IR 05000445/1997001
| ML20141K150 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/21/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20141K138 | List: |
| References | |
| 50-445-97-01, 50-445-97-1, 50-446-97-01, 50-446-97-1, NUDOCS 9705280345 | |
| Download: ML20141K150 (14) | |
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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.:
50-445 50-446 License Nos.:
NPF-87 NPF-89
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Report No.:
50-445/97-01
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50-446/97-01 Licensee:
TU Electric Facility:
Comanche Peak Steam Electric Station, Units 1 and 2 Location:
January 27-31, and April 28 through May 2,1997
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Inspectors:
W. Wagner, Reactor inspector, Engineering inspector j
D. Passehl, Senior Resident inspector, Callaway
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(January 27-31, 1997)
T. Polich, Project Manager
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(April 28 through May 2,1997)
Approved By:
C. VanDenburgh, Chief, Engineering Branch Division of Reactor Safety ATTACHMENT:
Supplemental Information i
9705280345 970521 PDR ADOCK 05000445 G
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TABLE OF CONTENTS EX ECUTIVE SU M M ARY............................................. iii R e p o rt D e t a il s.................................................... 1
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111. Engineering
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E1 Engineering Self-assessment................................ 1
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l E1.1 Introduction
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E1.2 Self-Assessment Team Qualifications, Objectivity, and l
I nde pe nde nc e..................................... 1 l
E1.3 Scope and Depth of Self-Assessment
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l E1.4 Significant Self-Assessrnent Team Conclusions
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E1.5 Independent NRC laspection........................... 5 l
V. M a n ag e m e nt M e e ting s............................................ 9 l
X1 Exit Meeting Summary.................................... 9 i
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ATTACHMENT: Supplemental Information
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EXECUTIVE SUiviMARY
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Comanche Peak Steam Electric Station, Units 1 and 2 NRC Inspection Report 50-445/97-01; 50-446/97-01 This inspection evaluated the licensee's self-assessment effort that was performed as an alternative to an NRC engineering team inspection.
Enaineerina The NRC inspectors concluded that the team was composed of well-qualified
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individuals who were sufficiently independent and objective. The team was capable of performing good assessments (Section E1.2).
j The NRC inspectors concluded that the scope and depth of the engineering
self-assessment was sufficient to satisfy the requirements of NRC Inspection Procedure 37550 (Section E1.3).
The NRC inspectors concluded that the findings identified by the self-assessment
team were generally representative of engineering's strengths and weaknesses.
The inspectors also concluded that these findings were consistent with those j
observed by the inspectors (Section E1.4).
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The inspectors concluded that the licensee's self-assessment team identified
examples of weaknesses in design control and procedure compliance. Two noncited violations were identified for design control discrepancies and for procedure compliance violations (Section E1.5).
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Report Details Ill. Enaineerina E1 Engineering Self-assessment E1.1 Introduction An NRC Region IV team inspection was planned to accomplish the core inspection program requirements of NRC Inspection Procedure 37550, " Engineering," in January 1997, at the Comanche Peak Steam Electric Station. In a letter dated November 13,1996, Texas Utilities Electric, the licensee, proposed to perform a self-assessment of Comanche Peak Steam Electric Station engineering activities and programs in lieu of the planned NRC engineering team inspection. The licensee included its engineering self-assessment plan as an attachment to the letter. The
plan provided the details regarding the schedule, objectives, scope, methodology,.
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and team qualifications for the self-assessment.
In a letter dated January 29,1997, the NRC informed the licensee that their request was reviewed in accordance with the guidelines contained in NRC Inspection
Procedure 40501, " Licensee Self-Assessments Related to Team inspections," and i
accepted the proposed timing, scope of effort, and the credentials and experience of the assessment team. The option of permitting licensees to conduct a self-assessment in lieu of an NRC inspection was an NRC program aimed at minimizing regulatory impact and utilizing NRC resources more efficiently. Consequently, the
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NRC informed the licensee that it planned to reduce the level of inspection effort planned for inspection Procedure 37550, " Engineering," at the Comanche Peak Steam Electric Station. The NRC further informed the licensee that it planned to perform an inspection of the self-assessment effort in accordance with inspection Procedure 40501.
