IR 05000445/1989002
| ML20235U997 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/28/1989 |
| From: | Bitter S, Burris S, Joel Wiebe Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20235U992 | List: |
| References | |
| TASK-1.A.1.1, TASK-1.A.1.2, TASK-1.A.1.3, TASK-1.A.2.1, TASK-1.C.2, TASK-1.C.3, TASK-1.C.4, TASK-1.C.5, TASK-1.D.1, TASK-TM 50-445-89-02, 50-445-89-2, 50-446-89-02, 50-446-89-2, IEB-75-04, IEB-75-4, IEB-79-05, IEB-79-06, IEB-79-5, IEB-79-6, NUDOCS 8903090472 | |
| Download: ML20235U997 (21) | |
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6 II. S./ NUCLEAR REGULATORY COMMISSION
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, OFFICE OF NUCLEAR' REACTOR REGULATION
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l NRC Inspection Report: 50-445/89-02 Permits: CPPR-126 !
50-446/89-02 CPPR-127 ,
i Dockets: 50-445 Category: A2 50-446 :
Construction Permit Expiration Date: i Unit 1: August 1, 1991 Unit 2: August 1, 1992 Applicant: TU Electric Skyway Tower 400 North Olive Street Lock Box 81 ,
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Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station.(CPSES),
Units 1 and 2 ,
Inspection At: Comanche Peak Site, Glen Rose, Texas Inspection Conducted: January 11 through February 7, 1989
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Inspector: . 2 87 D. Bitter, Resident Inspector, Date ;
R Operations
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Inspector: I di P. Burris, Senior Resident Inspector, ' Ddte Operations j l
Reviewed by: M Ikf) O MI f. S. Wiebe, Lead Project Inspector /Datb 8903090472 890301 i"i 4 PDR ADOCK 05000445 I G PNV '
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Inspection Summary Inspection Conducted: January 11 through February 7, 1989 (Report 50-445/89-02; 50-446/89-02)
Areas Inspected: Unannounced resident safety inspection of the applicant's actions on previous inspection findings, follow-up on violations / deviations, preoperational retesting program, TMI action items (Safety Issue Management System Items I.D.1, open; I.A.1.1.1, open; I.A.1.1.3, open; I.A.1.2, open: I.A.1.3.2.A, open; I.A.2.1.4, open; I.C.2, open; I..C.3, closed; I.C.4, open; I.C.5, open),
significant meetings, piant tours, technical specification review, follow-up of NRC bulletins, follow-up of 10 CFR 50.55(e) reports, and maintenance program implementatio Results: Within the areas inspected, no significant weaknesses were identified. The applicant's maintenance training facilities were identified as a strength. During the inspection, no significant safety matters or deviations were identified. One violation was identified (paragraph 4) in the area of preoperational testing program activities; however, in accordance with the NRC revised enforcement policy effective October 13, 1988, no Notice of Violation was issued. The violation occurred when a shift test engineer failed to document that a test prerequisite had been complete .
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DETAILS i Persons contacted i
- R._W. Ackley, Jr., Director, CECO .
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- J. W. Beck, Vice President, Nuclear Engineering, TU Electric
- H. D. Bruner,. Senior Vice President, TU Electric
- W. J. Cahill, Executive Vice President, Nuclear, TU Electric ' '
- W. G. Counsil, Vice Chairman, Nuclear, TU Electric
- W. G. Guldemond, Manager of Site Licensing, TU Electric
- L. Heatherly, Licensing Compliance Engineer, '
TU Electric
- J. C. Hicks, Licensing Compliance Manager, TU Electric
- J. J. Kelley, Manager, Plant Operations, TU Electric
- 0. W. Lowe, Director of Engineering, TU Electric
- J. W. Muffett, Manager of Engineering, TU Electric ,
- E. F. Ottney, Program Manager, CASE i
- D.'M. Reynerson, Director of Construction, TU Electric
- A. B. Scott, Vice President, Nuclear Operations, TU Electric 4
~*C. E. Scott, Manager, Startup, TU Electric The NRC inspector also interviewed other applicant employees ;
during this inspection perio * Denotes personnel present at the February 7, 1989, exit intervie . Applicant Action on Previous Inspection Findings (92701)
(Closed) Unresolved Item (445/8601-U-02): Battery room j
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thermostats are not explosion proof, as recommended by Section 14 of ANSI C2, 1977, The National Electrical Safety Code. The applicant reviewed the unit heaters in Class 1E battery rooms and found that these components were not powered from redundant Class 1E power supplies and were not l seismically qualified. The applicant has determined that this deficiency is limited to Class 1E battery rooms and is reportable under provision of 10 CFR 50.55(e). In accordance with procedural requirements, the applicant issued SDAR (Significant Deficiency Analysis Report) CP-88-08 and generated Design Change Authorizations (DCAs) 66141 and 5992 These DCAs will change existing Class lE battery room thermostats for ones which are seismically qualified and safety class as required by design. In addition, the nonsafety-related nonexplosion proof electric unit heaters are being replaced with safety-related explosion proof electrical unit heaters in the Class-1E battery room The inspectors are closing this item based on the applicant's corrective actions and will use the SDAR (CP-88-08) to track the applicant's completion of the modifications. The inspectors consider this item close _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _ - _ _ _ _ . . - - _
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O 4 Follow-up on Violations / Deviations (92702)
(Closed) Violation (445/8431-V-05): Changes to Procedures and Responsibilities of the Station Operations Review Committee (SORC). This issue originally dealt with the applicant's propriety in changing and the ultimate review / approval of
" Acceptance Criteria" of Procedure ELM-302, Revision 0,
"480 Volt Air Circuit Breaker Inspection."
