems are tracked for resolution. The team was informed tnat the independent safety engineering group wrote approxirnately 10 percent of all the suggestion-occurrence-solution reports generated. The team confirmed this approximation. There were 229 suggestion-occurrence-solution reports written by the independent safety engineering group engineers for the period of January 1996 through February 199 Of these 229 reports, approximately 87 were open, but not overdue, and were being processed for closure. Five were Open and overdue. The team reviewed the five overdue reports and an additional six other reports that were still open and did not identify any problems. This review indicated that the independent safety engineering group was identifying plant issues and that these issues were being tracked to completion.
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j l  The team also reviewed the licensee's listing of items entered in their centralized l  action tracking system by independent safe':y engineering group personnel to determine the extent of the group's involverient with the system. The centralized action tracking system was used to track istues identified in the industry. The team
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Union Electric C I Meeting August 15,1997 I
noted that a total of 64 items were written t y the independent safety engineering group over the past year. Of these items,311 were closed and 26 were still open.
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NOTE:
]  Of the 26 open items,14 were overdue for closure. Thase 14 items were reviewed 1  with licensee personnel. Based on this review, the team determined that they were l  properly classified as low priority and were bring tracked for resolution, i
  (1) This meeting is open to attendance by members of the general publi (2) N?.0 personnel, not listed above, that desire to attend this meeting should notify Mr. T. Stetka at (817) 860-8247 by COB on August 8,1997, l
The team's interviews of four independent saiety engineering group engineers and the group's supervising engineer did not identify any issues. All personnel interviewed indicated that they were qualified to perform their assigned activities
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and that they were proactive toward resolving plant issues. In addition, they felt that they had a low threshold for writing suggestion occurrence-solution reports and i  that plant personnel were responsive to their fi1 dings, Conclusions The team concluded that the independent safety engineering group was effective in providing an independent assessment of plant o?erations, providing an independent assessment of the effect of internal and externa; events on plant operations, and in
:  providing recommendations to improve plant safety. The team considered the use of an independent safety engineering group engineer as a shift technical advisor to be an effective and notable application of the group's experience base.
 
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E7 Quality Assurance in Engineering Activities (37550)
u Inspection Scooe i
The team reviewed six quality assurance audit reports and eight quality assurance a  surveillance reports of plant engineering that were performed from December 1994 through August 1996. These reports were reviewed to evaluate the effectiveness of the licensee's process to self identify and resolve plant problem Observations and Findinas
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The team found that the licensee considers their self assessments to be the audits and surveillances performed by the quality assurarle department. These audits and surveillances were conducted at the request of the nuclear engineering departmen The team noted that 27 self assessments of engineering activities were performed during the period of 1994 through 1996. The team found that the quality assurance audit and surveillance findings were generally consistent with those
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identified by the tearn. Specifically, the team noted that system engineers were
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1 knowledgeable and qualified in their systems and were effective in identifying plant I problems. However, the team also found that self-assessment findings were not consistent with the team's findings in that the self-assessments did not identify the failure to perform 10 CFR 50.59 safety evaluations as discussed in Section E2.4 of I this repor The team found that the responses to the quality assurance audits and surveillances l were timely and acceptable. An example of this responsiveness was evidenced by l Surveillance Report SP96103. The team reviewed the discrepancies identified in ,
this report and found that they were properly evaluated and that necessary changes I to the Final Safety Analysis Report and/or plant procedures were being implemente c. Conclusions The team concluded that the self assessmei.ts of engineering activities were generally consistent with the team's findings except in the area the 10 CFR 50.59 safety evaluations. The team concluded that licensee responses to identified discrepancies were timely and acceptabl V. Manaaement Meetinas X1 Exit Meeting Summary The team presented the inspection results to members of licensee management at the conclusion of the inspection on February 28,1997. In addition, a final exit meeting was held on June 24,1997. The licensee acknowledged the findings presented. During both meetings, the licensee stated the following objections with regard to the inspection findings:
  * The licensee disagreed with the violations for a failure to report the inoperability of main steam safety valves as discussed in Section E1.3.b.1. The licensee's position was that they comply with the requirements of 10 CFR 50.73 and the guidance provided in NUREG 1022, Revision O. They also do not believe that they need to comply or are committed to guidance prouded to other licensees. (This statement was made in reference to a lett.sr dated December 8,1993, that was sent to the existing Region IV power reactor licensees by Samuel J. Collins regarding the interpretation of reporting requirements for setpoint drifting of main steam and pressurizer safety valves.)
 
