IR 05000261/1979011
| ML14175B068 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/09/1979 |
| From: | Belisle G, Kellogg P, Mchenry T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14175B063 | List: |
| References | |
| 50-261-79-11, NUDOCS 7909110490 | |
| Download: ML14175B068 (12) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 Report No. 50-261/79-11 Licensee:
Carolina Power and Light Company 411 Fayetteville Street Raleigh, North Carolina 27602 Facility Name:
Robinson 2 Docket No. 50-261 License No. DPR-23 Inspection at Robnson Site near Hartsville, South Carolina Inspectors:
T. 'J. McHenr
/ate Signed fWG iat Sied Approved by:
,
/,
,
P.Z n Chief, RONS ate Signed SUMMARY Inspection on May 21-25, 1979 Areas Inspected This special, announced inspection involved 84 inspector-hours onsite in the area of followup on IE Bulletin Results Within the area inspected, two apparent items of noncompliance were identified; (Infraction - Failure to change procedures as required by Technical Specifica tions -
paragraph 5.b);
(Deficiency -
Post maintenance procedures failed to restore equipment to operable status - paragraph 5.d).
79I911OL-V0
DETAILS Persons Contacted Licensee Employees
- R. B. Starkey, Jr., Plant Manager
- C. W. Crawford, Operations Supervisor
- H. S. Zimmerman, Maintenance Supervisor
- J. M. Curley, Engineering Supervisor
- B. W. Garrison, Quality Assurance Supervisor D. S. Crocker, Environmental and Radiation Control Supervisor C. Wright, Engineering Technician M. Page, Engineer R. H. Chambers, Senior Engineer R. S. McGirt, Senior Nuclear General Specialist W. T. Traylor, Administrative Supervisor F. Bishop, Electrical Engineer F. Lowery, Training Coordinator Other licensee employees contacted included construction craftsmen, operators, and office personne *Attended exit intervie.
Exit Interview The inspection scope and findings were summarized on May 25, 1979 with those persons indicated in Paragraph 1 abov The licensee acknowledged the items of noncompliance identified in paragraph 5.b and In ad dition plant management made several commitments associated with the items of noncompliance and other inspection findings as follows: Prior to reactor startup (Note 1), auxiliary feedwater pump discharge valves AFW-45 and 46 would either be fully opened or throttled and tested to assure the design flow rate of 300 gpm for each pump is maintained (paragraph 5.b). Prior to reactor startup (Note 1),
both emergency diesel generator governers will be set to the proper Load Limit and Synchronizer Indi cator values (paragraph 5.b.). Prior to reactor startup (Note 1),
all safety-related instrument cabinets, relay panels and termination cabinets will be inspected and cleaned if necessary (paragraph 9). Prior to reactor startup (Note 1), Emergency Procedure EI-1 (Incident Involving Reactor Coolant System Depressurization) will be reviewed by all operating shift personnel; in addition, a procedure walk through will be performed by all licensed shift personnel (paragraph 5.a).
-2 Prior to use, each Operating Work Procedure (OWP) will be-reviewed to assure that the OWP properly restores the system, component, etc. to an operable alignment (paragraph 5.d). Licensee Action on Previous Inspection Findings Not-inspecte.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve noncompliance or deviation New unresolved items identified during this inspection are discussed in paragraphs 12 and 1.
