IR 05000416/1987029
| ML20236U527 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/24/1987 |
| From: | Butcher R, Dance H, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236U502 | List: |
| References | |
| 50-416-87-29, NUDOCS 8712030120 | |
| Download: ML20236U527 (11) | |
Text
UNITED STATES
[ka nf og% NUCLEAR REGULATORY COMMISSION
[\, //n REGION 81 101 MARIETTA STREET, N.W.
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- 1 g AT L ANTA, GEORGI A 30323 k9 * * . . + ,o'
Report No.: 50-416/87-29 Licensee: System Energy Resources, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Inspection Condu,cte - October 17 - November 13, 1987 Inspecto s: - [/' A R.' C. Butcher, enior Resident Inspector
///24![('7 Date Signed-lMathis,fResicentLJ% lll24/[~1 Dae Signed'
J.~ Inspettor Accompanying Insp tor: , Bernhard (October 19-23,1987)
Approved by: _
2-5" // 9 H.'C. Dance, Section Chief D6te Signed Division of Reactor Projects SUMMARY Scope: This routine inspection was conducted by resident and regional inspectors at the site in the areas of Licensee Action on Previous Enforcement Matters, Operational Safety Verification, Maintenance Observation, Surveillance Observation, ESF System Walkdown, Reportable Occurrences, Operating Reactor Events, Inspector Followup and Unreso?ved Items, Compliance with 10 CFR 50.62, Refueling Activities, and Scram Discharge Volume Capabilit Results: One violation was identified - Failure to follow procedure for cold weather protectio a71125 PDR ADOCK 05000416 G PDR
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REPORT DETAILS I
i l Licensee Employees Contacted j J. E. Cross, GGNS Site Director
- C. R. Hutchinson, GGNS General Manager R. F. Rogers, Manager, Special Projects 1
- A. S. McCurdy, Manager, Plant Operations 1 J. D. Bailey, Compliance Coordinator l M. J. Wrisht, Manager, Plant Support l
- L. F. Daughtery, Compliance Superintendent
- D. G. Cupstid, Start-up Supervisor R. H. McAnuity, Electrical Superintendent j
- J. P. Dimmette, Manager, Plant Maintenance !
- W. P. Harris, Compliance Coordinator l J. L. Robertson, Licensing Superintendent L. G. Temple, I & C Superintendent J. H. Mueller, Mechanical Superintendent
- L. B. Moulder, Operations Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent S. M. Feith, Director, Quality Programs
- S. F. Tanner, Manager, Quality Services Other licensee employees contacted included technicians, operators, security force members, and office personne * Attended exit interview Exit Interview (30703)
The inspection scope and findings were summarized on November 13, 1987, with those persons indicated in paragraph 1 above. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. The licensee had no comment on the following inspection findings:
416/87-29-01, Inspector Followup Item. Lack of a program to identify and evaluate abnormal area temperatures. (paragraph 4)
416/87-29-02, Violation. Failure to follow the cold weather protec-tion procedure. (paragraph 4)
416/87-29-03, Inspector Followup Item. Analysis to determine if silicone filled transmitters and Magnetrol transmitter response times are acceptable. (paragraph 13) Licensee Action on Previous Enforcement Matters (92702)
Not inspected this report perio !
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2 Operational Safety, Radiological Protection and Physical Security Verifi-cation (71707,71709,71714and71881)
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. The inspectors kept themselves informed on a daily basis of the overall i
plant. status and any significant safety matters related to plant opera-tions. Daily discussions were held with plant management and various
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members of the plant operating staf The inspectors made frequent visits to the control room such that it was visited at least daily when an inspector was on sit Observations included instrument readings, setpoints and recordings, status of operat-ing systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation,.
temporary alterations in effect, daily journals and data sheet entries, control room manning, and access controls. This inspection activity included numerous informal discussions with operators and their super-visor Weekly, Feature ESF) (when systemsthe inspectors were confirmed were onsite, selected operabl The confirmationEngineered is made-Safety by verifying the following: Accessible valve flow path alignment, power supply breaker and fuse status, major component leakage, lubrication, cooling and general condition, and instrumentatio General plant tours were conducted on at least a biweekly basi Portions of the control building, turbine building, auxiliary building and outside areas were visited. Observations included safety related tagout verifica-tions, shift turnover, sampling program, housekeeping and general plant conditions, fire protection equipment, control of activities in progress, problem identification systems, and containment isolation. The licensee's onsite emergency response facilities were toured to determine facility readines l The inspectors reviewed at least one Radiation Work Permit (RWP), observed health physics management involvement and awareness of significant plant activities, and observed plant radiation controls. The inspectors verified licensee compliance with physical security manning and access control requirements. Periodically the inspectors verified the adequacy of physical security detection and assessment aid The following comments were noted: !