* The licensee did not agree that the guidance in Procedure APA-ZZ-00140, which states that a safety evaluation was not required for plant
" improvements," was misleading as discussed in Section E2.4.1.b.3.(1).
 
The licensee's position was that the statement would not result in a plant change that affected the design, function, or method of performing the function in an associated syste . __ - -. . _ - - . . - -_ __ _ _ . . _. . .- __ _ _ _
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The inspectors asked the licensee whether any materials examined during tlie inspection should be considered propriotary. The licensee identified some information reviewed by the team that was considered to be propriety. The team was aware of this information, which involved maximum flows through component cooling water heat exchangers, and stated that this information had no bearing on inspection results and would not be discussed in the repor !
 
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ATTACHMENT SUPPLEMENTAL INFORMATION -
PARTIAL LIST OF PERSONS CONTACTED Licensee R. Affolter, Plant Manager D. Bono, Supervising Engineer, Site Licensing B. Hampton, System Engineer D. Hollabaugh, Supervising Engineer, Technical Support G. Hughes, Supervising Engineer, independent Safety Engineering Group L. Kanuckel, Supervisor Civil Design Group K. Kuechenmeister, Superintendent, Design Engineering J. Laux, Manager, Quality Assurance J. McGraw, Superintendent, Technical Support Engineering C. Naslund, Manager, Nuclear Engineering A. Passwater, Manager, Licensing and Fuels C. Pilkington, Outage Supervisor G. Randolph, Vice President, Nuclear M. Reidmeyer, Engineer, independent Safety Engineering Group R. Rice, Design Engineer T. Sharkey, Supervising Engineer, NESM W. Witt, Superintendent, Systems Engineering NRC D. Passehl, Senior Resident inspector LIST OF INSPECTION PROCEDURES USED iP 37001 10 CFR 50.59 Safety Evaluation Program IP 37550 Engineering LIST OF ITEMS OPENED AND CLOSED Ooened 50-483/9705-01 VIO Failure to report the inoperability of main steam safety valves as required by 10 CFR 50.73 (Section E1.3.b.1).
 
50-483/9705-02 IFl Review the final analysis for the f ailed pressurizer safety valve in Refueling Outage 8 to assure that the accident analysis is still valid (Section E1.3.b.2).
 
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50-483/9705 03  IFl Review the licensee's efforts to determine the extent of Final Safety Analysis Report discrepancies and the adequacy of corrective actions (Section E2.1b).
 
50-483/9705-04  VIO Failure to report the 10 CFR 50.59 safety evaluations performed for temporary modifications as required by 10 CFR 50.59(b)(2) (Section E2.4.1.b.2).
 
50-483/9705-05  APV Failure to perform a 10 CFR 50.59 safety evaluation for changing the method of operation of the post-accident sampling system (Section E2.4.1.b.3(2))  )
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50-483/9705-06  APV Failure to perform a 10 CFR 50.59 safety evaluation for changing the setpoints on the refueling machine (Section E 2.4. 2.b.1 ).
50-483/9705-07  APV Failure to perform a 10 CFR 50.59 safety evaluation for the substitution of manual operator action for the automatic function for the diesel generator building supply fans. Based on the increase in probability of failure, this change was an unreviewed safety question as defined in 10 CFR 50.59 (Section E2.4.2.b.2).
 
