IR 05000416/1986041
| ML20210D905 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/28/1987 |
| From: | Butcher R, Dance H, Will Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20210D873 | List: |
| References | |
| 50-416-86-41, IEB-86-001, IEB-86-003, IEB-86-1, IEB-86-3, NUDOCS 8702100189 | |
| Download: ML20210D905 (14) | |
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[km na r UNITED STATES
- o NUCLEAR REGULATORY COMMISSION
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REGION 11 -
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101 MARIETTA STREET, N.W.
- ATLANTA. GEORGI A 30323
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Report No.:
50-416/86-41 Licensee:
System Energy Resources, Inc.
Jackson, MS 39225-3070 Docket No.:
50-416 License No.: 'NPF-29
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Facility Name: Grand Gulf Nuclear Station Inspection Condu tedi December 20, 1986 - January 16, 1987-Inspecto
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R.C.jBu cher, Senior' Resident Inspector Date Signed
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W.T.
mith, Resident Inspector Date Signed rk7 Approved by:
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H'.C'. Dance, 3ection Chief, 4 ate Sigrfed Division of Reactor Projects SUMMARY Scope:
This routine inspection was conducted by the resident 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 Unresolved Items, Design Changes and Facility Modifications and IE Bulletin Followup.
Results:
No violations or deviations were identified.
8702100189 870129 PDR ADOCK 05000416 PDR o
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I REPORT DETAILS 1.
Licensee Employees Contacted
- J.E. Cross, GGNS Site Director
- C.R. Hutchinson, GGNS General Manager
- R.F. Rogers, Manager, Unit 1 Projects A.S. McCurdy, Manager, Plant Operations
- J.D. Bailey, Compliance Coordinator M.J. Wright, Manager, Plant Support
- L.F. Daughtery, Compliance Superintendent 0.G. Cupstid, Start up Supervisor R.H. McAnuit R.V. Moomaw,y. Electrical Supe *intendent W.P. Harris, Compliance CoordinatorTech. Assistant to the Manager, Plant M
- 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
- J.P. Dimmette, Manager,/ Radiation Control Superintendent Plant Maintenance
- H.D. Morgan, Unit 2 Construction
- S.M. Feith, Director, QA
- F.W. Titus, Director, NPE
- 0. Stonestreet, Manager, PM&C
- W.E. Edge, Manager, QA
- W.C. Eiff, Principle QE, NPE Other licensee employees contacted included technician force members, and office personnel.
s, operators, security
" Attended exit interview.
2.
Exit Interview The inspection scope and findings were summarized on Janu those persons indicated in paragraph 1 above ary 16,1987, with as proprietary any of the materials The licensee did not identify
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inspectors during this provided to or reviewed by the inspection.
The following inspection findings:
licensee had no comment on the 416/86-41-01, Inspector Followup found during ESF walkdown (paragraph 8). Item (IFI).
Correction of deficiencies 416/86-41-02, Unresolved Item.
cance of apparent overpressurization of pressure gauge E22 P graph 8).
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REPORT DETAILS 1.
Licensee Employees Contacted
- J.E. Cross, GGNS Site Director
- C.R. Hutchinson, GGNS General Manager
- R.F. Rogers, Manager, Unit 1 Projects A.S. McCurdy, Manager, Plant Operations
- J.D. Bailey, Compliance Coordinator M.J. Wright, Manager, Plant Support
- L.F. Daughtery, Compliance Superintendent D.G. Cupstid, Start-up Supervisor R.H. McAnuity, Electrical Superintendent R.V. Moomaw, Tech. Assistant to the Manager, Plant Maintenance W.P. Harris, Compliance Coordinator
- 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
- J.P. Dimmette, Manager, Plant Maintenance
- H.D. Morgan, Unit 2 Construction
- S.M. Feith, Director, QA
- F.W. Titus, Director, NPE
- D. Stonestreet, Manager, PM&C
- W.E. Edge, Manager, QA
- W.C. Eiff, Principle QE, NPE Other licensee employees contacted included technicians, operators, security force members, and office personnel.
- Attended exit interview.
2.
Exit Interview The inspection scope and findings were summarized on January 16, 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:
416/86-41-01, Inspector Followup Item (IFI).