The inspectors reviewed the applicant's responses to various NRC inspection reports and found that this issue involved the review and approvai of requirements and responsibilities of the SORC, as defined in technical specification The original referenced documents have been revised or conservatively interpreted by the applicant to ensure that work instructions receive the same level of review that a procedure would receive. The procedures and their changes or interpretations were reviewed as follows:
. Maintenance Department Administrative (MDA)-201 no longer contains criteria for deficiency procedures or instructions. This procedure required that all maintenance department procedures be reviewed and approved in accordance with Station Administrative Procedure (STA)-202. Changes to the maintenance procedure are to be accomplished in accordance with STA-20 . STA-202 specifies that all work instructions are to be treated as procedures during the review and approval process. This procedure also contains the procedure review and approval matrix which specifies those procedures requiring SORC revie . The CPSES Technical Specifications, Proof and Review Copy, Section 6.5.1.6 describes the SORC responsibilities for review of procedure changes and safety evaluation TU Electric management assured the inspectors that any intent or criteria change to a procedure or instruction would be reviewed by the SORC prior to implementing the procedure for us Based on this review of the program and the current copy of technical specifications, the inspectors consider this item close . Preoperational Retest Program Activities (39301, 70300, 70301, 70302, 70311, 70312, 70333)
NRC inspections of the applicant's preoperational retest and operational preparedness phase activities were performed
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' through direct observation, personnel-interviews and reviewLof preoperational test activities to verify that:
. . Systems and components important.to the safetyfof the plant were being fully tested to demonstrate their operability and design' requirement .- kil management and administrative control procedures, including QA requirements, which were required had been implemented, followed, and' documente To accomplish these goals, the NRC inspectors reviewed available. test procedures, witnessed selected ongoing 1 test activities, and. reviewed completed test procedure results.<
The inspectors-used the following criteria to perform the'~
pretest review to ensure that:
.. Test objectives met the referenced Regulatory Guide and FSAR Section 14 commitment . Acceptance criteria were identified and clearly define . Prerequisite conditions were established, adequately defined, and easily understoo . ' Content, format, and requirements were incorporated in the final approved procedures as required by administrative procedur . When test equipment was used, the appropriate custody control and required calibration data were'specifie . procedures were clearly written and appeared to be able to be easily followe Test witnessing of the identified systems was performed to ensure that all testing was performed in accordance with approved procedures and to verify the adequacy of test program records including preliminary evaluation of test results. The NRC inspectors accomplished this by ensuring that:
. The latest revision of the test procedure was in use by test personne ;
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. All crew manning requirements were me . All test prerequisites were me . Proper plant systems were in servic ;
. Test equipment required by the procedure was calibrated and in service, if applicabl _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _
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. Testing was performed in accordance with an approved procedur . Criteria for interruption of testing and continuation of testing were followe . All data was collected by the proper personne . All temporary modifications, such as, jumpers, strainers,
\ spool pieces, or blank flanges were installed and tracked per established administrative control The inspectors reviewed and/or witnessed the following test as discussed below:
. Test Procedure 1CP-AT-03-02, Revision 1, "345 KV Station g Service Transformer (1ST)." The stated purpose of this test is to demonstrate that the station service transformer 1ST can supply the Unit i normal busses and the station common bus, and to demonstrate that the protective relaying system can isolate transformer 1ST in the event of a 345 KV transformer tri In particular, the NRC inspectors witnessed the performance of Section 7.5 of this test. Section 7.5 was performed to demonstrate the ability to manually transfer the non-class 1E loads from transformer 1ST to the unit auxiliary transformer, 1UT, and to demonstrate the ability to automatically transfer the loads from 1UT to IST using both the fast and slow transfer mode I Following the pretest briefing conducted by the shift l supervisor, the inspectors witnessed the completion of the specified test prerequisites and the tripping of the Unit 1 345 KV switchyard generator breakers. Subsequent to tripping these breakers, the inspectors noted a deficiency. Specifically, coincident with the loss of normal power, both radio repeaters were lost. Thus, there was no remote communications capability. This occurrence resulted in the temporary suspension of test activitie This deficiency will be reviewed during a future inspection. Any findings will be documented in future inspection report . Test Procedure 1CP-PT-04-01, Revision 0, " Station Service Water," was performed by Startup personnel and reviewed by the NRC inspection staff during this inspection perio The inspectors followed up on an open issue discussed briefly during the previous inspection report which dealt with Startup personnel failing to properly document prerequisite conditions. During review of the