LIST OF DOCUMENTS REVIEWED Plant Procedures Procedure  Revision Title APA-ZZ-00007  11 Quality Assurance Organization, Responsibility and Conduct of Operations JDP-ZZ 04100  8 Operating Experience Review Procedure APA-ZZ-00107  3 Review of Current Industry Operating Experience JDP-ZZ-04400  2 Callaway Plant Event Reduction Program JDP-22-01100  5 ISEG Tracking Log -
JDP-ZZ-02000  5 STA Personnel Qualification and Training JDP-ZZ-03C00  6 ISEG Engineer Control Room Watch JDP-ZZ-04000  2 Document Reviews l
JDP-ZZ-04200  4 Callaway Operating Experience input to Nuclear Network JDP-ZZ-04300  3 Review of Nuclear Safety Review Board Material
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APA-ZZ-00140  20 Safety, Environmental and Other Licensing Evaluations
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Procedure  Revision Title    )
 
APA-ZZ-00104  7 Technical Specification Interpretations and Notice of
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Enforcement Discretion  l l
 
APA-ZZ-304  11 Control of Callaway Equipment Lists  l APA-ZZ-500  27 Corrective Action Program APA-ZZ-604  16 Requests for Resolution  ,
 
APA-ZZ-605  7 Temporary System Modifications APA-ZZ-600  15 Design Change Control APA-ZZ-325  4 Initiating, Authorizing, and Removing Condition Reports EDP-ZZ-4023  13 Calculations EDP-ZZ-4055  2 Design Bases Control MSM-BB-QV001  16 Pressure Safety Valve Testing MSM-AB-OV001  10 Main Steam Safety Valve Set Pressure Test
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OPS-EG-V001 A  17 CCW Train a Section XI Valve Surveillance OPS-EG-V001 B  13 CCW Train B Section XI Valve Surveillance TDP-ZZ-0065  0 Training and Qualification of Engineer Support Personnel Plant . Modifications Modification Title FiFR 16981 A Approved Replacement Gages and Transmitters RFR 17402A Change Heat Exchanger End Cover Gasket Material MP 93-1055 Modification to Hangers EJ01-R502/134 and EJ02-R504/133 MP 93-1058 Move Close Torque Switch Bypass to Rotor 3 on EGHV62 MP 96-1003 Modification to Provide a Level Instrument on the Turbine Exhaust Line of Turbine Driver (KFCO2) Associated With the TDAFP MP 96-1014 Installs An isolation Valve In The B Train ESW To AFP Suction Line Temocrary Modifications Modification Title TM-960E010 This TM removes the SR power supply source from differential pressure gauges GKPDIS50028, 39,100 and 103 on SGK04A, SGK04B, SGK05A, and SKG05B, respectivel a
Approved By:
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g Arthu'r T. Howell, Director i Division of Reactor Safety DISTRIBUTION:
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Garry L. Randolph, Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, Missouri 65251 Nuclear Consulting, In Raines Drive Derwood, Maryland 20855 Gerald Charnoff, Es Thomas A Baxter, Es Shaw, Pittman, Potts & Trowbridge 2300 N. Street, Washington, D.C. 20037 H. D. Bono, Supervising Engineer Site Licensing Union Electric Company P.O. Box 620 Fulton, Missouri 65251 Manager - Electric Department Missouri Public Service Commission 301 W. High P.O. Box 360 Jefferson City, Missouri 65102
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l Suaaestion-Occurrence-Solution Reports    l l
l SOS Title 96-0926 EO discovered NB01 room warmer than NB02 room. He found the A/C unit would start and valve GKV0767 opened but would go immediately shu During the performance of OSP-EC V001 A, EC-HV-0011 was timed stroked closed, but not open as require Fuel and/or lube oil was discovered by the NRC in the 'B' diesel firepump sum During performance of OSP-SA-0017A, there were complaints that the resultant lineup caused excessive HVAC pressure in the fuel builoin Letdown HX outlet temp control valve (BGTV0130) experienced problems again at controlling letdown temperatur During restoration of some WPA on the heater drain pump, a tag was in place but the valve was out of the tagged position,  j 96-1775 Due to high vibration on the "B" RCP, the pump was secured and the
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reactor shutdown per TS 3.4. While attempting to isolate ESW on WPA 21514, it was discovered that the pointer for EFV0275 was 180 degrees out of positio Significant oil leak noted on the "A" MF ]
96-1981 Resolution of a level 4 violation 95-1952 A single active failure of the CCW train could result in loss of cooling to the CCP miniflow 95-0230 Suspect valve seat leakby l
95-1428- Unexpected levelincrease noted in B CCW surge tank 96-0355 Review Velan valves for their applicability to OE 7640 95-2140 Equipment required for remote shutdown is not being tested on a periodic basis 95-1593 CCW valve must be positioned open or throttled to prevent CCW low flow to RCP motor cooler 95 2297 CCW valves have not been verified to be in their correct position 95-2094 Only one CCW penetration has an automatic isolation valve 95-0787 Pin hole leak discovered in weld upstream of CCW valve 96-1380 Pin hole leak in vent pipe of CCW heat exchanger
 