Correction of deficiencies found during ESF walkdown (paragraph 8).
416/86-41-02, Unresolved Item. Identification of causes and safety signifi-cance of apparent overpressurization of pressure gauge E22-PI-R002 (para-graph 8).
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416/86-41-03, IFI.
Licensee determination if any containment penetrations have' been accepted based upon misapplication of Regulatory Guide 1.11 (paragraph 11).
3.
Licensee Action on Previous Enforcement Matters (92702)
(Closed) Violation 416/84-16-01. Failure to perform a safety evaluation on a procedural change. This violation was reviewed for closure in Report 416/86-02 but it was found that the licensee failed to incorporate the corrective actions committed to in their response.
Subsequently, in the review of the Updated Final Safety Analysis Report (UFSAR), Revision 1, paragraph 13.1.2.2.16.6, Conduct of Operations, the overall responsibility for the protective tagging of equipment under. operations control is defined as that of the Shift Superintendent.
Paragraph 18.1.13 of the UFSAR which addresses TMI action item I.C.1 also references paragraph 13.1.2.2.16.6.
This violation is closed.
(0 pen) Violation 416/86-20-04. Surveillance procedure 06-IC-SZ51-SA-0001, Chlorine Detector Calibration, was inadequate in that it failed to specify chlorine detector valves Z51-F079 and Z51-F080 to be locked open as required by system piping and instrument drawing (P&ID) M-0049. On June 30, 1986 the valves were found open, but not locked.
On July 3, 1986 the inspectors found errors in the temporary alteration jumper log, which indicated non-compliance with Administrative Procedure (AP) 01-S-06-03, Control of Temporary Alterations.
As corrective action, the licensee revised the surveillance procedure such that it now requires the chlorine detector valves to be locked open. The AP was revised to delete the use of the temporary alteration jumper log because the log was redundant.
The information provided by the log is already required by the temporary alteration request form and log index.
In addition, an internal review of outstanding temporary alterations was performed. Minor deficiencies were identified and subsequently corrected. The above corrective actions were inspected to verify completion. The licensee also committed to verify that all instrument root valves are procedurally controlled in the Operations Department procedures, and upon completion to issue a :o1%y statement addressing the manipulation of instrumentation valves by personnel other than those in the Operations Department, by December 1986.
This is not complete as of the end of December 1986, however, the licensee issued letter AECM-86/0405 dated December 26, 1986, to NRC Region II requesting the commitment date to be changed to July 1987, due to the manpower resource impact caused by the extensive refueling outage conducted during September through November 1986.
The licensee is in the process of revising all t
applicable P& ids to identify all safety related instrument root valves.
This project is scheduled for completion by June 30, 1987. As the project is completed, Operations Department will be in a position to verify that all instrument root valves are procedurally controlled. This was discussed with the resident inspectors, and appears to be acceptable. This violation shall remain open pending completion and NRC followup inspection of thr. instrument root valve identification and control project, t
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(Closed) Violation 416/84-21-03.
Failure to implement Section 3.5.6 of
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MPL-Topical-1A which requires that designs be reviewed to assure that design characteristics can be controlled, inspected and tested.
Nuclear Plant Engineering Administrative Procedure (NPEAP)01-304, Performance of Design and Preparation of Design Change Packages, Revision 10, Section 7.0 of attachment 6 now requires that testing and inspection requirements be specified. NPEAP 01-306, Specifications, Revision 7, attachment 1, requires examination and testing requirements be specified.
This item is closed.
(Closed) Violation 416/84-21-04.
Failure to establish procedures for implementing the requirements of MPL-Topical-1A, paragraph 3.5.13, which requires that errors and deficiencies in the design process, including computer programs, that could adversely affect safety-related structures, systems or components be documented and corrective action taken to preclude repetition.
NPEAP 01-800 Nonconformance Reporting, paragraph 2.0, states that this procedure applies to all personnel assigned to NPE who become aware of any nonconformances or deficiency in items, services, activities, including errors in the design process or computer program, that could
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f affect structures, systems of components.
This item is closed.
(Closed) Violation 416/84-21-05.
Failure to establish adequate quality requirements and procurement prescribed by procedures appropriate to the circumstances.
The licensee has issued NPEAP 01-401, Procurement of material, Equipment and Services, Revision 4.