Review of IE Bulletin 79-06A - Operational Errors and System Misalignments Identified During the Three Mile Island Incident The inspector reviewed the licensee's actions with regard to IE Bulletin (IEB)
79-06 The inspection activities involved a review of operator training, engineered safety feature (ESF)
systems and procedure These reviews were conducted to verify the adequacy of the licensee's system alignments, administrative controls, procedures and training in regard to each item addressed in IEB 79-06A. The following items summarize the inspector's findings in the areas inspecte Review of Operator Training The inspector reviewed the licensee's training records to verify that operators and supervisors had received training in the following areas:
(1) Procedures changes initiated as a result of IEB 79-06 (2) Specific measures which provide assurance that ESF system are available, in particular, measures for returning such systems to operable status following maintenance and testin (3) Specific measures to assure that automatic actuation of ESF systems are not overridden except as permitted by IEB 79-06 (4) Automatic actions initiated by reset of ESF systems that could effect the control of radiactive liquids and gase (5) Directives for early NRC notification of serious event The licensee's training records indicated that all licensed and non licensed operating shift personnel had received training on May 18 and 21, 1979 in the five areas identifie Specifically, the licensee's
training program was outlined in the following topics:
-3 (1)
Review of the Three Mile Island (TMI) incident including component failures and judgement error (2) Review of IEB 79-06A and the licensee's response including details of procedure and design changes resulting from the bulleti (3) Review of design differences between ThI and the licensee's facilit In addition to the licensee's training, 58 licensee personnel attended briefing sessions conducted onsite by the NRC on April 23, 197 Details of this training are discussed in Inspection Report 50-261/79-0 The inspector interviewed licensed operators and supervisors on each shift to assure that the information included in the training programs was known and understoo Specifically, the inspector verified that each person interviewed adequately understood the following:
(1) The requirements for maintaining reactor coolant pumps running following accident (2) The requirements to maintain a 50 degree subcooling and high head safety injection in operatio (3) The requirements to utilize multiple indication of parameters when availabl (4) The contributing factors to the TMI inciden Based upon the inspection activities performed in the training area, the inspector determined that the licensee's actions with regard to training appear satisfactor However, as a result of substantial last minute changes to emergency procedures (described in paragraph 5.f.) it appeared that additional training on the finalized procedures should be conducted. This item was discussed at the exit and plant management made a committment that; prior to reactor startup, (Note 1)
the following would be completed:
(1) EI-1 (Incidents Involving Reactor Coolant System Depressurization will be revised, reviewed and approve (2) EI-1 will be reviewed by all operating shift personne (3) A walk through of EI-1 will be performed by all licensed shift operating personne Inspection of Engineered Safety Features (ESF) Procedures and Alignments The inspector reviewed ESF system valve, breaker and switch alignment operating procedures against current system drawings to verify the adequacy of alignment procedure In addition, a system walk down of
-4 each ESF operating procedure was performed to verify that all acces sible valves, breakers, and switches were in the proper position, including verification that all valves required to be locked were actually locked. The following ESF system Operating Procedures (OP's)
were reviewed and walked down:
(1) OP-6, Service Water (2) OP-7, Diesel Generators (3) OP-14, Auxiliary Feedwater (4) OP-38, Residual Heat Removal (5) OP-40, Component Cooling Water (6) OP-42, Safety Injection and Containment Spray During the walk down of the above systems the inspector identified the following discrepancies between the required position of valves or settings and the actual position of the valves or setting (1) Diesel generator day tanks drain valves (FO-11A and B) are re quired to be closed by OP-7A valve checklis These valves were found open with tygon tubes attached for level indicatio (2) Auxiliary feedwater pump discharge valves (AFW-45 and 46) are required to be open by OP-14A checklist. These valves were found to be throttle (3) The Load Limit and Synchronizer Indicator settings on both emer gency diesel generator governors were found not set at values required by OP-7A checklis The inspector discussed these above discrepancies with licensee person nel and identified these items as examples of an apparent item of noncompliance with Technical Specification 6.8.2 which requires that proposed changes to operating procedures be reviewed by the PNSC and approved by the Plant Manager. Prior to the conclusion of the inspec tion, licensee personnel were unable to determine the reason for the discrepancy between the required and actual diesel governer setting This item was discussed at the exit and plant management made a com mitment that prior to reactor startup (Note 1), the proper setting of diesel governors will be determined and the governors adjusted to the required value The failure to review and approve operating procedure changes has been identified as an infraction in the Notice of Violation (261/79-11-01).
