IR 05000282/1989008

Partially Withheld Insp Repts 50-282/89-08 & 50-306/89-08 on 890305-0415 (Ref 10CFR2.790 & 73.21).Violations Noted.Major Areas Inspected:Operational Safety,Maint,Security,Refueling & Temporary Instructions

ML20247L428
Person / Time
Site:	Prairie Island
Issue date:	05/02/1989
From:	Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20247L389	List: IR 05000282/1989008 ML20247L385 ML20247L419
References
50-282-89-08, 50-282-89-8, 50-306-89-08, 50-306-89-8, GL-87-12, GL-88-17, IEB-88-008, IEB-88-8, IEIN-87-023, IEIN-87-059, IEIN-87-23, IEIN-87-59, IEIN-88-001, IEIN-88-051, IEIN-88-055, IEIN-88-067, IEIN-88-1, IEIN-88-51, IEIN-88-55, IEIN-88-67, NUDOCS 8906020165
Download: ML20247L428 (13)

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'U. S. NUCLEAR REGULAT0kY COMMISSION

REGION III

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Reports No. 50-282/89008(DRP);.50-306/89008(ORP) R

. Docket Nos. 50-282; 50-306 Licenses No. DPR-42;.DPR-60 l

Licensee: Northern. States Power Company J 414 Nicollet Mall  ;

Minneapolis, MN 55401 L Facility Name: Prairie Island Nuclear Generating Plant Inspection At: Prairie Island Site, Red Wing, MN i Inspection Conducted: March 5.through April 15', 1989 Inspectors: J. E. Hard T. J. O'Connor

! Approved By: B. Burgess, Chief Reactor Projects Section 2A E[d///

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-l Inspection Summary Inspection on March 5 through April 15, 1989 (Reports No. 50-282/89008(DRP);-

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No. 50-306/89008(DRP))

Areas Inspected: Routine unannounced inspection by resident inspectors of 4

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previous inspection findings, plant operational safety, maintenance, surveillance, facility modifications, security, refueling, temporary instructions, and followup of LERs and Information Notice Results: During this. inspection period, Unit 1 operated continuously at 100%

power, except for power reductions associated with axial offset calibration testing, and quarterly turbine generator valve testing. Additionally, a power

. reduction was commenced on March 22, 1989, as a result of refueling water storage tank (RWST) boron analysis indicating boron concentration being significantly less than technical specification (TS) limits. Subsequent analysis verified the concentration was within TS limits and the power reduction was terminate RCS _ radio chemistry has indicated the presence of a failed fuel ro Activity levels, although higher than normal, are still less than one percent of TS

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l Attachment Contains SAFEGUfmLS I R OITATION Upon Sol a d 51 W G

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8906020165 890504 Page ID Dooo:1 trolled ADOCK0500gg2

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Unit'2 entered the inspection period at 90 percent power. At 10:15 p.m. on

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March 28, 1989, Unit 2 came off line for a scheduled maintenance and refueling outag The Unit 2 outage has progressed smoothly and has included 100% eddy current testing of the steam generators which resulted in a total of eight tubes'being plugge In general, the plant continues to be well operated, as noted by no reactor trips since July 1987, and few personnel error Of the 9 areas inspected, two violations of NRC requirements were identifie Attachment Contains St.FEGO.Anas INFORMATION Upon Sc212 i10' 5hl" page Is Docera roll ed

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DETAILS o Persons Contacted i Licensee Employees

1. L. Eliason, General Manager, Nuclear Plants

1. • E. Watzl, Plant Manager D. Mendele, General Superintendent, Engineering and Radiation Protection R. Lindsey, Assistant to the Plant Manager

1. M. Sellman, General Superintendent, Operations G. Lenertz, General Superintendent, Maintenance

'D.'Schuelke, Superintendent, Radiation Protection G. Miller, Superintendent, Operations Engineering P. Kamman, Superintendent, Nuclear Operations QA

1. K. Beadell, Superintendent, Technical Engineering

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1. • M. Klee, Superintendent, Quality Engineering R. Conklin, Supervisor, Security and Services D. Vincent, Project Manager,' Nuclear Engineering and Construction

'D. Musolf, Manager Nuclear Support Services-J.' Goldsmith, Superintendent, Nuclear Technical Services