This NPEAP is applicable to safety related procurement processed by NPE.
This item is closed.
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4.
Unresolved Items *
One new unresolved item identified during this inspection is discussed in paragraph 8.
5.
Operational Safety Verification (7.1707)
The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control room such that it was visited at least daily when an inspector was on site. Observations included instrument readings, setpoints and recordings, status of operating 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 I
access controls.
This inspection activity included numerous informal discussions with operators and their supervisors.
- An unresolved item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
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l Weekly, when the inspectors were onsite, selected Engineered Safety Feature (ESF) systems were confirmed operable.
The confirmation is made by verifying the following: Accessible valve flow path alignment, power supply breaker and fuse status, major component leakage, lubrication, cooling and general condition, and instrumentation.
General plant tours were conducted on at least a biweekly basis. 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, radiation protection controls, physical security, problem identification systems, and containment isolation.
No violations or deviations were identified.
6.
Maintenance Observation (62703)
During the report period, the inspectors observed portions of the maintenance activities listed below. The observations included a review of the work documents for adequacy, adherence to procedure, proper tagouts, adherence to technical specifications, 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.
MWO M70107, Replacement of sheared stop screw on turbine generator hydraulic speed governor.
MWO 170142, Troubleshoot and reset source range monitor D retract permissive circuitry.
07-5-53-C41-4, Revision 3, Preventive maintenance on Standby Liquid Control System storage tank heater controls.
No violations or deviations were identified.
7.
Surveillance Observation (61726)
The inspectors observed the performance of portions of the surveillances 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 surveillances, 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.
06-IC-1C71-M-0002, Revision 22, Turbine Stop Valve Trip Fluid Low Pressure j
(RPS/EOCRPT) Functional Test.
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06-IC-1C71-M-1003, Revision 22, Turbine Control Valve Fast Closure (RPS/
E0CRPT) Functional Test.
06-RE-1C51-0-0001, Revision 24, Local Power Range Monitor Calibration.
06-RE-1J11-V-0001, Revision 30, Power Distribution Limits Verification.
No violations or deviations were identified.
8.
Engineered Safety Features System Walkdown (71710)
A complete walkdown was conducted on the accessible portions of the High Pressure Core Spray (HPCS) 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 conditions.
The overall condition and status of the HPCS system appeared satisfactory, however there were a few discrepancies as listed below:
The HPCS pump suction pressure gage, E22-PI-R001 was mounted with only one of the three required fasteners. One of the missing fasteners was lying on the gauge rack structure below the gauge, and the other was not found. This was subsequently corrected.
The condensate bypass pressure gauge, E22-PI-R002 is identified on Piping and Instrument Drawing (P&ID) M-1086, Revision 19 as E22-PI-R003, but the gauge label and operations procedures identify it as E22-PI-R002.
Some insulating pads and fasteners were laying loose near the Jockey Pump suction piping.
Two hangers in the HPCS pump room are incorrectly labeled "R0" for restrictive orifices. One hanger is on the 14 inch minimum flow and test return line, but the restrictive orifice is actually located in an adjacent room. The other hanger is on the 4 inch minimum flow line, but the orifice appears to be located about 10 feet downstream of the hanger, next to E22-F012.
Several HPCS motor operated valves are labeled on the adjacent insulation F004C but the P&ID identifies them as F004-C and the System Operating Instruction (501) refers to them as F004. The licensee explained that the
"-C" identified the power source (Division 3) and was not to be construed as the valve number.
It would appear appropriate to remove the
"C" from the labels to prevent confusion, because there are valves in the plant which have a letter suffix, i.e., E12-F0640.
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Step 4.1.2.h of SOI 04-1-01-E22-1, Revision 21, High Pressure Core Spray System, required the operators in the control room to verify all alarms cleared on panel 1H13-P601, section 16A, except reactor level 8 high, which is not expected to clear until reactor temperature exceeds approximately 350* F.
At the time of the walkdown, reactor temperature was 133* F, and the alarm annunciator was lit as expected, but the HPCS out of service (00SVC)
annunciator light was also on. The 00SVC light comes on whenever any HPCS system status lights are on, and as noted in step 4.1.2.e of the SOI, the HPCS line break status light stays illuminated until the reactor reaches a temperature of approximately 400* F.