The inspector interviewed operators to determine the bases for the fact that the motor driven auxiliary discharge valves (AFW-45 and 46)
discussed above were throttled instead of fully ope It was deter mined that the identified valves were throttled due to fact that the auxiliary feedwater pumps provided flow in excess of the design and desired requirments. It was further determined that no procedural or administrative controls existed to insure that these valves were
-5 placed in the required position when auxiliary feedwater is required to be operabl This item was discussed with plant supervisory per sonne The inspector stated that administrative controls are re quired to assure the design flow rates described in the FSA This item was discussed at the exit and plant management made a commitment that AFW-45 and 46 would either be fully opened or throttled and
- tested to assure 300 gpm from each pump prior to reactor startup (Note 1). This item has been identified as an open item (261/79-11-02). Review of ESF Periodic Test Procedures The inspector reviewed periodic test procedures for ESF systems to verify that when completed, the systems will be returned to an oper able conditio In addition, the last completed periodic test for each ESF system was reviewed to assure that the acceptance criteria were met. The following periodic test procedures (ST's) were re viewed:
(1) PT-2.9A, B and C, RHR System Component Test (2) PT-3.4 A, B and C, Containment Spray System Component Test (3) PT-4.1 A, B and C, Service Water System Component Test (4) PT-2.7 A, B and C, Safety Injection System Component Test (5) PT-22.1 A, B and C, Auxiliary Feedwater System Component Test (6) PT-36.2, Component Cooling Water System Component Test (7) PT-23.1, Emergency Diesel Test (8) PT-40.0, Inservice Quarter Valve Test No problems were identified in the areas of system restoration fol lowing ESF periodic testing or acceptance criteria of ESF periodic test Review of ESF Systems Restoration Following Maintenance of Exten ded Outage The inspector conducted a review of the licensee's administrative controls to assure proper restoration to service of ESF components and systems following maintenance and extended outage Maintenance restoration for ESF system is controlled by Operating Work Procedures (OWP's) which are approved by the Operations Supervisor. The following OWP's were reviewed to insure that ESF systems are restored to oper able alignmen (1) SW-l,3 and 5, Service Water Systems (2) RAR-1, 3 and 5, Residual Heat Removal System (3) S15-i, 3, 4 and 5, Safety Injection System (4) CS-1, 2, 3 and 4, Containment Spray System (5) AFW-1, 3 and 7, Auxiliary Feedwater System (6) CC-1, 2 and 7, Component Cooling Water System (7) DG-1 and 2, Diesel Generators
-6 During the review of the above OWP's it was determined that the system restoration alignment requirements specified in SIS-3 and CS-2 for safety injection pump C and containment spray pump B respectively, did not restore these components to an operable alignment conditio Specifically, the discharge and recirculation valves on the safety injection pump remained closed and the suction valve on the spray pump remained closed following system restoratio The inspector reviewed the licensee's maintenance file on the affected pumps and determined that OWP's SIS-3 and CS-2 had not been used since issue This item was discussed at the exit and identified as an apparent item of non compliance with Technical Specification requirement 6.8.1, which requires that written procedures meet the requirements of ANSI N18.7 1972, Section 5.3. ANSI N18.7-1972, Section 5.3.5(3) requires post mainten ance return to service instructions which provide special attention for restoration of normal conditio Due to the potential affect of the failure to restore ESF systems following maintenance, plant man agement committed to a review of the restoration to service section of all OWP's prior to their us This item has been identified as a deficiency in the Notice of Violation (261/79-11-03).
In the area of extended outages, the inspector determined that the licensee's administrative controls require complete valve alignment of all related systems following refueling outages and all affected safety related system following extended maintenance outage These controls appear adequate to assure restoration of ESF systems following outages and extended maintenanc Review of Pressurizer Level Protection System The inspector reviewed the licensees action with regard to Item 3 of IEB 79-06A which requires pressurizer level bistable input to safeguards actuation be placed in the tripped condition. The inspector determined by logs and records that the pressurizer level bistables had been tripped during operation in accordance with the requirements of IEB 79-06A. In addition, the licensee had submitted a license amendment request to modify the existing logic for safeguards actuation to a 2/3 actuation on low pressure with no input from pressurizer leve The licensee plans to install the design modification prior to plant start up (Note 1) if the license amendment is approved; otherwise, operation will continue with pressurizer level bistables trippe The inspector also reviewed PT-5.2 (Pressurizer Pressure Protection Channel Test) and PT-5.3 (Pressure Level Channel Test) to verify that these procedures reflected the changes described above. It was deter mined that these procedures had not been revised; however, revisions were pending based upon either installation of the design change or operation with pressurizer level bistables trippe In any case, the licensee stated that appropriate revisions would be made prior to implementing PT-5.2 and PT-Assessment of Operating Procedures and Plant Practices The inspector reviewed the following areas to ascertain information concerning the licensee's operating policies, practices and procedures:
(1) The inspector reviewed the licensee practices for placing tags on control panels to ascertain whether the potential exists for obscuring status indicatio The licensee routine practice in this area is to secure control panel tags such that indication is not blocked from view. This practice is accomplished by either taping the tag in place or by rolling the tag and placing it under the switch handle to secure the tag. In both cases strings are also used to preclude the loss of a tag from its proper location. The inspector verified the above tagging practices in effect during all control room inspections and plant tour No problems were identifie (2) The inspector discussed plant transients with operating personnel to ascertain if additional charging pumps are required to assist pressurizer level control during transients. Based upon operator responses, control of pressurizer level during transients has not been a problem and does not require operation of safety injection pump (3) The inspector reviewed operating and emergency procedures to ascertain whether a procedure existed for recovery from a dry steam generator condition. It was determined that no procedure for feeding a dry steam generator existed. The licensee's posi tion on this matter is that the steam generator inventory and auxiliary feedwater capability insure that operable steam gener ators will not go dry during accidents analyzed in the FSA Review of Emergency Instructions The inspector reviewed EI-i (Incidents Involving Reactor Coolant System Depressurization), to ascertain if all revisions required by IEB 79-06A had been incorporate Specifically, the inspector veri fied that the appropriate revisions to address the following concerns had been incorporated into EI- (1) Procedure instructs operators to not override automatic actions of ESF systems unless continued operation will result in unsafe plant condition (2) Procedure requires a 50 degree Fahrenheit subcooling as a condi tion for maintaining high pressure injection operatio (3) Procedure requires at least two reactor coolant pumps remain in operation as long as the pumps are providing forced flo (4)
Procedure requires operator to utilize various parameters in evaluating plant conditions and do not rely upon single indica tors, especially pressurizer leve In regard to the above items and the other procedure revisions re quired by IEB 79-06A, no problems were identifie However, a review
- of subsequent actions of EI-1 indicated that the sequence of steps contained numerous steps not required as part of the actual recircu lation switchove Specifically, the operator is given only ten minutes to complete the switchover sequence and during that time the procedures require manual valve closure and breaker operations which are not required to accomplish the recirculation switchove This item was discussed with licensee personne The inspector determined that additional steps in the switchover sequence resulted from pre vious NRC requirement The inspector stated that the requirement to insure that core cooling water is restored promptly during the switch over sequence is paramount. Therefore, the recirculation switchover sequence in EI-1 should be revised to limit the operator actions to only the steps necessary to accomplish the switchove Licensee representatives stated that the swtichover sequence of EI-1 would be revised as recommended prior to startup (See Note 1).
The inspector identified additional emergency procedure discrepancies which were discussed with cognizant individuals responsible for pro cedure revision. The following apparent emergency procedure discre pancies are open pending review on subsequent inspection (261/79-11-04).
(1)
The adequacy of emergency procedures in the area of natural circulation cooldow (2)
The compatability of Overall Plant Operating Procedures when referenced by Emergency Procedure.
Review of IE Bulletin 79-07
-
Seismic Stress Analysis of Safety-Re lated Piping The inspector reviewed the licensee's response to IEB 79-07 which indicated that upon reanalysis, the maximum stress was calculated to be 3.0 ksi as compared with allowable stress of 13.1 ksi. In addition, the licensee's response has been reviewed by NRC Division of Operating Reactor This review has concluded that the licensee's reanalysis was conducted with both approved and acceptable results. Therefore IEB 79-07 is close.
Review of IE Bulletin 79-04
-
Incorrect Weights For Velan Swing Check Valves The inspector reviewed the licensee's preliminary response to IEB 79-04 in which the licensee committed to a complete response to IEB 79-04 prior to startup (Note 1) and discussed this item with cognizant licensee personne The licensee had determined that six three inch valves had been analysed at 60 pounds and the actual weights for each valve was 85 pound The licensee
-9 had evaluated the effect of additional weight from a piping seismic analysis standpoint and no problems were identified. However, data on reanalysis of supports had not been receive Subsequent to the above inspection the licensee's response to IEB 79-04 was received in the Region II offic Based upon an in office review of the response, it appears that the licen see's actions in regard to IEB 79-04 are acceptabl However, this bul letin will remain open pending a detailed review of the analytical tech nique and resultant data included in the respons.
