IR 05000321/1989015

Insp Repts 50-321/89-15 & 50-366/89-15 on 890724-28.No Violations or Deviations Noted.Major Areas Inspected:Snubber Surveillance Program,Nrc Bulletin 80-11,masonry Wall Design & Licensee Action on Previous Insp Findings

ML20246P506
Person / Time
Site:	Hatch
Issue date:	08/16/1989
From:	Belisle G, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20246P496	List: IR 05000321/1989015 ML20246P492
References
50-321-89-15, 50-366-89-15, IEB-80-11, NUDOCS 8909110135
Download: ML20246P506 (7)

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UNITED STATES

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o NUCLEAR REGULATORY COMMISslON REGION li r,

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. MARIETTA STREET, * E t ATLAN1 A GEORGI A 30323

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- Re' port Nos.:- 50-321/89-15 and 50-366/89-15

. Licensee: Georgia Power Company P. 0.' Box 1295 Birmingham, AL 35201 Docket Nos.: 50-321 and 50-366 License Nos.: DPR-57 and NPF-5

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' Facility Name: Hatch 1 and 2 Inspection Conducted: . July 24-28, 1989 Inspector: n w ./ . /- f/$

J. J. Lenah n H~

/ Date' Signed

Approved By: -

<W p /2/3/{,f G. A. Belisle,~ Chief 'Date' Signed Test Programs.Section Engineering Branch -

Division of Reactor Safety SUMMARY Scope:-

This routine, unannounced inspection was conducted in the areas of the snubber surveillance program, IE Bulletin 80-11, Masonry Wall Design, and licensee action on-previous inspection finding Results:

In the areas inspected, violations or deviations were not identifie One minor weakness was identified in the lack of details in the emergency response procedure regarding connecting the backup air supply to the transfer canal seals when normal air supply is interrupted (paragraph 4.b).

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8909110135 890822 PDR ADOCK 05000321 o r- DC

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i REPORT DETAILS Persons Contacted Licensee Employees

• E. Burkett, Acting Manager, Engineering Support

• P. Fornel, Maintenance Manager

• 0. Fraser, Site Quality Assurance (QA) Manager

• J. Hammonds, Supervisor, Nuclear Safety and Compliance R. Keck,' Supervisor, Engineering Support C. Moore, Assistant General Manager, Plant Support

• G. O'Donnell, Instrumentation and Control (I&C) Supervisor

• S. Tipps, Manager, Nuclear Safety and Compliance Other licensee employees contacted during this inspection included engineers, operators, technicians, and administrative personne Other' 0rganations R. Dewberry, Southern Company Services, Modification Group Supervisor

'J. Norman, Mechanical Engineer, Bechtel NRC Resident Inspectors

• R. Musser

• Attended exit interview Snubber Surveillance Program, Units 1 and 2 (70370)

The inspector examined procedures and quality records related to the snubber surveillance program and inspected snubbers on safety-related piping systems. Acceptance criteria examined by the inspectors appear in Unit 1 Technical Specification 3/4.6.L and Unit 2 Technical Specification 3/4. Review of Snubber Surveillance Procedures The inspector examined the following procedures which control snubber surveillance activities:

52SV-SUV-001-IS and 52SV-SUV-001-2S, Hydraulic Shock and Sway Arrestor Inspection and Functional Test 52SV-SUV-004-IS and 52-SUV-004-25, Inspection and Testing Pacific Scientific Mechanical Snubber

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52GM-MME-025-0, Removal of Pipe Hangers - Restraints and Supports 52CM-MME-003-0S, Bergen-Patterson Hydraulic Snubber Rebuild Inspection of Snubbers The inspector performed a visual inspection of the following snubbers:

IE11 RHR 193*, 213, 216, 224 A, 224 B, 244 A, 250 A, 250 B, 251, 252*, 254, 255, and 306 A on the Unit I residual heat removal (RHR) system 1E21 CSH 57 on the Unit I core spray system 1E41 HPSEH 85 on the Unit I high pressure coolant injection (HPCI) system IT48 CPH 6A* and 6B* on the Unit I containment vent and purge system 2E11 RHR 106*, 112*, 235*, 250, 251 A, 251 B, 252, 268, 269 A, 269 B, 290 A*, 290 B*, - 292 A*, 292 B*, 295 A*, 295 B*, 297 A* ,

and 297 B* on the Unit 1 RHR system 2E 21 CSR 37A*, 37B*, 40, 53A, 538, 72A, 72B, 706*, and 707* on the Unit 2 core spray system 2 T48 CPH 54* and 727* on the containment purge vent system 2 E51 CR0 110* on the Unit 2 RCIC system