During a tour of the Reactor Core Isolation Cooling (RCIC) system room it was noted that the room temperature was significantly higher than previously experienced. The licensee had already initiated Material Nonconformance Report (MNCR) 0285-87 noting RCIC room temperatures had reached 126 F at the ceiling and 122 F at the pum On October 18, 1987, the inspector measured 117.5*F in the RCIC roo The higher room temperatures are the result of steam leaks in the RCIC/ Reactor Water Cleanup Pump rooms. Equipment qualification
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calculations assumed a 105 F normal room temperature while the RCIC system .is not operating. The MN.CR requested an operability deter-mination based on long term exposure to temperatures greater than equipment qualification. Nuclear Plant Engineering (NPE) evaluated the conditions by taking worse case assumptions- and concluded that the most temperature sensitive component service life would end in October 1988. NPE requested more specific temperature data versus duration to determine the actual qualified lif There does not appear to be a program in place by the licensee to identify and evaluate abnormally high temperatures in areas that contain safety or safety-related equipment. For example, the RCIC room normal _ tempera-ture for equipment qualification is 105'F but TS 3.7.8 allows the area temperature for auxiliary building Emergency Core Cooling System (ECCS) rooms (the licensee treats the RCIC room as an ECCS room for this TS) to reach 150*F without any action, eight hours at 150 F requires a special report to the NRC and temperatures can go up to 180 F with no further actions required. These temperatures could significantly affect equipment qualification life. The lack of a program to identify and evaluate abnormal area temperatures will be identified as Inspector Followup Item 416/87-29-0 The inspectors checked Equipment Performance Instruction 04-1-03-A30-1, Revision 0, Cold Weather Protection and although the procedure was signed off as complete, severtl action items had not been completed. For example paragraph.7.3 requires portable heaters be setup in the breezeway between the auxiliary building and .the diesel generator building and a covering placed over the grating at both ends of the breezewa These tasks had not been accomplishe There are no individual sign offs for certain tasks and since the weather has been mild, operations decided to delay accomplishing certain tasks. This could create a problem since no task card would be issued to remind people certain tasks were not complete. The licensee has committed to revise the above procedure and require l individual task sign offs prior to closing the procedure as complet !
Failure to follow the procedure or otherwise identify that actions i were incomplete is a violation 416/87-29-0 l t Maintenance Observation (62703)
During the report period, the inspectors observed portions of the mainten-ance activities listed below. The observations included a review of the i Maintenance Work Orders (MW0s) and other related documents for adequacy, ,
l adherence to procedure, proper tagouts, adherence to technical specifica-tions, radiological controls, observation of all or part of the actual work and/or retesting in progress, specified retest requirements, and adherence to the appropriate quality controls. Maintenance activities were:
MWO E75421, Replace overload device and calibrate per 07-S-12-82.
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'l MWO M74975, Replace vent valve (F022A) on CRD system with Rockwell )
Edward l MWO M74957, Clean Inter-Condenser of Steam Jet Air Ejector !
MWO M75812, Chemically Clean ESF Switchgear Room Coole f i
MWO M75331, Removal Mechanical Snubber Q1G41G018R06 for Functional Testing During RF0 No violations or deviations were identifie . Surveillance Observation (61726)
The inspectors observed the performance of portions of the surveillance listed below. The observation included a review of the procedure for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual surveillance, removal from service and return to service of the system or components affected, and review of the data for acceptability based upon the acceptance criteria. Surveillance observed were:
06-0P-SP64-W-0001, Revision 23. TCN 5, Fire Pump Weekly Operabilit S-53-39, Revision 5, Calibration of 11521DP,1152DP and 1153DP Square Root Flow Transmitte EL-1R65-R-0001, Revision 26, Motor Operated Valve Thermal Overload Protection Device Functional Tes ME-1000-R-0002, Revision 22, TCN 1, Snubber Functional Tes IC-1821-R-0008, Revision 24, Division 2 ECCS Reactor Vessel Water Leve P-1C71-V-0002, Revision 23, Refueling Interlock Chec S-03-14, Revision 6, Chemical Additions to Plant System P-1C51-V-0001, Revision 28, Source Range Monitor Channel Func-tional Tes ME-1M61-V-0001, Revision 30, TCN 29, Local Leak Rate Test. (Valves M61F016 and M61F018).