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official test on January 8, 1989, the inspectors found that the startup test engineer.(STE) was performing Section 7.13. This section was intended to verify proper operation of the interlocks and controls for the screen wash pump (CPX-SWAPTS-02), the screen wash header control valve (X-CU-4289), and the traveling screen (CPX-SWTSTS-02). The first step in this section, Step 7.13.1, required the STE to verify that the applicable prerequisites of Section 6.0 had been complete Section 6.0, step 6.3, required the STE to align the screen wash portion of the system for operation in accordance with SOP-501A for test sections 7.5 and 7.1 Although Step 7.13.1 had been signed as complete i
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on January 8, 1989, step 6.3 of the prerequisite conditions was not signed when the inspector reviewed the'
tes The inspectors questioned the STE as to whether or not this particular prerequisite condition had been met. The STE assured the inspectors that the prerequisite condition had been performed, but due to an oversight, he had failed'to' sign the ste The inspectors informed the applicant that failure to properly document prerequisite conditions was contrary co their test Department Administrative Manual Procedure TEA-303, Revision 0, Section 6.2.6.11, Step 6.2.6.11.1, which states "The test engineer shall ensure that all prerequisites have been completed which are applicable to the test section which is to be performed."
Although this is a clear violation of administrative procedures, the inspectors have determined that a Notice of Violation will not be issue This is in accordance with the revised statement of policy effective October 13, 1988, based on the following:
. It appears that this is an isolated case because no other instances have been identified to dat . The applicant took immediate corrective actions to rectify the identified problem and prevent similar occurrence . The violation was not willful in natur The following inspection report item number will be considered open and closed by the above documentation (50-445/8902-V-01).
. Test Procedure 1CP-AT-34-03, Revision 0, " Steam Dump Preoperational Test." The inspectors reviewed the procedure and witnessed selected portions of testing
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activities during this inspection period. During review of testing activities, the inspector found that the test was being conducted such that stated test requirements and criteria were me . Test Procedure 1CP-PT-49-01, " Charging and Letdown,"
was reviewed and portions of this test were witnesse In general, it was determined that the test-was progressing satisfactorily. However, the inspector questioned the STE and startup supervision as to the method of verifying reference documents prior to test commencement. Discussions between startup supervision and NRC personnel are currently ongoing as to whether or not the applicant's current program is acceptable. This issue will be followed up on during a future inspection perio . Test Procedure 1CP-PT-57-01, Revision 0, " Safety Injection Pump Performance," was reviewed and the inspectors witnessed Section 7.7, Train A and B Hot Leg Flow Balance. The inspectors questioned when and how the applicant would secure the flow balance valves once the correct flows were obtained and verified. The test procedure was not explicit as to when these valves would be secured in the cor.cect settin During discussions with the STE, the inspectors were informed that the valves would be secured by welding as soon as practical after flow balancing. However, later review' revealed that this l was not performed until the next day. The l inspectors discussed with the applicant the necessity for maintaining control over preset flow settings once the flow balancing portion of preoperational testing was complete. The STE !
assured the inspectors that future items requiring l absolute control would be handled in a more expeditious manner. The inspectors will follow-up '
on this item during future inspection period . Test Procedure 1CP-PT-57-05, " Safety Injection Accumulators Preoperational Test," was reviewed and selected portions of the test were witnessed. The inspectors found the testing activities to be acceptable as witnessed. One item of interest was discussed with the STE and startup supervision I concerning the verification of a special precautio )
This special precaution stipulated that the STE was !
