IR 05000416/1984046
| ML20112K141 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/03/1984 |
| From: | Burris S, Jape F, Mathis J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20112K144 | List: |
| References | |
| 50-416-84-46, IEB-84-03, IEB-84-3, IEIN-84-70, TEIN-84-70, NUDOCS 8501180486 | |
| Download: ML20112K141 (6) | |
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Report No.: 50-416/84-46 Licensee: Mississippi Power and Light Company Jackson, MS 39205 Docket No.: 50-416 License No.:
NPF-29 Facility Name: Grand Gulf 1 Inspectio onducted: November 13 - 16, 1984 Inspectors:
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P,4 J. L/. Mathis Date Signed
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25f(Y 5. P. Burris Date Signed Approved by:
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/ A[3/ h F. Jape, Section Chief C/ /
Date Signed'
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Engineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection entailed 52 inspector-hours on site in the areas of Power Level Plateau Data Review, Independent Inspection Effort and Followup on IEN 84-70 and IEB 84-03.
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Results: No violations or deviations were identified, i
l 8501180486 841210 PDR ADOCK 05000416 G
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O REPORT DETAILS 1.
Licensee Employees Contacted
- R. F. Rogers, Technical Assistance to General Manager
- J. L. Robertson, Operation Superintendent
- L. A. Daughtery, Compliance Superintendent
- J. D. Bailey, Compliance Coordinator
- D. Cupstid, Start-up Supervisor G. Ceasar, Licensing R. Patterson, Shift Technical Advisor R. V. Moomaw, I&C Superintendent J. Mueller, Mechanical Superintendent J. Buckalew, Mechanical Supervisor T. Miller, Engineer, Bechtel B. Levenson, Group Leader, Mechanical I&C Other licensee employees contacted included two operators, seven mechanics, three security force members and five office personnel.
NRC Resident Inspector
- J. Caldwell, Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on November 16, 1984, with those persons indicated in paragraph 1 above. The licensee acknowledged the inspection findings with no significant comments.
3.
Licensee Action on Previous Enforcement Matters Not inspected.
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Unresolved Items Unresolved items were not identified during this inspection.
5.
Power Level Plateau Data Review (72532)
The inspectors reviewed the completed data packages for start up testing during test condition 1 listed below.
The data packages were reviewed for compliance with the applicable portions of the Quality Assurance program section on start-up testing.
Data packages for the following tests were reviewed:
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SU-11-1, LPRM Calibration
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SU-12-1, APRM Calibration
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SU-19-1, Core Performance
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SU-29-1, Recirculation Flow Control System
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SU-71-1, Residual Heat Removal System
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The review verified that test changes were approved in accordance with administrative procedures, actions required by test changes had been completed, deficiencies and exceptions identified during testing have been resolved, including any retest requirements, and individual test steps and data sheets were properly completed.
The inspectors verified that the following requirements were met for the completed data packages:
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LPRM Calibration Program identification of the calibration, such as recording of
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data, time and calibration number in the LPRM calibration log book or equivalent Destination of all the individual amplifier input calibration
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currents required to produce a selected standard deflection on each meter All LPRM detector readings were obtained at the conditions
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existing prior to amplifier gaii adjustments Following the post gain adjustment P-1 calculation, the gain
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adjustment factors were verified within the established limits
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Determination following LPRM calibrations, that APRM settings were reviewed to assure they were within required Technical Specifica-tion limits b.
APRM Calibration Verified that the core was maintained at steady state operating
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conditions at the desired power level and recirculation flow during the calibration Verified that only one APRM channel per Reactor Protection System
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bus is bypassed at a time Verified that the APRMs were adjusted to read the calculated
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percent rated power Verified that the "as left" APRM readings had been properly
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recorded
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c.
Core Performance Verified that the Maximum Linear Hea. Generation Rate of all rods
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during steady-state conuitions did not exceed the limit specified by the plant Technical Specifications Verified that the steady-state Minimum Critical Power Ratio
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exceeded the minimum limit specified by the plant Technical Specifications Verified that the Maximum Average Linear Heat Generation Rate did
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not exceed the limits specified by plant Technical Specifications Verified core flow did not exceed its rated value
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Recirculation Flow Control System Verified that the decay ratio for each controlled mode of response
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q was less than or equal to 0.25
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Residual Heat Removal System
_ Verified that RCIC pump discharge flow rate recorded on data
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sheet 3 was approximately 414 gpm for two loop operation Verified that the temperature differences across the RHR side of
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the heat exchanger were greater or equal to 247.7"F l
Verified that the heat transfer rate for each RHR loop was greater
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than or equal to 112.8 x 106 BTU /hr two loop operation Verified that a transient response of any system related variable
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to any test input did not diverge In the areas inspected, no violations or deviations were identified.
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6.