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SOS Title 96-1795 CCW system temperature below Final Safety Analysis Report minimum 95-2065 Several valves credited for operating following a single failure have been designated as passive
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95-0860 A potential operability concern with EGD02A and the A train Diesel ]
Generator
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95-1792 Valves did not fully stroke per their indicators 96-1263 During testing a PSV exceeded its Technical Specification tolerance 96-1247 During testing a MSSV exceeded its Technical Specification tolerance 90-2908 During testing a PSV exceeded its Technical Specification tolerance l 95-0508 During testing 14 MSSVs exceeded their Technical Specification tolerance 90-1682 During testing 3 MSSVs exceeded their Technical Specification tolerance l 95-0013 Failed surveillance of the CCW surge tank level transmitter 95-2219 Cases were identified reviewing CCW flow verification tests that did not meet the testing requirements 95 2105 Resolve Callaway's definition of safe shutdown 95-2126 Discrepancies noted during review on auxiliary building flood calculations 95-1852 No CEL ECN was initiated when a hanger was rernoved 94-0775 During review it was found that some valves listed as active in the Final j Safety Analysis Report were listed as passive in the equipment list i i
94-1047 RFRs had been generated without correction of errors in the equipment list or Final Safety Analysis Report 94-0702 The valve nozzle and guide ring settings could not be verified 96-0877 A separation violation exist in the vendor terminal box for GKPDIS0103 97-0255 Work done on Diesel Generator A consisted of replacing the cam cover I gasket on the left bank with new gaskets made of Bluegard 3000 which leaked oil    j i
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Reauests for Resolution
!  RFR  Title -
176726  Establishing range of design flows for CCW components
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16449  Revise valve drawing and vendor manual to' include size of vent plug hole l  16458  Evaluate seal water heat exchanger capacity.
 
;  16528  Evaluate safety function of excess letdown heat exchanger 16444  Operability of CCW pumps with pump room coolers inoperable
;  16457  Update CEL' Q-list reason fields for various components 16448  Raise CCW surge tank low level alarm setpoint 15246  Evaluate differences between CEL and Final Safety Analysis Report Table 3.9(B)-16 17572  Update drawings for storage items 15489A  To provide positive indication of the status of the Turbine Driven Auxiliary, Feedwater Pump (TDAFP)
17206A  Modification request to rewire GKPDIS0028,39,100 and 103
          .
Calculation . Revision  Title    -
J EG-32  0  Calculation Determines Volume Contained in the CCW l Surge Tanks Versus Fluid Level  j EG-34  0  Upper Recommended Flowrate Determination for Components Using CCW Work Reauests  Title W1b3699  Change EG-LSHL-000224 setpoint -
G525907-181  Install and remove glove bag G561893  Generic WR to perform troubleshooting W163006  Replace lower wedge in valve internals W169645  Remove PSA snubber and replace W172312  Actuator air supply regulator is plugged W173733  Replace vent cap W174151  Adjust open limit switch on EGHV0016 W177243  Valve handwheel required to be locked in place
 