Review of IEB 79-02 -
Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts The inspector reviewed the licensee's procedure for inspection and testing of expansion anchor bolts and discussed the inspection schedule with the plant supervisor responsible for licensees activities on IEB 79-0 The inspector determined that the license has divided the support plates into three groups as follows:
a. Priority 1 -
Seismic 1 safety-related support plates on nonredundant component b. Priority 2
-
Seismic 1 safety-related support plates on redundant component c. Priority
-
Seismic 1 safety-related support plates on non safety-related Q-List component The inspector reviewed the licensee's testing program which is outlined as follows:
a. Priority 1, 2 and 3 support plate loadings have been recalculated using flexible vice rigid base plate Using a detailed procedure, inspect/test one anchor per support plate having a design factor less than te If an anchor fails the inspection/test then the inspection/testing of all anchors in the associated plate is require All priority 1 and 2 anchors failing the inspection/test are repaire Detailed records of the testing program are to be recorded and main taine The inspector determined that the licensee's schedule requires that all priority I and inaccessable priority 2 support plates be inspected and repaired if necessary prior to startup (Note 1).
In addition, inspection/
testing of all other supports are to be completed by the required IEB com pletion dat The inspector discussed the failure rate and inspection results with cogni zant licensee personne Based upon this preliminary review it appears that the licensee's actions on IEB 79-02 will satisfy the principle cri terion; that testing must reasonably assure that each Seismic Category I piping system will perform its intended functio However, this IEB will remain open pending the completion of the testing program, a review of the lieensee's response and a detailed inspection of the technical adequacy of the licensee's inspection progra.
Fire Retardant Material On Plant Equipment During discussions with plant personnel, the fact that residue from fire retardant materials sprayed on cables during the outage had settled on instrument panels and relay rack The inspector toured the instrument cabinet rooms and determined that some residue had deposited on exposed relay/termination racks but there was no evidence that the materials were inside of safeguards and protection instrumentation cabinet This item was discussed with the maintenance supervisor who stated that a cleaning program had been implemented. This matter was also discussed at the exit, plant management committed to inspect and clean if necessary, all safety related instrument cabinets, relay panels and termination racks prior to plant startup (Note 1).
The inspector identified this item as open pending a followup inspection of associated equipment during a subsequent inspec tion (261/79-11-05).
10. Electrical Distribution Manual During the inspection activities discussed in Paragraph 5, the inspector noted that the electrical distribution manual in the control room did not appear to be current. Further it did not appear that the electrical dis tribution manual was controlled by administrative measures to assure up dating and appropriate review. This item was identified as open pending a review during a future inspection to determine the utilization of the electrical distribution manual for safety-related activities (79-11-06).
11. BFD Relays This item was previously identified and discussed in inspection report 50-261/ 78-27 and LER 50-261/78-29. The inspector discussed the licensee's actions on this item with cognizant licensee personne It was determined that all BFD relays used in safeguards and protective systems had been replaced with a new type of NBFD relay The licensee also intends to continue a testing program on approximately twenty various styles of NBFD relay. This test program consists of monthly testing of spare NBFD relays to monitor their performance. The inspector had no further questions in regard to this ite. Service Water Valve Checklist During the valve lineup verification on the service water system, the inspector found that service water valve SW-564 was not listed on the valve
-11 checklist (OP-6A) but was actually in plant and shown on the Q-List drawing This item was identified to plant personnel who immediately initiated the necessary changes to include valve SW-564 on valve checklist OP-6 The inspector reviewed previous OP-6A revisions and service water design modi fication to in order determine why SW-564 was not on the alignment check list, since it appeared that the valve may have been added as a modifica tio However, no determination was made as to whether the valve was installed as part of a design chang The inspector identified this item as unresolved pending a determination of the reason SW-564 was omitted from the valve alignment checklist (261/79-11-07).
13. Valve Tagging During the valve alignment system walkdown the inspectors noted many missing identification tags on valves. The inspector discussed this item at the exit and identified this item as unresolved pending a determination of the licensee's commitment to 10 CFR 50, Appendix B, Criterion XIV (261/79-11-08).
NOTE 1 -
Startup refers to the initial reactor startup following the refueling outage.