I to ensure that no one was near the reactor vessel flange during performance of the accumulators discharge test. Startup management assured the I inspectors that verification of all prerequisite I conditions and special precautions were completed l l
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. 9 and documented in the official test log prior to commencement of actual test activitie One violation was identified. However, a notice of violation will not be issued as discussed abov . TMI Action Items (SIMS) (25565_1 (Open) TMI Action Item I.D.1, " Control Room Design Review": Supplement 6 (SSER 6) to the CPSES Safety Evaluation Report (NUREG 0797) required a supplement to the applicant's " Human Factors Control Room Design Review of Comanche Peak Steam Electric Station" report. This supplement was issued as Supplement 3 on September 15, 1988. On January 18 and January 19, 1989, the inspector and three NRC reviewers audited applicant's progress on the items identified in SSER 6 and Supplement 3 to the applicant's repor The auditors verified that all of the Human Engineering Discrepancies (HEDs) that require correction and NRC audit before licensing have been corrected and verified by the NR The HEDs that require environmental surveys are still open because the environmental surveys have not yet been performed. The surveys are awaiting completion of control room modifications that may affect the results of the surveys. The NRC will review the results of the surveys and any necessary corrective action. The inspectors understand that the surveys and any necessary corrective action will be completed prior to fuel loa Four of the five HEDs that require assessment and design improvements have been closed by the applicant and were reviewed and accepted by the auditor The remaining HED (Control No. 354), concerning high temperature in the Remote Shutdown Area, requires additional evaluation of the temperature, humidity, and airflow to determine corrective action. The NRC will review and verify the corrective action when the applicant completes the evaluation. The inspectors understand that the evaluation and corrective action will be completed prior to fuel loa Six HEDs were identified subsequent to the Detailed Control Room Design Review Survey. Resolutions for three of these were reviewed and accepted by the auditors. The three remaining HEDs are open and require action as follows:
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l . .Feedwater Bypass Valve Control Switch label does not-adequately reflect the switch: functio Although Supplement 3 to the applicant's. report indicated the labels.were changed, the auditors determined that this action was not yet complete.- The NRC will verify that the label is changed when th applicant's action is complet . The removable handles for the'four Auxiliary Feedwater Isolation Valve Control Switches does not have the valve code symbol utilized for~J-handle type switches. A modification has been issued for replacing.each switch with a fixed handle switc The NRC will review this modification when it is complete . The ventilation panel layout was.such that it was difficult to perform routine tasks and time sensitive emergency tasks in an efficient manne Panel X-CU-01 does not yet have a mimic installe ;
The NRC will review this item when the mimic is installe !
The inspector understands that action on these three HEDs '
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will be completed prior to fuel loa Details of the audit that do not require inspection '
effort will be documented in a future supplement to the CPSES Safety Evaluation Report (NUREG-0797). This TMI Action Item remains open pending completion of applicant action and subsequent NRC review as described abov (Open) TMI Action Item I.A.l.1.1, " Shift Technical'
Advisor on Duty." This item requires that each applicant provide an onshift technical advisor to the shift supervisor. The CPSES FSAR in Section I.A of the
" Response to the NRC Action Plan Developed as a Result of the TMI-2 Accident" commits to having available onsite, to each operating shift when the plant is being operated in Modes 1-4, an individual who is qualified to provide technical support to the shift supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis. The FSAR also states that the STA position is to-be eliminated when the qualifications of the shift supervisors have been upgraded. NUREG 0797 SER for CPSES indicates in Section 22.I.A.1.1 that the need for the shift technical advisor (STA) may be eliminated when the qualifications of the shift supervisors and senior operators have been upgraded and the man-machine interface in the control room has been acceptably upgraded. NUREG 0797 also states that the staff has not made a decision on the level of upgrading required for
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licensed operating personnel and the man-machine interface that could be-acceptable for eliminating the STA positio The qualifications and training of the STAS will be reviewed under TMI Action Item I.A.1.1.3. The progress of upgrading the man-machine interface in the control room will be reviewed under TMI Action Item I.D.1. TMI Action Item I.A.1.1.1 remains open pending NRC review pursuant to TI 2515/65) of administrative procedures concerning requirements for having the STA onsite to ensure they meet applicant's commitments in the FSAR and NRC requirements in NUREG-0737 and NUREG 079 c. (Open) TMI Action Item I.A.1.1.3, " Shift Technical Advisor (STA) Training." The item requires that the STA have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific L'
training in the response and analysis of the plant for transients and accidents. The STA is also required to receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The CPSES FSAR in Section 1.A of the
" Response to the NRC Action Plan Developed as a Result of the TMI-2 Accident" commits to the above stated requirement and additionally commits to an STA training program which consists of:
(1) College Level Fundamental Training (2) Management and Supervisory Training (3) Plant Specific Training (4) Simulator Training (5) Requalification Training .