Step 4.1.2.h failed to address this, and as such could not be followed verbatim to place the HPCS system in standby, The licensee subsequently corrected this discrepancy in
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l Revision 22 of the SOI.
Correction of the above minor discrepancies shall be tracked by Inspector Followup Item 416/86-41-01.
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The condensate bypass pressure gauge, E22-PI-R002 was apparently overranged by a pressure transient to the extent that the pointer was bent about 50 psi low, and the gauge was indicating approximately 800 psig, which is not appropriate for the existing conditions.
This line should be at no more than Jockey pump discharge pressure which is considerably less. The gauge has since been replaced and is indicating zero. The licensee was requested l
to identify the cause of the gauge being damaged, and if pressure transients l
were involved, to identify the boundaries and what effect this may have had on the integrity of the HPCS system piping and pressure boundary components.
This shall be Unresolved Item 416/86-41-02.
No violations or deviations were identified.
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Reportable Occurrences (90712 & 92700)
The below listed event reports were reviewed to determine if the information provided met the NRC reporting requirements.
The determination included i
adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance
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of each event. Addition 41 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.
The following License Event Reports (LERs) are closed:
LER No Event Date Event
- 86-003-01 January 22, 1986 Reactor Scram Resulting From Generator Load Reject.
- 86-033 September 23, 1985 Containment Cooling System Isolation on High Radiation.
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f LER No Event Date Event
- 86-032-02 September 16, 1986 Inadvertent Secondary Containment Isolation and ESF Actuations.
- 86-030 August 26, 1986 Breached Area of Control Room Boundary Exceeds Operating License Limit.
- 86-034 October 14,1986 Inadvertent Isolation of Shutdown Cooling.
- 86-035 October 21, 1986 Effluent Releases Exceeded TS Limits.
- 86-036 October 16, 1986 TS Time Limit Exceeded.
- 86-039 October 27, 1986 TS Time Limit Exceeded.
- 86-040 October 27, 1986 TS Time Limit Exceeded.
- 86-041 October 30, 1986 Auxiliary Building Isolation Switches Found in the
Bypass Position.
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- 86-043 November 6, 1986 Inadvertent Drywell Purge Compressor Start.
- 86-044 November 18, 1986 ESF Actuation Due to Incorrect Procedural Step.
The event of LER 86-003 is discussed in Report 416/86-04.
The event of LER 86-033 is discussed in Report 416/86-32.
The events of LER 86-032 resulted in IFI 416/86-32-03, violation 86-32-04 (first example), and violation 86-37-01 (first and second examples).
The event of LER 86-035 is discussed in Report 416/86-37.
The event of LER 86-039 is discussed in Report 416/86-37.
The event of LER 86-040 is discussed in Report 416/86-37.
The event of LER 86-043 is discussed in Report 416/86-37 under violation
416/86-37-01.
The event of LER 86-044 is discussed in paragraph 10 of this repor.
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LER 86-045, dated December 17, 1986 was reviewed and is considered unacceptable. The LER does not list similar events that have occurred such as those documented in LER 82-045 and 85-048.
Also, licensee followup actions with the responsible party and the safety significance were not addressed.
This event was discussed with the licensee on January 8,1987 and the licensee committed to submit an updated LER.
No violations or deviations were identified.
10. 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 action. The inspectors examined instrument recordings, computer printouts, operations journal entries, scram reports and had discussions with operations, maintenance and engineering support personnel as appropriate.
On November 18, 1986 while plant personnel were performing the Division 2 Diesel Generator (DG) 18 month functional test an electrical wire was removed from the isolation logic circuitry resulting in a Division 2 auxiliary building isolation and actuation of the Division 2 Standby Gas Treatment System (SGTS). A design change to the DG circuit had installed a permanent wire in the location where a temporary jumper was previously installed.
Because of the design change modification, the temporary jumper installation step in the test procedure had been deleted, but the licensee failed to delete the procedural step directing the technician to remove the temporary jumper. Consequently, when the technician removed the permanently installed wire from the circuit, the inadvertent actuation noted above occurred.
This event is another example of the events noted in violation 416/86-37-01 and will be tracked as part of the corrective action of
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violation 416/86-37-01.