- The licensee performed an engineering self-assessment at the Comanche Peak Steam Electric Station on January 27 through February 7,1997. The results of the self-assessment were contained in a self-assessment report that was submitted to the NRC on March 20,1997. The NRC inspectors monitored the performance of the engineering self assessment on January 27-31,1997, reviewed the self-assessment results, and independently inspected the areas assessed by the licenscu on April 28 through May 2,1997. The purpose of the NRC reviews and
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inspections was to determine the quality and results of the licensee self-assessment.
E1.2 Self-Assessment Team Oualifications. Objectivity. and Independence
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insoection Scone (40501)
The purpose of this inspection was to determine the qualifications, objectivity, and independence of the licensee's self-assessment team in accordance with NRC Inspection Procedure 40501, " Licensee Self-Assessments Related to Team
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Inspections." The inspectora determined the qualifications, objectivity, and independence of the licensee's engineering self-assessment team and individual team members by performing the following:
The NRC inspectors reviewed the licensee's engineering self assessment plan
prior to, and during, the onsite inspection.
The NRC inspectors interviewed selected team members.
- The NRC inspectors accompanied and observed selected team members
during their assessment activities. At the time of the onsite inspection, the ongoing team assessment activities were predominantly interviews of l
licensee personnel and reviews of engineering work products.
The NRC inspectors observed daily team meetings.
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Observations and Findinas l
The NRC inspect ~ s found that the engineering self-assessment team consisted of eleven Texas Ut, :ies Electric engineering personal, and one consultant. The inspectors deterrained that the team was composed of diverse and well-qualified members who were sufficiently objective and independent to allow meaningful j
assessments. Specifically, the inspectors noted the following:
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t All engineering team members were degreed engineers. The major
engineering disciplines of mechanical, electrical, and nuclear engineering l
were represented on the team.
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All team members had ten or more years of industry experience in des'ign,
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systems, or plant support engineering functions.
l Team member assessment activities observed by the NRC inspectors were
independent and self-critical. Team meetings observed by the inspectors included critical discussions of team member observations.
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Conclusion The NRC inspectors concluded that the team was composed of well-qualified
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t individuals who were sufficiently independent and objective. The team was capable l
of performing good assessments.
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E1.3 Scoce and Deoth of Self-Assessment i
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InsDection ScoDe (40501)
The NRC inspectors reviewed the scope and depth of the licensee's self-assessment j
to determine if it was equivalent to those requirements specified in NRC Inspection
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Procedures 37550, " Engineering." The inspectors reviewed the self-assessment
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plan, observed in-process assessment activities, interviewed licensee personnel, and l
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reviewed the licensee's self-assessment report.
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Observations and Findinas Ennineerina Self-Assessment The NRC inspectors determined that the licensee's engineering self-assessment
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consisted of both a horizontal and a vertical review. The licensee chose to review the station service water system, emergency diesel generator system, instrument air
system, and the switchyard system for the vertical review of engineering activities
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for those systems. The adequacy of engineering activities regarding design and configuration control, surveillance and testing, and identification and resolution of problems associated with those systems were reviewed.
The licensee chose seven engineering subject areas for the horizontal review of engineering activities. Recent operations-related issues and maintenance-related issues that required engineering assistance, system health and performance, routine reactor engineering testing, temporary modification and maintenance alteration processes, inservice testing, followup to corrective actions, and engineering administration were the subject areas reviewed.
The team prepared and completed a matrix of team assessment activities as compared to the inspection requirement attributes of NRC Inspection Procedure 37550.
The inspectors found that the licensee had addressed most portions of inspection Procedure 37550. The inspectors noted that the licensee had not reviewed any modifications required by paragraph 2.08 of NRC Inspection Procedure 37550, because modifications of the type had not been performed within the last 2 years.
This inspection requirement pertained to design changes and modifications installed as part of NRC regulations 10 CFR 50.62,10 CFR 50.63, and Supplement 1 to NUREG-0737 for Regulatory Guide 1.97 instruments and the safety parameter display system.
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Conclusion The NRC inspectors concluded that the scope and depth of the engineering self-assessment was sufficient to satisfy the requirements of NRC Inspection Procedure 37550.
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E1.4 Sionificant Self-Assessment Team Conclusions a.
Inspection Scope (40501_J The NRC inspectors reviewed the licensee's self-assessment report. The NRC inspectors discussed the results of the self-assessment with licensee managers, the team leader, and team members. The inspectors summarized the licensee's conclusions.