NUREG-0797, "CPSES SER," Supplement 1 refers to a September 29, 1981, letter in which the applicant committed to meet the STA education, experience, and training requirements of NUREG-073 This item remains open pending NRC review (pursuant to Temporary Instruction 2515/65) of the applicant's STA training program to verify that it meets the commitments identified in the FSAR and September 29, 1981, letter and meets the provisions of NUREG-0797 and NUREG-073 d. (open) TMI Action Item I.A.1.2, " Shift Supervisor Administrative Duties." This NUREG-0737 item requires the following actions:
(1) The highest level of corporate management is to periodically reissue a directive that emphasizes the responsibility of the shift supervisor for safe
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operation of the plant and that clearly establishes the supervisor's command dutie (2) Plant procedures are to be reviewed to ensure that I
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the duties of the shift supervisor and control ^ room operator are properly. defined to establish a definite line of command and a clear delineation of-the command decision authority of the shift supervisor in the control room relative-to other plant management personne (3) Training programs for shift supervisors are to emphasize and reenforce the responsibility for safe operation and management function to ensure safet (4) The administrative duties of the shift supervisor are to be reviewed by the senior officer responsible for plant operations. Detracting or subordinate administrative functions are to be delegated to other operation's personnel not on duty in the control roo The CPSES FSAR in Section I.A.I.2 of the " Response to the NRC Action Plan as a Result of the TMI-2 Accident,"
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commits the applicant to the fourth item delineated abov Lack of commitments for the other three'
requirements does not relieve the applicant from having to implement them. NUREG-0797, "CPSES SER," Supplement 1 dated October 1981, icentifies an office memorandum dated July 8, 1981, and Procedure ODA-102, Revision 3 dated January 27, 1981, as containing the information required by items 1, 2, and 3 abov Because of the elapsed time and procedure revisions, this item remains open pending NRC review (pursuant to 1 Temporary Instruction 2515/65) of administrative procedures, applicable office memoranda, and the shift supervisor training program to ensure they meet applicant commitments in the FSAR and the provisions in NUREG-0737 and NUREG-079 e. (Open) TMI Action Item I.A.1.3.2.A, " Minimum Shift Crew Requirement." This NUREG-0737 item required implementing i
interim criteria for shift staffing in accordance with
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the letter of July 31, 1980, from D. G. Eisenhut to all power reactor licensees and applicants. Subsequently, this issue was the subject of a rule which wa promulgated as 10 CFR 50.54m on July 11, 1983, and was effective January 1, 198 This item remains open pending NRC review of plant l administrative procedures (pursuant to Temporary I r
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Instruction 2515/65) to ensure compliance with 10 CFR 50.54 (Open) TMI Action Item I.A.2.1.4, " Upgrade of Reactor Operator and Senior Reactor Operator Training and Qualifications / Modify Training." This NUREG-0737 item established the experience and training criteria required to sit for a hot license examination by referencing the March 28, 1980, letter from H. Denton. NUR'3-0737-specifically. excepted cold license applicancs from some of the specified requirements. The proof-and-review version of the CPSES Technical Specifications, paragraph 6.3.1, requires that licensed operators and senior licensed operators meet or exceed the requirements specified in' Sections A and C of Enclosure 1 of the March 28, 1980, letter from H. R. Denton-to all licensee This item remains open pending NRC review of the CPSES training program (pursuant to Temporary Instruction 2515/65) to ensure that the requirements of Sections A and C of Enclosure 1 of the March 28, 1980, letter from H. Denton are me (Open) TMI Action Item I.C.2, " Shift Relief and Turnover Procedure." This item required specific items to be included in a turnover checklist for the shift supervisor, the control room operators, the auxiliary operator, and technicians. In addition, a system was to be established to evaluate the effectiveness of the shift relief and turnover procedures (for example, periodic independent verification of system alignments).
Supplement 1 to NUREG-0797, "CPSES SER" documented a review of Procedure ODA-302, Revision 2, " Relief of Personnel" which described the shift relief and turnover procedures to be used by the applicant. The procedure was considered adequate except for a checklist for the auxiliary operator which was still under developmen Guidance and a check sheet were provided for the operation's engineer to periodically monitor and evaluate the effectiveness and completeness of the turnove Supplement 6 to NUREG-0797, documented a review of l Procedure ODA-302, Revision 3, " Relief of Personnel."
The revision was found to retain the acceptable checklists from Revision 3 and, in addition, incorporated a shift relief checklist for the auxiliary operators at the various watch stations throughout the plant. The auxiliary operator checklists were also considered acceptable.
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This item ^ remains open pending'NRC. verification that the'
current procedure that contains the' shift relief and turnover requirements still includes the items identified in NUREG-0797,-that the procedure is being properly implemented,-and that the effectiveness.of the turnover
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is being periodically monitored and evaluated by someone-in the current organization who is in a position equivalent to the operation's enginee (Closed) TMI Action Item I.C.3, " Shift Supervisor Responsibilities." This item concerns issues that are addressed in TMI Action Item I.A.l.2 which is discussed in paragraph'd above. TMI Action Item I.C.3 is considered closed to eliminate duplicate. trackin (Open) TMI Action Item I.C.4, " Control Room Access."