06-0P-1C11-R-0011, Revision 23, Scram Discharge Volume Isolation Time Tes P-1F11-V-0001, Revision 21, TCN 1 & 2, Fuel Handling Platform Interlock Chec I l
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07-S-53-P75-34, d'sion 4, Fuel Oil Storage Tank Level' Loop Calibra-
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tion Instruction No violations or deviations were identifie . Engineered Safety Features System Walkdown (71710) l A complete walkdown was conducted on the accessible portions of the Containment Spray System. The walkdown consisted of an inspection and verification, where possible, of the required system valve alignment, including valve power available and valve locking where required, instru-mentation valved in and functioning; electrical and instrumentation cabinets free from debris, loose materials, jumpers and evidence of rodents, and system free from other degrading condition A No violations or deviations were identifie . Reportable Occurrences (90712 & 92700)
The below listed event reports were reviewed to determine if the informa-tion provided met the NRC reporting requirements. The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event. Additional inplant reviews and discussions with plant personnel as appropriate were conducted for the reports ,
indicated by an asterisk. The event reports were reviewed using the !
guidance of the general policy and procedure for NRC enforcement actions, regarding licensee identified violation The.following License Event Reports (LERs) are close LER N Event Date Event
- 87-015 September 17, 1987 Reactor Water Cleanup (RWCU) System Isolatio No violations or deviations were identifie . Operating Reactor Events (93702)
The inspectors reviewed activities associated with the below listed reactor events. The review included determination of cause, safety significance, performance of personnel and systems, and corrective actio The inspectors examined instrument recordings, computer printouts, operations journal entries, scram reports and had discussions with opera- L
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tions, maintenance and engineering support personnel as appropriat On November 4, 1987 at 6:30 p.m., valve G41F040 failed the Local Leak Rate Test. The plant was in Mode 1 at approximately 96% power. The G41F040 valve is an inside containment isolation check valve for the Fuel
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Pool Cooling and Cleanup system. TS 3.6.4 action statement requires that with one of the containment isolation valves shown in Table 3.6.4-1 inoperable, maintain at least one isolation valve operable in each affected penetration that is open and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the inoper-able valve to operable status or isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position. The licensee shut and deactivated motor operated valve G41F028 which is an outside containment isolation valve for the same penetration line as G41F040. Since there was no fuel in the containment fuel pool at this time, this system was not required. The licensee made a phone report to the NRC under 10 CFR 50.72 but later determined this condition was not reportable. The Resident Inspectors concurred in the licensee's determination that this event was not reportabl No violations or deviations were identifie . Inspector Followup and Unresolved Items (92701)
(Closed) Inspector Followup Item 416/87-18-06. The licensee has posted a red sign on Door 0C223 warning personnel that the normal position of that door is open to ensure proper room cooling. The identification sign on Door 0C223 has been changed to identify the Division 2 remote shutdown panel room as 0C208 . Compliance With The ATWS Rule, 10 CFR 50.62 (25020)
This inspection was counducted per TI 2500/20 to determine the licensee's compliance with the Anticipated Transients Without Scram (ATWS) rule, 10 CFR 50.62, determine the effectiveness of the Quality Assurance (QA)
controls applied to major activities and assess the operational readiness of ATWS equipment that is not safety related. The inspection findings for completed ATWS changes were as follows:
Licensee's Plan - In a letter dated October 14, 1985, (AECM-85/0322) the licensee submitted a plan to implement the requirements of paragraphs C 3, C.4 and C.5 of 10 CFR 50.62 prior to startup following the second refueling outage. The second refueling outace is scheduled to start on November 6,1987 and end on January 9,1988. Additional information has been submitted by letters to the NRC dated April 3, 1987 (AECM-87/0055),
June 22, 1987 (AECM-87/0109, August 13, 1987 (AECM-87/0152), and October 23, 1987 ( AECM-87/0190) . The licensee's plan requires them to neet the QA guidance of Generic Letter 85-06. The inspectors reviewed the licensee's letters as they were submitted and coordinated comments thru Regional Management to the Licensing Project Manage Design Engineering - The licensee has issued three Design Change Packages (DCPs) to incorporate the ATWS changes. The DCPs are 87/4005, 85/4053 and 87/300 Nuclear Plant Engineering (NPE) issued a Q list of safety