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Work Recuests Title W178013 Perform ISI VT-3 exam on EF02-C003 W178273 Retorque bolting due to chemical leakage W531533 Replace relays W531535 Replace relays W531564 Replace relays    j W531568 Replace relay due to outgassing concern W575857 Install DP gage and take readings G579598 Generic work request for cleaning boric acid leaks l
W174735 Drill hole in versa valve vent plug to 11/32 in, diameter 10 CFR 50.59 Screenina and Safety Evaluations For The Followina Documents:
Plant Title Modifications RMP 94-2005A Redesign of PASS System (screening)
CMP 95-1027A Rewire Cabinet Cooling Fans to Receive Safety Grade Power (screening)
CMP 95-1019 Change Cable Size for EJHV8701B (screening)
CMP 96-1008 Add Local Control Stations to S/G PORVs "B" & "C" (safety evaluation)
CMP 95-1004 Modify SSPS Power Supply (safety evaluation)
CMP 95-1007 Change RHR Miniflow Valves to Limit Close (safety evaluation)
CMP 95-2015 Remove Flow Switches from Control Logic (safety evaluation)
Temocrary Modifications q
_N_o ,. Title TM 95-E0006 Install Recorder to Monitor Battery Charger NK22 (screening)
TM 96-E0013 Add Manual Switch to Control Speed of Refueling Machine (screening)
TM 94-M005 Provide Lube Water for the Cirewater Pump (screening)
TM 95-M019 Install Temporary Lube Oil Fill Line for RCP-C (screening)
 
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TM 96-E0005 Jumpering of Temperature Switches for UHS Sump Heaters (screening)
TM 96-M004 Install Splash Shield on "B" CCP (screening)  i
 
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TM 95-E0006 Remove Failed Signal from COPS (safety evaluation)  !
TM 95-M002 Temporary Filter and Piping for BTRS Chill Water Loop (safety evaluation)
TM 96-M014 Installation of Blank Flanges on Cooler SGN01 A (safety evaluation)
New Procedures and Procedure Chances Number Title ETP-AE-STOO8 AEFV0042 Repair and Retest Procedure (safety evaluation)
ETP-RJ-ST001 Test of Rod Drop Software (safety evaluation)
ETP-BG-ST015 Letdown Heat Exchanger Flow Test (safety evaluation)
ETP-MB-ST001 Main Generator Excitation Stability Test (safety evaluation)
ETP-ZZ-ST019 Plant Radio Testing (cafety evaluation)
ETP-ZZ-ST006 Bank Reactivity Worth Measurement (Rod Swap) (safety evaluation)
OTN-NE-0001 A Standby Diesel Generator System - Train "A" (screening)
OTN-NE-00002 Standby Diesel Generator Auxiliary Systems (screening)
MSM-KJ-QT001 10 Year Emergency Diesel Generator Fuel Oil Storage Tank Cleaning (screening)
Reauests for Resolution RFR Title 16337 A & B Sediment in Diesel Generator Fuel Oil Storage Tank (safety evaluation)
17402 A Change HX End Cover Gasket Material (screening)
13573B Modify PORV Leakoff Line (screening)
13877 A Cavity Return Temperature Alarm Setpoint (safety evaluation)
16805 A Increase DP Capabilities of Valves (screening)
16981 A Approve Replacement Gauges and Transmitters (screening)
14464 B Pressurizer Safety Valve Drawing Changes (screening)
16311 B Evaluate Containment Cooler Motor Acceptability (screening)
 