This item concerned limiting access to the control room to those individual necessary for safe plant operation Supplement 1 to NUREG-0797, "CPSES SER" documented the NRC staff's review of. Procedure ODA-306, Revision 2,
" Control Room and Observation Area Access," and Procedure ODA-102, Revision 3, " Shift Complement Responsibilities and Authorities." The review concluded that these procedures satisfied the requirement This item remains open pending NRC review that the current procedures which contain these requirements include'the provisions identified in NUREG-079 (open) TMI Item I.C.5, " Procedures for Feedback of Operating Experience to Plant Staff." This item required applicants for an operating license to prepare procedures to ensure that operating information pertinent to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. Specific aspects to be included in the procedures are identified in NUREG-0797, "CPSES SER,"
along with clarification of these aspect Supplement 1 to NUREG-0797 documented review of draft Procedure NOA-XXX, " Review of Significant Event Reports,"
Procedure ODA-107, " Reporting of operational Incidents,"
and Procedure ODA-106, Revision 0, " Review of Documents."
Other procedures were being developed to complete the L applicant's action. The staff concluded that the l- applicant was making acceptable progress in this area, but withheld final closecut until receipt and review of the remaining procedure Supplement 6 to NUREG-0797 documented the staff's review of an applicant description of the additional procedures
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! I that were established. On the basis of the review,.the NRC staff concluded that the procedures are in conformance with the TMI Action Plan requirements. This item remains open pending NRC verification that current 1 procedures contain the aspects described in NUREG-0797,-
that current procedures contain the essential aspects of
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the procedures identified in Supplement 1 to NUREG-0797, I and that current procedures contain the items described in Supnlement 6 to NUREG-079 . Significant Meetings (30702)
J During this inspection period, Dr. Thomas E. Murley, Director I of the Office of Nuclear Reactor Regulation, visited the CPSES sit The purpose of his visit was to apprise himself of the status of construction completion activity and the applicant's preparedness for licensing. He met with TU Electric corporate management personnel and with representatives from the Citizens Association for Sound Energy (CASE). Dr. Murley expressed a special interest in the Shutdown Control Board Simulator, the control room, and the training facility. He was favorably impressed with the applicant's maintenance training facility. His observations will be considered in the preoperational inspection effor . plant Tours (71302)
The NRC inspectors conducted numerous plant tours during this inspection period. These tours provided coverage during normal, off-normal, and backshift working hours. NRC 1
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inspection activities included reviewing work documentation, witnessing ongoing work activities, observing and interviewing shift operations personnel, reviewing the status of control room construction work, reviewing the status of system and component completion, determining the status of Units 1 and 2 equipment lay up, observing housekeeping activities, and inspecting for general safety complianc To support these activities, NRC inspectors attended plan-of-the-day meetings, discussed plant status with operations personnel, and observed shift turnover briefing During the course of the tours and inspections, no violations, deviations, or discrepancies were note . Technical Specification Review (71301)
Portions of the proof-and-review version of the CPSES technical specifications were reviewed to ascertain whether they are clear and enforceable and reflect the installed referenced system The review included accuracy of set points based on FSAR values, capability to perform
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surveillance as' stated in the Technical Specification, and i the_ policies in Section.6 of the Technical Specification This review is continuing and will be included in future ,
inspection' period . Follow-up of NRC Bulletins (92700)
Applicant action on the following bulletins will not be specifically reviewed because NRC review of related issues is, considered adequate to evaluate applicant. action. This administrative closure of the bulletins does not relieve the applicant of ensuring items in the bulletins are adequately; addressed and the NRC may choose to review the applicant action at a future date. The following bulletins are administratively closed for Units.1 and (Closed) NRC Bulletin 75-04, " Cable Fire at Browns-Ferr Nuclear Station." 10 CFR 50, Appendix'R, specifies the requirements to address fires. The NRC's review.and inspection of Appendix R issues adequately covers this are (Closed) NRC Bulletin 75-04 See Bulletin 75-04'abov (Closed) NRC Bulletin 75-04 See Bulletin 75-04 abov rl . (Closed) NRC Bulletin 79-05, " Nuclear Incident at Three Mile Island." NUREG 0737 issues are inspected in accordance with Temporary Instruction 2515/65 "TMI Action Items." The NRC's review and inspection of TMI Actions Items adequately covers this area.
.. (Closed) NRC Bulletin 79-05 See Bulletin 79-05 abov (Closed) NRC Bulletin 79-05 See Bulletin 79-05 abov (Closed) NRC Bulletin 79-05 See Bulletin 79-05 above.