• Indicates mechanical snubbe All other snubbers listed are hydrauli During the above inspection, the inspector verified that the snubbers were not damaged, that attachment to the supporting structure (hanger) and piping was secure, and for hydraulic snubbers, that the reservoir levels were acceptabl Review of Quality Records The inspector examined quality records documenting visual inspection of Unit I accessible and non-accessible snubbers performed in September - October,198 Within the areas inspected, no violations or deviations were identifie . _ -__ _ _ _ _- _ - _____ _ _ -_____ _ _- -

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3. (Closed) IE Bulletin 80-11, Masonry Wall Design (25537)

The licensee responded to IEB 80-11 in letters to NRC Region 11 dated July 3,1980, and November 4,1980. These responses reported that the design re-analysis required by IEB 80-11 disclosed that approximately nine walls required modifications to meet seismic design requirements. The problem regarding the overstressed masonry walls was also reported to NRC as Licensee Event Report number 50-321/1980-115 on November 1, 1980. The licensee notified NRC Region II on April 16, 1981, that the computer program used to perform the seismic analysis of the masonry walls was found to contain some nonconservative assumptions. This problem was reported as LER 50-321/1981-031. As a result of the reanalysis performed with the corrected computer program, the licensee determined that one additional wall required modifications to meet seismic design criteri The licensee submitted additional information regarding their masonry wall design criteria to the NRC Office of Nuclear Reactor Regulation (NRR) ir; a letter ' dated April 20, 198 Final acceptance of the licensee's design methodology for analysis of masonry walls was documented in a Technical Evaluation Report dated June 25, 1982, which was transmitted to the licensee by the NRC in a letter dated August 16, 198 The licensee prepared Design Change Request (DCR) 81-10 to implement modifications to the ten overstressed masonry wall The wall modification included complete removal of two walls, and addition of steel supports to reinforce the walls. The inspector examined the modifications completed for eight walls during several inspections and documented the inspection results in NRC Inspection Report Numbers 50-321/83-24, 50-366/83-25; and 50-321,366/84-05. During the current inspection the inspector reviewed DCR 81-10, which was closed out on August 20, 198 The inspector reviewed As Built Notices83-595 and 86-747 which document the as-built wall modifications and Field Deviation Report numbers 81-10-01 through 8. The inspector verified that the two remaining overstressed walls, numbers C 164-4 and C 139-39 had been removed as per licensee commitment Based on review of the licensee's actions to complete IE Bulletin 80-11 requirements which were reviewed during several inspections and documented in NRC Inspection Report Numbers 50-321, 366/81-02; 50-321, 366/81-11; 50-321/83-24, 50-366/83-25; and 50-321, 366/84-05, and review and acceptance of the licensee's design evaluation methodology by the NRC Office of Nuclear Reactor Regulation, IE Bulletin 80-11 is close Within the area inspected, no violations or deviations were identifie . Action on Previous Inspection Findings (92701, 92702) (Closed) Unresolved Item 321, 366/84-05-02, Timely Closeout of Design Change Requests. During previous inspections, the inspector reviewed the status of Design Change Requests (DCRs) and determined that the DCRs may not have been closed out within a reasonable time after the l

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actual physical modifications to the plant had been completed. The licensee formed a task force -in 1984-85 to address 'DCR closeout Examples: of items that need to be. resolved prior to DCR closeout are;

' verifications by have as-built drawings ) knowledgeable been completed, engineers that all.affr.cted that as-built notices ( procedures have been revised, that nonconformance reports. and field deviatio reports have been resolved, and that various .other documents, equipment indexes have been. updated. As: a result of the efforts of

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- the:DCR closeout task force, the backlog of open DCRs .was _ reduce Since the task force was disbanded in 1985, DCR implementationLand

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closecut has been the responsibility of the Modifications- Group within the Engineering Support Department. The inspector' examined Procedure '42 EN-ENG-001-05, Engineering Service Procedure, which provides instruction for initiating, controlling, and. documenting DCR Section. 7.8 of this procedure specifically addresses .DCR closeou The inspector interviewed personnel within the modifications group - and reviewed the DCR status report. .. The inspector also examined selected DCRs which have been closed ( DCRs 81-10, 87-99,87-100). Based on these discussions and review, the inspectot determined that the licensee has initiated efforts m which will result in timely DCR closeout. DCRs which are_ ready for closeout have been identified on a. prioritized list to expedite .