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*e          j Final Safety Analysis Report Chanae Notices    l ff
  . _C N Title I
95-046 Correct Final Safety Analysis Report Descriptions of MStV Status on Loss-
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of-Electrical Power (screening)
,  94-011 Remove Final Safety Analysis Report Table 3.11(8)-3 from Final Safety Analysis Report - Put in CEL (screening)
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96-020 . Delete Note on Figure 6.2.4-1 in Final Safety Analysis Report for Type A CIV Test (safety evaluation)    ;
i-  Technical Specification Interoretations i  TSI Re TS/ Final Safety Analysis Report Section
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1 1 TS 3/4.5.1 i  4  Final Safety Analysis Report 16. O TS 3/4.3.1
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11 4 TS 1.2, TS 3/4.9 _
Final Safety Analysis Report 16.0-2 13 2 TS 3/4.1 TS 3/4.3.3, TS 3/4.4.6
)    Final Safety Analysis Report 16.3. TS 3/4.8.1.1 -
19 9 TS 3/4. TS 3/4.0 l  23 1 TS 3/4.7.6e, 3/4.9.13d & 3/4.7.7d 25 2 Final Safety Analysis Report 16. TS 3/4. TS 3.7.1.5 & 3.7. Final Safety Analysis Report 16. TS 3/4.9.2 l  44 1 TS 3/4.7.6, 3/4.7.7 & 3/4.9.13 46 6 TS 3/4.7.6, 3/4.7.7 & 3/4.9.13 49 6 TS 3/4.8.1.2, 3/4.4.1.4.2, 3/4.7.6 & 3/4. Final Safety Analysis Report 16.1.2.1 & 16.1. s      _ .  -  -
 
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e TSI Re TS/ Final Safety Analysis Report Section 50 14 TS 3/4. TS 3/4.6.1.4 & 3/4.6. TS 3/4. . TS 3/4.3. TS 3/4. TS 3/4.6. Final Safety Analysis Report 16.6. TS 3/4. . 4 TS 3/4 TS 3/4.3.1 & 3/4. TS 3/4. O TS TS 3/4.4. O TS 3/4. O TS 3/4.9.12 68 1 TS 3/ TS 5.1.3,' 6.8.4.e, 6.9.1.5, 6.12 & 6.14a Final Safety Analysis Report 16.1 ,
70 5 TS 3/4. TS Final Safety Analysis Report 16. TS 3/4.1, 3/4.1.1, 3/4.9 & 3/4. Final Safety Analysis Report 16,16.7.2.1 & 16. O TS 3/4.5.2 & 3/4. Final Safety Analysis Report 16.1.2.1,16.1.2.2,16.1.2.3 16.1. O TS 3/4.1. O TS 3/4. l
 
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0 t i  Nuclear Enaineerina Quality Assurance Audits
 
l  AP94-019 Quality Assurance Audit of Materials and Nuclear Engineering Technical
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Support (Materials Engineering)
$;  AP95-004 Quality Assurance Audit of Qualifications of Nuclear Division Personnel
 
i  AP96-001 Quality Assurance Audit of Qualifications of Nuclear Division Personnel i
i  AP96-002 Quality Assurance Audit of Design Control
 
AP96 011 Quality Assurance Audit of Operator and Engineering Support Personnel Training
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l  AP96-010 Quality Assurance Audit of Corrective Action
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Nuclear Enaineerino Quality Assurance Surveillances l  SP94-107 Fire Barrier Penetration Seal Qualifications
!  SP95-013 10 CFR 50.59 Safety Evaluation for Final Safety Analysis Report Change
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Notice 94-31      i i
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i  SP95-061 Plant Modification Configuration Control SP95-073 Review of EDP-ZZ-04023, Calculations
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SP95-101 Vendor Equipment Technical Information Program SP96-027 Technical Adequacy and Configuration Management of the Request for i  Resolution Program I
  .SP96-059 Cancellation of Modifications    !
SP96-103 Review of Component Cooling Water System Operating Parameters Miscellaneous Documents s
Technical Specifications Final Safety Analysis Report Operating Review Committee Meeting Minutes and Materials Review Log for the Operating Review Committee Ineeting minutes for the period of December 15,1995 through February 7,1997 SEGR 96-05-001, independent Safety Engineering Group 18 Month Summary, dated May 23,1996
 