, (Closed) NRC Bulletin 79-06, " Review of operator errors and system misalignment during TMI incident." See Bulletin 79-05 abov IClosed) NRC Bulletin 79-06 See Bulletin 79-05 abov (Closed) NRC Bulletin 79-06A, Revision See Bulletin 79-05 abov (Closed) NRC Bulletin 79-06 See Bulletin 79-05 abov . (Closed) NRC Bulletin 79-06 See Bulletin 79-05 above.
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10. Follow-up of 10 CFR 50.55(e) Reports (SDARs) (92700) (Closed) SDAR CP-86-23: "P-10 Permissive." This deficiency was initially reported to the applicant by Westinghouse. Additional background dealing with the deficiency is contained in NRC Information Notice N .
The deficiency involves the failure, under certain circumstances, of intermediate range reactor trips to be i reinstated when reactor power is reduced below 10 percen Specifically, the required.3-out-of-4 logic might not be satisfied when reactor power is reduced below 10 percent, if one power range channel had been placed out-of-service when reactor power was greater that 10 percent, and if the P-10 bistable for another power range channel had failed such that it was incapable of resetting. Therefore, a single failure could result in the loss of the intermediate reactor trip The potential loss of these intermediate reactor trips is of concern because, for certain accident scenarios, the margin to the departure-from-nucleate boiling-ratio (DNBR) could be reduce Therefore, the applicant has reported this deficiency under the provisions of 10 CFR
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50.55(e).
To correct this deficiency, Westinghouse suggested several actions. In general, the applicant should monitor the status lights for the F-10 bistables and for the P-10 permissive interlocks in both trains of the reactor protection system when power is being reduced below 10 percent. Furthermore, Westinghouse recommends that if the applicant cannot confirm proper P-10 interlock operation, then either the affected P-10 bistable should be placed in a nontrip position when ,
operating at or reducing power below 10 percent, or the reactor should be shutdown and the reector trip breakers opene The applicant has responded to this guidance by revising procedures to ensure that when reactor power is reduced below 10 percent, the operator verifies that the P-10 permissive has reset properly. Failure of this permissive to reset would be evidenced by numerous annunciator and bistable indication The NRC inspectors reviewed Procedures IPO-004A, Revision 2, " Plant Shutdown from Minimum Load to Hot Standby," and ABN-703A, Revision 2, " Power Range Instrumentation Malfunction." Procedure IPO-004A specifically directs the operator to verify that the intermediate reactor trips have been reinstated and that u___________
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each of the four power range P-10 bistables has rese procedure ABN-703A is entered if the operator is unable
]r to confirm indication of proper P-10 resetting. This procedure specifies that if P-10.cannot be reset within one hour.,.the plant should be shutdown, with all reactor trip breakers open, and all operations involving positive -'
reactivity additions should be. suspende The applicant's resolution of this' deficiency appears adequate. All procedures are~ consistent with Technical Specifications; the NRC inspectors have no further concern This: item is close .b.- (Closed) SDAR CP-86-74: "Rockwell Terminal Boxes." In 1985, numerous nonconformance reports identified internal wiring discrepancies and rusty fittings in'the terminal boxes for main steam isolation valves (MSIVs) supplied by Rockwell. The affected terminal boxes contain wiring for limit: switches located in solenoid actuated valves and-manually-operated bypass valve The applicant has concluded that the wiring discrepancies-and rusty fittings are not reportable under the provisions of 10 CFR 50.55(e) because, if the deficiencies had remained uncorrected, there would have been no adverse effect on the safety of plant' operation This conclusion is based on the following reasons:
. The bypass valve limit switches are used for valve position indication only. The bypass valves are'
open only when they are used to equalize pressure around the MSIV prior to opening the first MSIV on plant startup. During power operation, they will be locked closed. Therefore, a loss of position indication for the MSIV bypass valves would not have an adverse effect on plant safet . The solenoid valve limit switches are part of a test circuit for the MSIVs; their failure would not prevent proper operation of the MSIV The NRC inspectors have reviewed this matter and find its resolution acceptable. This item is closed, (Closed) SDAR 86-81: " BOP [ Balance of Plant]
Safety-related Instrument Setpoints. This issue originated from an applicant review of calculations for Balance of Plant (BOP) safety-related instrument setpoints which indicated that omissions of required data or use of incorrect information occurred in performing the calculations required per Regulatory Guide 1.105, Revision _ - _ _ - - _ _ - _ _ _ _ _ - -
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The inspectors reviewed selected BOP instrument setpoint o calculations performed-by Stone and Webster Engineering-L Corporation (SWEC). 'SWEC is in the process of performing-new calculations using revised methodology to supersede previous BOP'setpoint calculations. Approximately 80% of'
the calculations have been redone with no significant deficiency found. .The applicant's program will identify any deficiency during completion of the review using existing CPSES deficiency reporting programs, including safety significance. The inspectors consider this item
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close (Closed) SDAR CP-87-12): " Containment P/T [ Pressure /
Temperature] Analysis Computer Error." In June 1987, the applicant notified the NRC of a computer error-in a-containment pressure / temperature analysis. In order to remain conservative, the applicant reported'this
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deficiency under the provisions of 10 CFR 50.55(e).