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Observations and Findinas Enaineerino Self-Assessment The assessment team noted the following strengths in the engineering areas reviewed:
Knowledge and experience of the system engineers.
- Working relationships among engineering, transmission, and system
protection groups.
s The system health program was capturing key elements of system
performance, identifying performance enhancement needs, and prioritizing those needs.
The commitment tracking system as an identifier of regulatory obligations,
and partnering agreements and general services agreements which provide ready access to technical resources.
The assessment team noted the following weaknesses in the engineering areas reviewed:
Management of switchyard configuration needed improvement.
- Commitment tracking utility was decreasing because of inconsistencies in
d ttabase maintenance.
Reorganization may create a void in discipline-type design basis knowledge,
and such knowledge and system-specific knowledge may decline over time
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as resources shrink.
Increased attention was warranted in evaluating effectiveness of corrective
actions and the approaches taken to documenting deficient and nonconforming conditions.
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Additional controls or guidance on the maintenance alteration program may
be warranted to ensure that changes are not inadvertently made which require a licensing basis or a design basis change or a 10 CFR 50.59 and/or J
10 CFR 50.54 evaluation.
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Conclusions The NRC inspectors concluded that the findings identified by the self-assessment team were generally representative of engineering's strengths and weaknesses.
The inspectors also concluded that these findings were consistent with those observed by the inspectors.
i E1.5 Independent NRC Inspection a.
Insnection Scope (40501)
The NRC inspectors observed licensee performance of the self-assessment and reviewed the detailed self-assessment plan implementation matrix, the self-
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assessment team's field notes, the team's final report, and 17 associated ONE-j forms to determine the status of the corrective actions for the identified conditions and develop an understanding of the basis for the self-assessment team's conclusions. The NRC inspectors also walked down accessible portions of the station service water intake structure and the Unit 2 emergency diesel generator to
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confirm the as-built configuration of the systems. The NRC inspectors interviewed self-assessment team members and other cognizant licensee personnel. The inspectors also attended self-assessment team meetings.
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Observations and Findinas The NRC inspectors generally agreed with the conclusions of the licensee's self-assessment team. The inspectors considered the self-assessment team to be effective in conducting the assessment, in that, in-depth and critical reviews were performed.
The NRC inspectors noted that the licensee's self-assessment resulted in 17 ONE-forms. The NRC inspectors determined that two of the engineering i
assessment ONE-forms indicated a weakness in design control and that five of the engineering assessment ONE-forms indicated a weakness in complying with
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applicable procedures. Each of these two weaknesses are discussed below.
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b.1 Desian Control Incorrect Assumption Used in Design Document for Station Service System -
(Licensee Self-Assessment ONE-Form 97-094).
The licensee team identified that calculation 16345-ME-(B)-088 used an incorrect assumption about the screen wash pump supply line; that is, the calculation provided inadequate justification to support a code class break across normally open valves. 'The licensee determined that this discrepancy would not effect station service water system operability. An evaluation to determine the root cause of the identified condition and the corrective actions for the discrepancy was scheduled to be completed by June 3, 1997.
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No Calculation to Support Final Safety Analysis Report of Sufficient Jacket
Water System Makeup for Emergency Diesel Generator System -(Licensee Seil-Assessment ONE-Form 97-142).
The licensee team identified that there were no calculations supporting a Final Safety Analysis Report statement about jacket water system makeup availability and alarm levels. Final Safety Analysis Report Section 9.5.5.2 states, in part, "that there is potentially 310 gallons of water available to replace a leakage up to 1.5 gallons per hour for 7 days of continuous operation." Engineering performed a calculation which demonstrated a sufficient volume of water consistent with the Final Safety Analysis Report description. Based on this engineering calculation, the licensee concluded that operability was not affected. Licensee actions were scheduled to be completed by May 12,1997.
The two design control discrepancies discussed above were identified as a violation of 10 CFR 50, Appendix B, Criterion lil, " Design Control." This licensee identified violation is being treated as a noncited violation (50-445,446/9701-01), consistent with Section Vll.B.1 of the NRC Enforcement Policy.
b.2 Procedure Comoliance The inspectors determined that the self-assessment team identified the following procedure compliance violations:
No Procedure Provisions to Apprise Operators of Approaching Minimum Safe
Shutdown impoundment Temperature of 40 F (97-106)-(Licensee Self-Assessment ONE-Form 97-106).