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. SEGR 95-12-003, Independent Safety Engineering Group Review of NRC Region IV SALP
; and Violations, dated February 20,1996 i
! SEGR 96-10-003, Independent Safety Engineering Group Followup review of EF system
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failures, dated October 10,1996
) SEGR 9610-008, independent Safety Engineering Group 1995 and 1996 highlights, dated November 4,1996
)
: SEGR 96-11-007, independent Safety Engineering Group Review of SOS 96-1385, Gripper
' Damage, dated November 13,1996    l j independent Safety Engineering Group Responsibility List, dated Jar.uary 13,1997
. Reports to the industry via the INPO Nuclear Network submitted by Independent Safety
,
Engineering Group personnel, dated February 12,1997 l Centralized Action Tracking System general information report providing the status of j Independent Safety Engineering Group entered items
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Independent Safety Engineering Group 18 Month Summary, dated May 23,1996  I l
Suggestion Occurrence Solution Report listing for 1996 and 1997 occurrences identified by  !
Independent Safety Engineering Group personnel
        '
Independent Safety Engineering Group Action Tracking list for the past year
        >
Operating Experience Crew Briefing Sheets, dated November 4,1996, September 10, 1996, October 1,1996, October 8,1996, October 10,1996, and January 13,1997 Operating Experience Journal on Excerpts from NRC Violations and Reports, dated March 1996
        ^l Operating Experience Journal on Shutdown Operations, dated May 1996 Operating Experience Journal on Drain Down and Midloop Operations, dated May 1996 Operating Experience Journal on Radiation Protection, dated May 1996 Operating Experience Journal on Shutdown and Startup Operating Experience, dated  I June 1996      I l
l February 5,1997 Letter ULNRC-03530 to NRC, "NRC Enforcement Policy Revision"  l February 6,1997 Letter ULNRC-3531 to NRC "Callaway Plant Response to Request for  I Information Pursuant to 10 CFR 50.54(f) Regarding adequacy and Availability of Design Basis Information" Nuclear Energy Institute guidelines of NEl 96-05, " Guidelines for Assessing Programs for Maintaining the Licensing Basis," Revision 0, dated October 7,1996 I
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Union Electric C j Meeting August 15,1997 j l
D Ronald A. Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, Missouri 65102 Otto L. Maynard, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporatien P.O. Box 411 Burlington, Kansas 66839 Dan I. Bolef, President Kay Drey, Representative Board of Directors Coalition for the Environment 6267 Delmar Boulevard University City, Missouri 63130 Lee Fritz, Presiding Commissioner Callaway County Court House 10 East Fifth Street Fulton, Missouri 65151 Alan C. Passwater, Manager Licensing and Fuels Union Electric Company P.O. Box 66149 St. Louis, Missouri 63166-6149 J. V. Laux, Manager Quality Assurance Union Electric Company P.O. Box 620 Fulton, Missouri 65251
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Revision as of 11:51, 26 October 2020

Notification of Licensee Meeting W/Util on 970815 to Discuss Failure to Perform SE IAW 10CFR50.59 Requirements Identified in Insp Rept 50-483/97-05
ML20149G857
Person / Time
Site: Callaway Ameren icon.png
Issue date: 07/16/1997
From: Howell A
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
50-483-97-05, 50-483-97-5, EA-97-168, NUDOCS 9707240099
Download: ML20149G857 (4)
v • d • e


Text

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UNITED STATES

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. ht>* "%4  ;

y+1 *g NUCLEAR REGULATORY COMMISSION'

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q$ REGloN IV 0 '

811 RYAN PLAZA DRIVE, SUITE 400 I

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AR LINGTON, TEXAS 76011-8064 4'+4 * * * * * #

JUL I 61997 -

l NOTICE OF LICENSEE MEETING

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Name of Licensee: Union Electric Company i N vie of Facility: Callaway Plant E, .cket:

. 50-483 EA No.: 97-168 Date and Time August 15,1997 at 10:00 a.m. (CDT)

of Meeting:

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Location of Meeting: Region IV Office, Arlington, Texas j 4th Floor, Training Conference Room I Purpose of Meeting: Pre-decisional Enforcement Conference to Discuss the Failure to Perform Safety Evaluations in Accordance with 10 CFR Part 50.59 Requirements identified in Inspection Geport 50-483/97-