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No violations or deviations were identified.
11.
Inspector Followup and Unresolved Items (92701)
By letter dated September 15, 1986 the licensee submitted a proposed change to Technical Specifications (TS) associated with the addition of an inboard isolation valve and two test connection valves to containment penetration 71B which is a one inch line associated with the post-accident sampling system.
A Material Non-Conformance Report (MNCR) was written on November 25, 1985, which stated the isolation barrier configuration ' for penetration 718 did not meet the criteria for a closed system outsido-containment, which was the configuration described in Table 6.2-49 the Final Safety Analysis Report (FSAR). Justification for continued operation was based on penetration 71B meeting the intent of Regulatory Guide (RG) 1.11.
By letter dated November 12, 1986 the WRC issued license amendment 24 in
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response to the licensee's September 15, 1986 letter. The Safety Evaluation to Amendment 24 expressed NRC concerns with the improper application of RG 1.11 as an alternate basis for meeting the requirements of General Design Criterion 56 (GDC 56).
It was noted that the plant was shutdown at that time in a refueling outage during which primary containment isolation was not required and the proposed additional isolation valves would be installed per DCP 86/4000 prior to startup from the refueling outage. Although the penetration in question, 71B, appears to have been corrected after a year's delay, the inspectors are concerned that other penetrations may have similar misapplications of RG 1.11.
The inspector has requested that the licensee determine if any other containment penetrations have been accepted based on similar applications of RG 1.11 to isolation valve requirements.
This will be Inspector Followup Item 416/86-41-03.
Followup on License Condition (LC) 2.C.(12).
LC 2.C.(12) requires, within 30 days after plant startup following the first refueling outage, that SERI shall comply with Items 1, 2 and 3 of Bulletin No. 79-26 and submit a written response to the NRC on item 3.
The subject of the issue is boron loss from BWR control blades. Plant startup from the first refueling outage occurred when the reactor became critical on November 30, 1986. On December 23, 1986 the licensee issued letter AECM 86/0403 to the NRC which reported compliance to item 1,2 and 3.
Although the LC appears to have been satisfied, the acceptability of SERI's actions in response to the bulletin shall be evaluated by NRC Region II prior to closecut of the bulletin.
(Closed) Inspector Followup Item 416/86-26-01.
Followup on correction of deficiencies found by the inspectors during the Division 2 Diesel Generator (DG) system walkdown.
Briefly, the discrepancies consisted of valves not properly labeled, instrument root valves not identified on the P&ID, component description errors in the system operating procedures, a deterio-ration of general cleanliness and condition of equipment, and apparent oil and air leaks. Upon reinspecting to verify that actions had been taken to correct the deficiencies, the inspectors were satisfied to the extent that this item may be closed; however, instrument root valves are not yet properly identified.
Completion of this task is currently being tracked under Violation 416/86-20-04.
See paragraph 3 of this inspection report.
(Closed) Inspector Followup Item 416/84-19-01, Standby Service Water (SSW)
system loop A flow.
Potentially Reportable Deficiency (PRD) 82-21 was initiated on April 15, 1982 due to a low flow condition through the SSW pump A when started in response to an automatic initiation of the Reactor Core Isolation Cooling (RCIC) system. The licensee initiated interim corrective actions due to the planned replacement of the existing SSW pumps with pumps of larger capacity.
This interim design was accepted and closed out in Report 416/84-19. Until a final design fix was incorporated, IFI 416/84-19-01 was initiated for followup.
By letter dated November 25, 1986 (AECM 86/0376) the licensee submitted a final report on PRD 82-21. The final corrective action required only the change out of the butterfly valve in the basin recirculation line with a globe valve due to the potential for valve damage from the increased pump total developed head. This change was accomplished during the first refueling outage.
PRD 82-21 is closed.
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(Closed) Unresolved Item 416/84-21-02.
Failure to perform a 10 CFR 50.59 review on Nuclear Plant Engineering Administrative Procedures (NPEAPs).
NPEAP 04-101, Revision 6, Authors Guide for Procedure Preparation, paragraph 7.4.2, now requires all cover sheets contain a safety evaluation appli-cability review statement.