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related systems, structures and components for control during design, installation and/or modification purpose Appendix B to the Q list defines the QA program for non-safety related items. Appendix B was revised to include the Alternate Rod Injection System, the Standby Liquid Control System (SLCS) and the Reactor Coolant Recirculating Pump Automatic Trip Equipment and stated the requirements of Generic Letter 85-06 were to apply. Each DCP was reviewed and had a safety analysis attached and reflected the design as submitted to the NRC per the referenced letter At this time other changes are being discussed with the Licensing Project Manager and future changes will be verified as part of the DCP followu Quality Assurance and Qualifications - On July 31, 1985, the Director, Quality Programs issued a memorandum to the GGNS General Manager and the Director, Nuclear Engineering and Construction alerting them of the requirements specified in Generic Letter 85-06 to satisfy the implementa-tion of 10 CFR 50.6 As noted under the Design Engineering section, NPE issued a revision to appendix B of the Q list identifying the QA require-ments for non-safety related ATWS changes. A review of the affected DCPs showed that each had a complete engineering review cycle. Two of the three DCPs were treated as safety related due to their interface with safety related equipment such as the SLCS. The ATWS design, installation and test personnel for non-safety related tasks are generally the same people accomplishing safety related task Procurement - The inspector reviewed the procurement documents of the SI.CS hydraulic accumulators, observed that identification and storage controls were properly applied such that selected components could be identified and located. Some items were procured by Bechtel and some items we"e procured directly by the licensee. The requirements of Generic Letter 85-06 appeared to be me No violations or deviations were identifie . Refueling Activities (60710)
Refueling outage cycle two officially started on November 7,1987. The generator was disconnected from the power grid at 1:12 a.m. and the reactor was manually scrammed from 16% power at 2:22 a.m. on November 7, 1987. . The reactor was manually :: crammed to accomplish a surveillance of the scram discharge vent and drain valves from less than 50% rod density which is required every 18 months per TS 4.1.3.1.4.a. The inspectors verified the following:
The licensee had implemented controls for the conduct of refueling operations. Plant staff, NPE and OP issued organization charts defining personnel responsibilities and contact _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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A detailed outage schedule had been issued defining what minimum i
safety equipment had to be operable during each phase of the outage.
! Fuel receipt and inspection was previously reported in Inspection Reports 416/87-17 and 416/87-1 '
The inspectors attended various licensee outage presentations for operators and crafts personnel. The presentations were detailed and complet The licensee has submitted fuel reload data to the NRC for review and approva On November 8, 1987, the inspectors witnessed the licensee hydrolaze the refueling cavity in preparation for refueling activities. The inspectors verified that radiological controls were in place prior to the hydrolazing activit In addition, the following tagout of systeas were witnessed in part to verify that they were in accordance with establish procedures:
Safety Relief Valves Tagout Standby Liquid Control System Drain A Service Air Breakers 152-1501, 152-1514 and 152-1601 The loop licensee initiallytomaintained A was operable shutdown satisfy TS 3. cooling).with (Shutdown Selected RHR loop B and RHR maintenance and surveillance activities were witnessed as noted in paragraphs 5 and The inspectors reviewed the controlling procedure for refueling 101 03-1-01-5, Revision 20. TCN's 10 & 11. The procedure was reviewed to ascertain procedural and performance adequacy. The procedure was analyzed for embodiment of necessary test prerequisites, preparation instructions, acceptance criteria and sufficiency of technical conten The inspectors witnessed movement of new fuel into the containment fuel pool, removal of the drywell head, detensioning of the reactor vessel head and other in-containment refueling activities. Radiological and security controls were verified to be in plac No violations or deviations were identifie . Scram Discharge Volume (SDV) Capability (25590)
In conjunction with the PRA Inspection conducted October 13-23, 1987, a Regional Inspector conducted an inspection of the SDV per TI 2515/90, Scram Discharge Volume Capability. As a result of an incident at Browns Ferry in June 1980, in which control rods only partially inserted on a scram signal, the NRC issued IE Bulletins 80-14 and 80-17 (with
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supplements) and a Generic SER dated December 198 The IE Bulletins addressed short term fixes and the SER addressed the long term fix. TI !