In August 1987, the applicant reported that a reanalysis of the main steamline break (MSLB) accident had.been completed. This reanalysis resulted in an increase in the peak containment temperature (for the MSLB accident)
from the original (incorrect) value of 333 degrees F to 345 degrees F. Subsequently, the applicant has evaluated'
the effect on'the safety-related equipment inside containment and has determined that all equipment that is required to be qualified for the MSLB conditions is indeed qualified for the newly calculated peak containment temperature of 345 degrees Therefore, the applicant has concluded that this issue would not have created an adverse effect on plant safety if it had gone uncorrected. The NRC inspectors have j reviewed the applicant's actions in addressing this
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deficiency and find them acceptable. This item is close (closed) SDAR CP-87-14: " System Operability During Testing." This issue arose when Startup personnel conducted a tour of the Unit 2 cable spreading room and found both AC and DC power turned off to the LV-29A fire pane In addition to the power being secured, audio alarm logic cards No. 1 and No. 10 were remove Since this system was under control of the Startup organization's administrative procedures, it was required that the personnel deenergizing the panel and removing the cards officially notify the Startup group. The applicant generated two Deficiency Reports (P86-0015 and P86-0016) to document these conditions and has subsequently closed this issue based on the following:
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. A memo has been generated that identifies the status of fire detection and protection system In addition, this memo delineates the required method of operations /startup interfac . A "Special Order" that clearly outlines reporting requirements concerning deenergization and changes to fire detection and protection systems has been issue . Instructions on reporting requirements have been posted locally at the pane . Operations personnel have been trained on the requirements and necessity for reporting changes to system The inspectors have reviewed the applicant supplied documents and discussed this item with both operations and Startup personnel. No further examples were noted and the inspector is satisfied with the applicant's actions. Therefore, this item is considered close (Closed) SDAR CP-87-17: Validyne 15V DC Power Suppl This issue originated from vendor correspondence to the applicant. Specifically, Validyne Engineering Corporation (Validyne) sent a service bulletin to the applicant that described a deficiency in validyne brand power supplies. The power supplies involved contained 15V DC regulators that can exhibit a high frequency oscillation under certain load conditions. This oscillation can cause a minor zero drift in the output of some signal conditioning cards; however, calibration accuracy changes by less than 0.1 percen The applicant has evaluated this deficiency and has determined that Validyne power supplies are installed in only one system at CPSES, the Emergency Response Facility (ERF) computer. Because the ERF computer is a non-Class 1E system, and because the deficiency is of a minor nature, the applicant has determined that this deficiency is not reportable under the provisions of 10 CFR 50.55(e).
The NRC inspectors have reviewed the applicant's actions in addressing the deportability of this item and find them acceptable. This item is closed.
f 1 Maintenance Procram Implementation (62700, 62704, 62705)
The NRC inspectors met with the operations maintenance manager i
! and key personnel. These personnel discussed their program l l
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and goals'for maintenance. The inspectors then met with training" personnel and toured the electrical and mechanical training facilities. These facilities included laboratories where: electrical and mechanical equipment can be disassembled, tested, and reassembled. The training department has obtained a large number of faulty components which could'be operationally observe If effectively implemented, this training program can be of great benefit to TU Electric operations. The. personnel and facilities were impressiv This program is certainly a positive aspect of the TU Electric program During this inspection period, the NRC inspectors toured th maintenance shop several times to observe work in progres The reworking of a Unit 2 AFW pump was observed during different stages. The pump was being reworked because the machine surfaces that mate and hold the pump shaft were not exactly parallel with the centerline of the pump shaft and this caused the bearing to prematurely fail as a result of binding. -The. work was being accomplished in accordance with procedures and the work activity was inspected by TU Electric inspector No violations or deviations were identifie *. Exit Meeting (30703)
An exit meeting was conducted on February 7, 1989, with the applicant's representatives identified in paragraph 1 of this report. No written material was provided to the applicant by the inspectors during this reporting period. The applicant-did not identify as proprietary.any of the materials provided to or reviewed by the. inspectors during this
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inspection. During this meeting, the NRC inspectors'
summarized the scope and findings of the inspectio _ _ _ _ - _ - - - _ _ _ - ___- _. _ _ _ _ _ _