The inspectors review of the NPEAPs verified that the applicability review statement was being complied with. This item is closed.
(Closed) Inspector Followup Item 416/83-38-03. Inclusion of ASME Section XI requirements into the surveillance program. The licensee's ASME Section XI program was inspected by Region II inspectors as documented in Reports 416/66-29 and 416/86-34. The licensee has submitted their Pump & Valve Inservice Inspection Program and related relief requests to the NRC for approval.
This item is closed.
12. Design, Design Changes and Facility Modifications (37700 & 37701)
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The inspectors have been conducting document reviews and hardware inspec-tions to ascertain that design changes and facility modifications associated with TS License Conditions were in conformance with the requirements of the facility license, TS, and 10 CFR 50.59. Part of the reviews were conducted during the three previous reporting periods and are documented in NRC Inspection Reports 416/86-32, 416/86-37, and 416/86-39 and if not closed are being tracked as Inspector Followup Items.
Those reviews that were continued and/or completed during this reporting period are documented in paragraph 11 above. The following additional Design Change Packages (DCPs)
were reviewed during this reporting period:
(Closed) DCP 83/3515, part of License Condition (LC) 2.C.(18): This DCP installed reactor pressure interlocks so that the injection valves on the Low Pressure Coolant Injection (LPCI) and the Low Pressure Core Spray (LPCS)
lines will not-automatically open when called upon by an accident signal when the reactor is above the design pressures of LPCI and LPCS systems.
Previously this interlock was only in effect during the test mode of the systems, and during an accident condition the systems were relying on swing check valves as a protective pressure boundary. The completed Design Change Implementation Package (DCIP) was reviewed by the inspectors and no deficiencies were identified.
The applicable TSs, operating procedures, surveillance procedures, and drawings were inspected to ensure that changes were implemented as a result of the DCP work completed, with satisfactory results.
Closure of this DCP partially satisfies LC 2.C.(18), which requires the above described interlock to be installed prior to startup from the first refueling outage.
This item is closed.
(Closed) DCP 84/4072, License Condition (LC) 2.C.(35):
During pre-operational testing of the drywell purge and Post-LOCA vacuum breakers, the vacuum breakers were unable to open or close properly due to the resistance of the contact type position switches. Six vacuum relief check valves were
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affected.
E61-F004A and B are in the Post-LOCA vacuum relief system, E61-F001A and B and E61-F002 A and B are in two redundant drywell purge systems.
LC 2.C.(35) requires that position indicators with redundant indication and alarm in the control room for the drywell purge and Post-LOCA vacuum breakers be provided prior to startup following the first refueling outage. The licensee incorporated DCP 84/4072 to implement LC 2.C.(35).
An NRC letter dated July 23, 1985 contained a Safety Evaluation Report accepting the licensee's design. The inspectors reviewed DCIP 84/4072 and identified the following discrepancies:
The Review, Approval and Implementation Record (RAIR), Section 3, asks if the change is an unreviewed safety question or Technical Specification change. The question is incorrectly marked no.
A TS change was involved.
The Retest Control Form for MWO P63786 lists surveillance 06-0P-1E61-Q-0007 as a retest that must be completed prior to submittal to operations for returning the system to operation.
No documentation exists to show 06-0P-1E61-Q0007 was accomplished prior to that time.
The NPE Information Verification data sheet, Form 707.1, lists the Responsible Engineer as B. Levenson and the Engineering Supervisor assigned Peer Concurrence Review as the verification method to be used.
Contrary to NPE Administrative Procedure 01-707, Certification of Information, B. Levenson also signed the Peer Concurrence Review block.
Alarm Response Instruction (ARI) 04-1-02-1H13-P870-10A-D2 directs operator actions to be taken if an alarm is received regarding the subject valves (Division 2). This instruction appears inadequate.
If any of the subject valves are found open, the valve must be restored to the closed position within one hour or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per TS 3.6.5.
The applicable ARI does not alert the operators to this TS requirement.
The licensee has taken the following actions on the noted discrepancies:
Procedure 01-S-16-1, Plant Design Changes and Modifications, divides the previous question into two parts. The RAIR form now asks if an unreviewed safety question is involved and separately asks if a TS change is involved.
The RAIR form for this DCP was corrected to reflect a yes answer.