2515/90 was written to verify compliance with the Generic SE The inspector performed a review of the plant Final Safety Anaylsis Report (FSAR), Technical Specifications, plant drawings, Alarm Response Instruc-tions (ARI), Surveillance Instructions (SI) And vendor drawings and instructions. A walkdown was performed of both banks of the SDV and Instrument Volumes (IV). Control Room indications for valve position of the vent and drain valves were verified to exis Vent and drain valve indicating lights indicate the condition of both vent or drain valves open (red) or at least one vent or drain valve closed (green). This is adequate indication to the operator of the drain and isolation functions but it is not possible to verify individual valve position. Control Room alarms were verified against the ARI The inspectors findings are formatted using the numbers corresponding to the items in Section 4 of TI 2515/9 FSAR Amendment 51 of 11/81 addressed the SER criteria on pages Q and R 4.6-18 thru 27. The following are the inspectors findings:
Item 4.01, Scram Discharge Volume Header Sizing The FSAR states a minimum volume of 3.35 gallons per drive was in the design specification. With 193 control rods, this would require 645 gallons of total volume above the scram setpoin Preoperational test IC11PT01, Rev.1, Step 7.5.44 verified the available volume was 944 gallons. This item is close Item 4.02, Automatic Scram on High SDV Level The inspector reviewed ARI 04-1-02-1H13-P680-5A-A1, R33, CRD Disch Vol Wtr Lvl HI Trip; 06-1C-1011-M-0001, R21, CRD Scram Discharge Volume High Water Level (RPS) Functional Test; 06-1C-1011-R-0001, R24, Scram Discharge Volume High Water Level (RPS) Calibration; 06-10-1011-M-0003, R21, Scram Discharge Volume High Water Level Float Switches (RPS) Calibration. The FSAR was reviewed. The preoperational test procedure referenced in Item 4.01 verified adequate discharge volume over the scram setpoint. This item is close It was noted that ARI 04-1-02-1H13-P680-5A-A1, R33, did not have clear actions to indicate to the operator what actions to take if a scram did not occur as required. The misleading statements were going to be omitted in a procedure change request filed by the operations staff after the inspector questioned the procedure.
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Item 4.03, Instrument Taps Not On Connected Piping
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.A system walkdown was performed to confirm this item. This item is close Item'4.04, Detection of Water in the IV Grand Gulf employs a Gould PD/PDH3018 series remote seal transmitter, using a silicon filled capillary tube, for system-alarm and rod block circuits, and .as one of two diverse level indicators for the scram circuit. The transmitters output feeds Rosemount 510DU master and slave trip units. There is level indication of the back panel at these trip units that can be read by the operators. .The alarms'and rod-blocks are annuni-cated on the IH13-P680 panel. A review was conducted of 06-IC-1C11-M-0002, R24, Scram Discharge Volume High' Water Level (Rod Block) Functional Test and 06-IC-1011-R-0002, R-24 Scram Discharge Volume High Water Level (Rod Block) Calibration. I addition, ARI 04-1-02-1H13-P680-4A2-C4, Rev. 33, CRD Disch Vol Not Drained and 04-1-02-1H13-P680-4A2-C5, Rev. 34 Cont Rod Withdrawal Block, were reviewed. This item is close Item 4.05, Vent and Drain Valves System Interfaces This item requires that design of the system precludes backup of water through the vent or drain valves to the instrument volum , Piping and Instrumentation Drawing Diagram (number M-1081A, Rev.P&ID), Control Rod Drive Hydraulic Sys the configuration of the vent and drain valve There are vacuum breakers' installed on the vent lines that preclude wate from the suppression pool from backing up into the volume. The plant walkdown confirmed the existence of the vacuum breakers ;
Item 4.06, Vent and Drain Valves Close On Loss of Air A review of the FSAR and P&ID verified this item is complet Item 4.07, Operator Aid This item asks that instrumentation be provided to aid the operator in detection of water in the IV prior to scram initia-tion. See Item 4.04 for verification of operator indications.
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l This item is close Item 4.08, Active Failure in Vent and Drain Valves Visual Inspection and P&ID review verified redundant vent and drain valves are installed to ensure a single active failure will not prevent isolation of the system when required. This item is closed.
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Item 4.09, Periodic Testing of Vent and Drain Valves Review of 06-0P-1011-Q-0009, R24, Scram Discharge Volume Vent l and Drain Valves Operability Test, verified a quarterly func-l tional surveillance is provided to test the valve P-1C11-R-0011, R23. Scram Discharge Volume Isolation Time Test, provides ASME timing every 18 month This item is close Item 4.10, Periodic Testing of Level Detection Instrumentation Covered in Items 4.02 and 4.0 This item is close Item 4.11, Periodic Testing of the Entire System 06-0P-1011-R-0011, Covered in Item 4.09 also provides system testin This procedure scrams the plant with a control rod density of less than or equal to 50% and verifies all vent, drain, alarm and scram function Other Items Questions brought up by the inspector resulted in the need to re-evaluate the response time requirements for the transmitter The GE RPS Instrument Design Data Sheet requires a response time of less than one second. The silicone filled (Gould) trans-mitters response time is slightly higher than this, but is less than two second Analysis on Magnetrol response time due to the long piping length to the instrument volume is also being evaluated. This issue was raised by IE Notice 87-17. Followup of SDV IV instrument response times will be Inspector Followup Item 416/87-29-0 No violations or deviations were identified, i
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