closecut. The inspector determined that the licensee- complied with -

procedural and NRC requirements in closeout of DCR b.- (0 pen) Unresolved Item 321, 366/86-41-02, Design of Transfer Canal Seals and the Seal Leak Detection Syste When .the loss of 141,000 gallons of fuel pool water occurred on December 2-3, 1986, the leakage was not detected because it occurred on the Unit 1 side of the reactor building seismic gap 'where no leak detection system existed. The inspector examined DCR 87-99 which -

provided for installing a sheet metal channel to collect leakage from the transfer canal seals and modified the Unit 2 leak detection system to detect leakage on the Unit I side of the gap. The inspector also examined ABNs87-663, 88-45, 46, and 48 which provided as-built details of the completed modifications to the leakage detection system. The December 2-3, 1986, event occurred as a result of losing air to the inflatable seals in the transfer cana The inspector examined: DCR 87-100 which provided a redundant air supply for the seals and provided annunicators in the control room to alarm in the event of losing the normal air supply to the seals. The

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modification involved interconnecting the Unit I and Unit 2 service air: systems to provide two sources of air to the seals (i.e., with loss of service air on one unit, the service air system on the other unit would provide a backup air supply to the seals), addition of a receiver tank on the Unit I and Unit 2 service air supply lines to the transfer canal seals, and installation of check valves in the u - _ . n . _ _ - _ _ - _ _ _ _ _ _ - _ _ - _ _ _ _ - _

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service air systems to prevent back leakage of air from the receiver tanks and canal seals in case of failure of one or both service air systems. The normal pressure in the service air supply lines to the transfer canal seals is 100 psi Two redundant pressure switches were installed in both the Unit I and Unit 2 service air supply lines which alarm when the service air pressure drops below 75 psig. A pressure regulator reduces the pressure being supplied to inflate the transfer canal seals to 35 psi Two redundant pressure switches were installed in both the Unit I and 2 service air supply lines downstream of the pressure regulators to alarm when the air pressure in the transfer canal seals drops below 22 psig. The inspector examined the modifications to the transfer canal seal air supply system and. verified that the system was installed in accordance with DCR 87-100, and the as-built conditions shown on ABNs87-734, 88-43, 44, and 4 The inspector examined procedure 34AR-654-027-IN, Fuel Pool Bladder Low Air Pressure, which specifies the reactor operator's response to the low pressure alarm on control room panels H11-P654 and 2H11-P65 The inspector examined the results of the post-modification tests of the annunciators which were conducted in accordance with Special Purpose Procedure 42 SD-111187-00-1-0N, Functional Test for DCR 87-100 Air Supply to Spent Fuel Pool Transfer Canal Seals. Discussions with the reactor operators and review of the annunciator response procedure disclosed that a backup air supply is to be installed by the licensee's I&C technicians at the direction of the reactor operator Discussion with I&C personnel disclosed that they were some what familiar with this requirement, but that they did not have a specific procedure to provide the backup air supply. Procedure 34 AR-654-027-IN lacks detail regarding the backup air supply installation, although DCR 87-100 provided for storing a Nitrogen Bottle on the refueling floor near the transfer canal to provide the backup air supply. The inspector walked down the air supply system for the transfer canal seals with licensee I&C personnel and discussed the hookup of the backup air supply. After examining the system, the ISC personnel stated that the Nitrogen bottle would be connected to the quick connects installed on the distribution panel supplying air to the six transfer canal seal They stated that operations personnel would then open the bottle and adjust the regulator so that pressure to the seals is as stated in procedure 34 AR-654-027-I Although the I&C personnel did not have a detailed procedure indicating the requirements for installing the backup air supply system, they were able to adequately demonstrate to the inspector that they were knowledgeable of the system to install the backup air supply within 30 minutes to one hour after notification by the reactor operator This is adequate since the receiver tanks hold a 24 to 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> air supply for the transfer canal seals. The inspector questi 9d licensee personnel regarding the need fcr a more detailed procedure showing the installation of the backup air supply which would assist less experienced I&C personnel with its installatio Licensee management personnel stated they would review this problem to determine if the issuance of a procedure j

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is necessary. Unresolved Item 321, 366/86-41-02 remains open pending review by the NRC Office of Nuclear Reactor Regulation of the adequacy of the design of the originally installed canal seals and leak detection system, (Closed) Violation 321, 366/87-14-01: Failure to Have Written Procedure to Cover Inspection of Mechanical Snubbers After Severe Dynamic Event. The licensee's corrective actions for this violation are stated in their August 19, 1987, response to the NRC. The licensee deleted the mechanical snubber inspection procedures and replaced them with new revised procedures 52SV-SUV-004-IS and 52SV-SUV-004-2S. The inspector examined the procedures and verified that they contained requirements for inspecting mechanical snubbers after a severe dynamic event, and also complied with other existing Technical Specification requirement . Exit Interview The inspection scope and results were summarized on July 28, 1989, with those persons indicated in paragraph 1. The inspector described the areas inspected and discussed in detail the inspection results listed abov Proprietary information is not contained in this report. Dissenting comments were not received from the licensee.

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