Review and Witnessing of Maintenance Activities (62703, 92706)
During this inspection period, the inspectors reviewed selected maintenance activities in progress to verify that these activities were conducted in accordance with plant administrative controls and applicable NRC require-ments. This inspection consisted of both procedure review and witnessing of.
-work activity'in progress.
The mechanical maintenance procedures were reviewed to ensure that:
-Administrative controls have been estcblished for the review, approval
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and periodic updating'of maintenance procedures..
I Responsibilities have been assigned in writing to assure that main-
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tenance procedures will be reviewed, updated and approved as required.
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Plant maintenance personnel perfcrming work evolutions were qualified
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to perform the specific activity.
Required hold and inspection points were completed.
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Quality assurance personnel were available and on hand at the required
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hold points to verify that the work was performed in accordance with an approved procedure.
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Plant status was maintained current for all maintenance activities in progress which would have required the plant staff to place the plant in a safe shutdown condition within the specific amount time stated in the appropriate LCO.
This review included the following General Maintenance Instructions (GMI):
GMI-07-S-14-175, Replacement of HEPA Filters GMI-07-S-14-102, Replacement of Demisters GMI-07-S-14-172, Inspection of Filter Systems GMI-07-S-14-176, Inspection of Selected Dampers This inspection also included the witnessing of selected maintenance work in progress.
The inspectors reviewed the applicable plant administrative controls for conducting maintenance activities and witnessed the following maintenance activities to ensure that all applicable plant administrative controls and NRC requirements were met:
GMI-07-S-14-175, Replacement of HEPA Filters
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GMI-07-S-14-176, Inspection of Selected Dampers
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The inspectors verified that the maintenance met the following criteria:
Personnel performing the evaluation had a current copy of the GMI and
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were following it.
Any discrepancy identified during the performance of the activity'was
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identified.
Personnel performing the work activity were qualified to perform that
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particular part of the evolutions.
Maintenance activities performed during this inspection period appeared to be adequate with the following minor exception.
While performing GMI-07-S-14-176 for the first time, the licensee identified that the procedure did not completely satisfy all required maintenance requirements.
The procedure required that the valves be greased; however, the valves did
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not have a grease fitting.
The maintenance supervisor informed the
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inspectors that, although the manufacturer's technical manual had been used to write the procedure, it appeared that a more specific instruction was
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required to complete all necessary maintenance for these valves.
The maintenance group stopped all maintenance activities for this procedure and informed the inspectors that the licensee would evaluate the adequacy of the current procedure, and would write a new procedure for further maintenance work.
The licensee's assessment and corrective actions would be examined during a future inspection.
No violations or deviations were identified in the areas inspected.
7.
Follow-up on IEN and IEB (92703)
a.
IEN Reliance on Water Level Instrumentation with a Common Reference Leg Followup on IEN 84-70 Reliance on Water Level Instrumentation with a Common Reference Leg
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The inspectors reviewed the licensees evaluation for IEN 84-70, Re-liance on Water Level Instrumentation with A common Reference Leg.
The inspector interviewed the operations I&C Superintendent to discuss the applicability of this Bulletin for Grand Gulf. The licensee inform-ed informed the inspectors that Grand Gulf Unit 1 did have instrumenta-tion which shared a common reference leg and that the Nuclear Plant Engineering Group (NPE) was evaluating the Bulletin.
The inspectors discussed this issue with NPE and were informed that Grand Gulf was still evaluating the applicability of IEN-84-70.
The inspectors informed plant management that the NRC would review the final evalua-tion in a future inspection.
b.
IEB Refueling Cavity Seal Ring In addition to above, the inspectors followed-up on IEB-84-03, Refuel-ing Cavity Seal Ring.
On August 24, 1984, NRC distributed IEB-84-03 concerning the failure of a pneuma-seal at the Haddam Neck Nuclear Plant.
Actions by MP&L were to be taken prior to beginning refueling or within 90 days of receipt of this Bulletin, whichever is sooner.
During the inspection period, the inspectors received a copy of the licensee's response to IEB-84-03 and held discussions with NPE-Civil Section Staff.
At Grand Gulf Nuclear Station (GGNS), with the drywell head removed for refueling operations, the water barrier around the reactor cavity is maintained with reactor refueling seal bellows. The seal bellows were procured from Pathway Bellows, Inc., per the design requirements delineated in Bechtel Specification 9645-C-173.0.
This specification
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requires that the seal bellows be designed in accordance with applic-i
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able ASME and ASTM Codes to maintain their integrity during accident i
and seismic conditions. - The assembly was fabricated using welded stainless steel material and contains no rubbers or inflatable seals that could be susceptible to failure. Since the GGNS refueling cavity i
l water barrier does not. contain inflatable seals as in the subject l
event, IEB 84-03 is not applicable to GGNS.
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