The licensee team noted that the design basis established for the station service water system provided a low temperature limitation of 40 F. The team also noted that this limit did not appear in the Final Safety Analysis Report and did not appear to be captured in any implementing documents.
No alarms or procedures were found that would alert operators to a low
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temperature condition or direct any compensatory actions. The team determined that this limitation translated to a minimum component cooling water temperature which could have impacted lube oil cooler condensation.
The licensee concluded that the station service water system operability was not affected. This conclusion was based on the licensee assessment that the current station service water intake temperature was approximately 62 F and that the intake water temperature does not go below 50*F in the winter.
The licensee approved Procedure ABN-912, " Cold weather Preparations / Heat Tracing and Freeze Protection System Malfunction," Revision 4 on March 3, 1997. Attachment 1 to that revision added a bullet to Step 7 "to initiate shiftly monitoring of SSI temperaNre to ensure intake temperature remained
>40 F." Licensee actions on this item were completed on March 4,1997.
Unit Differences for Woodward Governor Modification Not identified in
Design Basis Documents - (Licensee Self-Assessment ONE-Form 97-110).
The licensee team determined that Procedure ECE 5.01-01, " Design Basis Documents," Revision 2, required that unit differences be identified in the design basis document. The team identified that Design Change Notice 8251 for the Woodward governors did not accurately reflect the (
fact that the electronic governor was installed on only one emergency diesel generator. The team concluded that Design Change Notice 8251 did not meet the requirements of Procedure ECE 5.01-01. The licensee determined that this condition had no impact on operability. Licensee action on this item was completed on May 1,1997.
Design Change Notice Not issued to Document Control Prior to Clearance
Release - (Ucensee Self-Assessment ONE-Form 97-130).
The IUensee team determined that Procedure ECE 5.01-07, " Maintenance Alteration Evaluation," Revision 2, required that final engineering documentation to be completed prior to work order closure / clearance release, whichever came first. The team identified that Design Change Notice 10291 was not issued to document control prior to clearance release on September 27,1996. The licensee determined that the installed hardware j
conformed to the design documentation, and concluded that there was no impact on operability. Licensee action on this item was completed by February 19,1997.
Emergency Diesel Generator System Design Basis Document Discrepancies -
(Licensee Self-Assessment ONE-Form 97-141).
The licensee team identified that the Jacket Water system Design Basis Document DBD-MF-0011, " Tornado Venting Analysis," was found to not identify four flowpath boundary active valves which were tested under the Inservice Testing Program. Further, Procedure SOP-609A, " Diesel G.9nerator System," Revision 10, provided a limitation on operability (i.e, the function and reset of slow start switches on Train A) that
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e was not provided in Design Basis Document DBD-ME-0011. The licensee
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concluded that these conditions did not impact operability. Licensee actions on this item were scheduled for completion by June 12,1997, Diesel Generator System Procedure inconsistencies - (Licensee Self-
Assessment ONE-Form 97-124).
The licensee team identified that Procedure SOP-609A was inconsistent with the Calculation 1.EB-302A and Design Basis Document DBD-ME-302A,
" Diesel Generator Area Ventilation System," regarding the limiting temperature for when diesel generator room exhnust fans were required to be operating. Further, the calculation did not address the availability requirements for two or three fan configurations. Another issue was i
identified by the team regarding whether unchecked and unserified calculations may be used directly in procedures. The team found that Procedure SOP-609A identified the maximum outdoor temperature to be 75 F for one fan availability, the design calculation indicated the maximum outdoor temperature as 76.7 F, and Design Basis Document DBD-ME-302A identified this limitation as 70 F referencing Calculation 1.EB-302A as the source.
To ensure system operability, the licensee determined that all diesel generator vent fan were operable on both units. The plant operators were alerted to this condition to ensure that all diesel generator vent fans were in standby if the outside air temperature was equal to or greater than 70 F.
Therefore, the licensee concluded that operability was not affected since all the documents referenced on this ONE-Form indicated that at less than 70 F only one vent fan was adespate to maintain the diesel generator rooms adequately cooled. Licensee actions on this item was scheduled for completion by June 30, ;997.