NRC Attendees: E. Merschoff, Regional Administrator A. Howell, Director, DRS K. Brockman, Deputy Director, DRP C. VanDenburgh, Chief, Engineering Branch, DRS W. Johnson, Chief, Projects Branch B, DRP T. Stetka, Senior Reactor inspector, Engineering Branch, DRS K Thomas, Project Manager G. Sanborn, Enforcement Officer Licensee Attendees: G. Randolph, Vice President and Chief Nuclear Officer C. Naslund, Manager, Nuclear Engineering R. Affolter, Manager, Callaway Plant A. Passwater, Manager, Licensing and Fuels J. Laux, Manager, Quality Assurance K. Kucchenmeister,' Superintendent, Design Engineering T. Sharkey, Supervising Engineer, Safety Related Mechanical Systems f

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9707240099 970716 PDR ADOCK 05000483 l 0 PDR , _ _ _ , _ .- _ _ _ - _ , .

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q Union Electric C Meeting August 15,1997

E-Mail To: FOR (EWM, ATH, KEB, CAV, WDJ, TFS, GFS)NRC Attendees PMNS Mtg Announcement Coordinator JKR E. L. Jordan, DEDO PAB H. L. Thompson, DEDR FJM F. J. Miraglia, A/D/NRR l ACT A. C. Thadani, ADT/NRR RPZ R. P. Zimmerman, ADP/NRR (TWA) Project Manager, NRR OEMAll J. Lieberman, D/OE ,

LJC L. J. Chandler, OGC I (DP) D. Passehl, Resident inspector CAH C. A. Hackney, RSLO j DMK1 D. M. Kunihiro, WCFO I GFS G. F. Sanborn, EO BWH B. Henderson, PAO MFH2 M. Hammond, PAO

, KEP Ken Perkins, Director, WCFO JAC1, CMS, CJG RA Secretaries LAT, DLF, CAM 3 DRP Division '

CLG, LMB DRS Division NLH, WSW DNMS Division LSO Linda Ousley cc: )

. DMB (IE 45)g

  • 'OEDO RIV Coordinator (17G21)

Receptionist ORA File RIV File DOCUMENT NAME:

To receive copy of document, Indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy RIV: SRI:EB l E C:EB l E D:DRP PAO gl RSLOLM{1 D:DRS l TFStetka/Imb" CAVanDenburgh" TPGwynrd BHendersonk CAHackney ATHowell lip 07/11/97 07/11/97 07/11/97 07/l(/97 07/% 97 07/4/97 OFFICIAL RECO 3D COPY

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Union Electric C I Meeting August 15,1997 I

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NOTE:

(1) This meeting is open to attendance by members of the general publi (2) N?.0 personnel, not listed above, that desire to attend this meeting should notify Mr. T. Stetka at (817) 860-8247 by COB on August 8,1997, l

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Approved By:

g Arthu'r T. Howell, Director i Division of Reactor Safety DISTRIBUTION:

Garry L. Randolph, Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, Missouri 65251 Nuclear Consulting, In Raines Drive Derwood, Maryland 20855 Gerald Charnoff, Es Thomas A Baxter, Es Shaw, Pittman, Potts & Trowbridge 2300 N. Street, Washington, D.C. 20037 H. D. Bono, Supervising Engineer Site Licensing Union Electric Company P.O. Box 620 Fulton, Missouri 65251 Manager - Electric Department Missouri Public Service Commission 301 W. High P.O. Box 360 Jefferson City, Missouri 65102

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Union Electric C j Meeting August 15,1997 j l

D Ronald A. Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, Missouri 65102 Otto L. Maynard, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporatien P.O. Box 411 Burlington, Kansas 66839 Dan I. Bolef, President Kay Drey, Representative Board of Directors Coalition for the Environment 6267 Delmar Boulevard University City, Missouri 63130 Lee Fritz, Presiding Commissioner Callaway County Court House 10 East Fifth Street Fulton, Missouri 65151 Alan C. Passwater, Manager Licensing and Fuels Union Electric Company P.O. Box 66149 St. Louis, Missouri 63166-6149 J. V. Laux, Manager Quality Assurance Union Electric Company P.O. Box 620 Fulton, Missouri 65251

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