A Records Supplement / Correction Form (Procedure 01-S-05-1, attachment V) was issued to reflect the accomplishment of Modification Special Test Instruc-tion (MSTI) 1E61-86-001-0-5 and the deletion of surveillance 06-0P-1E61-Q-0007. Modification Special Test Instruction (MSTI) 1E61-86-001-0-5 was accomplished in lieu of other specified retests but the DCP change documen-tation had been lost.
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NPE is clarifying Administrative Procedure 01-707, Certification of Informa-i tion, to ensure that independent certification is properly documented. The responsible engineer for preparing the document in question was not the same engineer that signed the Peer Concurrence Review.
ARI 04-1-02-1H13-P870-10A-D2 and 04-1-02-1H13-P870-4A-D2 were revised to reflect the applicability of TS 3.6.5 and the one hour LCO.
The items noted above consisted of minor paperwork errors but reflect poor attention to detail.
No further action is considered necessary.
(Closed) DCP 84/4084, LC 2.C.(33)(f):
This DCP modified the Automatic Depressurization System (ADS) logic to further automate the ADS system by providing automatic ADS initiation, if required, for events such as a break external to the drywell or a stuck open safety / relief valve. The modifica-tion also provided the capability to more easily inhibit ADS operation or operate multiple safety / relief valves in accordance with emergency procedure guidelines.
The inspectors witnessed portions of the work and inspected accessible portions of the completed modification, and reviewed the closed out DCIP for completeness. This DCP involved TS, drawing, system operating procedure, emergency procedure, alarm response instruction, and surveillance procedure changes.
The inspectors verified that the changes had been implemented except for one surveillance procedure, which was identified by a deferral notice in the DCIP. The deferral indicated that the particular surveillance requirements were satisfied by a Modification Special Test Instruction (MSTI) and that the Surveillance Program Tracking System (SPTS)
would ensure that the test is done when required again.
The inspectors t
follows! up and verified the test appeared on the SPTS. LC 2.C.(33)(f) was
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satisfied with the closeout of this DCP. The LC required the above modifi-cation and for SERI to provide, for NRC review, justification for the timer
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delay settings, revisions to the emergency procedures covering the use of
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l the manual inhibit switch and proposed TS surveillance procedures for the timer and switch. This information was submitted to the NRC on March 21 and May 30, 1986 (AECM 86/0082 and 86/0163 respectively). This item is closed.
(Closed) DCP 83/5024, LC 2.C.(33)(g):
This DCP provided the capability of l
continuously monitoring ADS air receiver pressure from the control room, l
with the appropriate alarm annunciators. The inspectors witnessed portions of the work and testing, inspected the accessible hardware installed and reviewed the DCIP for completeness. The DCP involved TS, drawing, operating
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procedure, and surveillance procedure changes and the publication of new j
l surveillance and alarm response procedures.
The inspectors followed up to
ensure they were properly implemented and found no problems. This completes the requirements of LC 2.C.(33)(g). This item is closed.
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'3.
IE Bulletin Followup (92703)
(Closed) IE Bulletin 86-01, Minimum Flow Logic Problems that Could Disable RHR Pumps.
By letter dated June 9, 1986 (AECM-86/0181) the licensee
responded to the subject bulletin. At GGNS the RHR A, B, and C and Low
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F
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Pressure Core Spray -(LPCS) pumps minimum flow piping. are equipped with-separate headers which discharge directly into the Suppression Pool. Each loop has independent flow indicating switches which provide signals to the
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corresponding minimum flow valve. Therefore, no single failure mode exists and the circumstances described in bulletin 86-01 do not apply to the RHR'
and LPCS systems at GGNS. This item is closed.
(Closed) IE Bulletin 86-03, Potential Failure of Multiple ECCS Pumps Due to
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a. Single Failure of Air Operated Valve in Minimum Flow Recirculation Line.
By letter dated October 27, 1986, (AECM-86/0333) the licensee responded to the subject bulletin.
In response to IE Bulletin 86-01 the licensee had investigated the possibility of a single failure vulnerability in the minimum flow recirculation line of any ECCS pump that could cause a failure of more than one ECCS train. The licensee's investigation determined that GGNS has no such single failure mode problem. This item is closed.
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