The NRC inspectors determined thet the procedure compliance violations were being corrected by the licensee within a reasonable time. These findings are being treated as a noncited violation consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-445,446/9701-02).
The inspectors reviewed portions of Nuclear Overview Department Evaluation Reports NOE-EVAL-96-200, " Corrective Action," and NOE-EVAL-97-00030,
" Maintenance Alterations." These two reports were ref,erenced in the self-assessment without any conclusions from the reports. The inspectors determined that both nuclear overview department evaluations were previously planned and independent of the self-assessment team. The corrective action evaluation started the week before the self-assessment and ended during the self-assessment. The maintenance alteration evaluation started the week after the self-assessrnent team
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finished. The licensee determined that further evaluation by the self-assessment team would have been a duplication of effort. The inspectors noted that the nuclear overview department evaluation reports supported the weaknesses identified by the j
self-assessment team in the areas of corrective actions and maintenance alterations.
The inspectors independent inspection identified no further violations.
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Conclusion i
The inspectors concluded that the licensee's self-assessment team identified j
examples of weaknesses in design control and procedure compliance. Two i
noncited violations were identified for design centrol derepancies and for
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j procedure compliance violations.
V. Manaaement Meetinas
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i X1 Exit Meetina Summary The 'NRC inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on May 2,1997. The licensee acknowledged the findings presented. The team asked the licensee whether any
j materials examined during the inspection were proprietary. No proprietary information was identified.
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d ATTACHMENT
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SUPPLEMENTAL INFORMATION
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I PARTIAL LIST OF PERSONS CONTACTED j
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LICENSEE
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1. Ahmad, Consulting Engineer, Electrical J. Amin, System Engineer Lead J. Barker, Manager, Nuclear Overview j
T. Bennett, Joint Engineering Team B. Bird, Manager, Nuclear Planning
M. Blavins, Plant Manager i
D. Buschbaum, Manager, Technical Compliance R. Calder, Executive Assistant
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R. Cockrel, System Engineer
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J. Curtis, Manager, Radiation Protection T. Evans, Supervisor, System Engineering j
J. Finneran, Modification Team
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R. Flores, Manager, System Engineering I
D. Goodwin, Manager, Operations Support W. Gufdemond, Manager, Shift Operations T. Hope, Manager, Regulatory Compliance J. Kelley, Vice President, Engineering and Support
M. Killgore, Manager, Reactor Engineering F. Madden, Manager, Technical Support i
I G. Merka, Senior Nuclear Specialist i
C. Montgomery, Modification Team
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J. Pettit, System Engineer
M. Quick, System Engineer j
L. Sailesh, Engineering Supervisor
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i L. Slaughter, Senior Engineer C. L. Terry, Group Vice President, Nuclear Production
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l R. Walker, Manager, Regulatory Affatrs B. Wells, Nuclear Overview G. Yezefski, System Engineer CONTRACTOR J. Polizzi, Westinghouse Site Engineering Manager i
NRC V. Ordaz, Resident inspector j
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INSPECTION PROCEDURES USED IP 37550 Engineering IP 40501 Licensee Self-Assessments Related to Team Inspections ITEMS OPENED AND CLOSED Ooened 50-445,446/9701-01 NCV Two Licensee Identified Design Control Discrepancies, E1.5.b.(1)
50-445,446/9701-02 NCV Five Licensee Identified Failure to Follow Procedure i
Requirements, E1.5.b.(2)
Closed 50-445,446/9701-01 NCV Two Licensee identified Design Control Discrepancies, E1.5.b.(1)
50-445,446/9701-02 NCV Five Licensee Identified Failure to Follow Procedure Requirements, E1.5.b.(2)
i LIST OF DOCUMENTS REVIEWED Licensee Procedures ECE 5.01, " Maintenance Alteration Evaluations," Revision 2 STA-421, " Operations Notification and Evaluation (ONE) Form," Revision 5 l
STA-602, " Temporary Modifications," Revision 11 STA-707,, "10 CFR 50.59 Reviews," Revision 13
STA-716, " Modification Process," Revision 14 l
ONE-Forms
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l 97-56,97-80, 97-87, 97-92, 97-94, 97-98,97-106, 97-109,97-110, 97-118, l
97-120,97-124, 97-128,97-130, 97-131,97-141, and 97-142.
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