• A. Hunstad, Staff Engineer T. Amundson, Superintendent Training

1. • A. Smith, General Superintendent, Planning and. Services

• J. Leveille, Senior Nuclear Safety / Technical Services Engineer

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E. Eckholt, Senior Nuclear Safety / Technical Services Engineer A.'Vukmir, Site Services Representative, Westinghouse Electric Cor G. Ortler, Manager Corporate Security NRC Representatives

1. A. Davis, Regional Administrator

1. W. Forney, Deputy Director, Division of Reactor Projects

1. J. Hickey, Acting Director, Division of Radiation Safety and Safeguards

1. B. Burgess, Chief, Reactor Projects Section 2A

1. I. Jackiw, Chief, Reactor Projects Section 28

1. W.' Shafer, Acting Deputy Director, Division of Reactor Safety .

The inspectors interviewed other licensee employees, including members of the technical and engineering staffs, shift supervisors, reactor and auxiliary operators, QA personnel, shift technical advisors, and shift manager * Denotes those present at the exit interview of April 17, 198 # Denotes those present at the management meeting of April 11, 198 ,

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^ 2. ' Licensee Event Report Followup (92700)

(0 pen) 282/89002: Vital Area Keys Lost for a Short Tim On March 5, 1989, while performing a tour of out plant equipment, the plant attendant's vehicle became stuck in the snow. During efforts to free the vehicle, the plant attendant lost his key ring which contained vital area keys. The keys were located approximately four hours late Prior to the location of the keys, the area was kept under surveillanc (0 pen) 282/89003: Unplanned Start of Emergency Diesel Generator D At 10:50 p.m. CDT on April 13, 1989, emergency diesel generator D1 experienced an unplanned auto start. The auto start occurred as a result of incorrect changes to the surveillance procedure, and oversight during the review proces . Operational Safety Verification (71707)

Unit 1 operated continuously at 100% power with power reductions associated with axial offset calibration testing and quarterly turbine generator valve testing. Additionally, a power reduction was commenced on March 22, 1989 as a result of refueling water storage tank (RWST)

boron analysis indicating boron concentration being significantly less than TS limits. Subsequent analysis verified that the concentration was within T.S. limits and the power reduction was terminated. Unit 2 entered the inspection period at 90% power and came off line on March 28, 1989, for a scheduled maintenance and refueling outage. The inspector observed control room _ operations, reviewed applicable logs, conducted discussions with control room operators and observed shift turnovers. The inspector verified operability of selected emergency systems, reviewed equipment control records, and verified the proper return to service of affected component Tours of the auxiliary building, turbine building and external areas of the plant were conducted to observe plant equipment conditions, including potential fire hazards, and to verify that maintenance work requests had been initiated for the equipment in need of maintenanc As documented in Inspection Reports No. 50-282/89002(DRP) and No. 50-306/89002(DRP), the licensee's experience with the radiation monitoring had not been good during the inspection period of December 18, 1988 through January 28, 1988, noting that 2R11 (containment air particulate monitor, train A) was out of service for eight days starting January 17, 198 During this inspection period 2R11 has been removed from service for an additional four days. The inspectors will continue to monitor the licensee's activities towards improving the operability of containment air particulate monitors, trains A and In Inspection Reports No. 50-282/89003(DRP) and No. 50-306/89003(DRP),

the licensee'was noted to have instituted measures to significantly minimize the number of individuals in the "at the controls" area of the

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shift during the outage have minimized the improvements to be realized on the Unit 2 side of-the control roo The inspectors anticipate that the full benefits of the policy will be realized when the unit is returned-to operation and during subsequent outage 'During the course of performing I&C Surveillance Test Procedure SP 1032A, Safeguards Logic Test, Rev. 10, the plant experienced a momentary rod block / turbine' runback. The technicians were able to recreate the momentary rod block / turbine runback. Additional trouble shooting will be done in order to fully correct the proble On March 27, 1989 the resident inspection staff.noted an unescorted visitor in the plant protected area. This is an apparent violation of the plant security plan. See attachment to this report for detail (Unclassified Safeguards Information).

' Maintenance Observation (62703, 60710)

Routine, preventive, and corrective maintenance activities were observed to ascertain that.they were conducted.in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications. The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from servicej, approvals ,

were obtained prior to initiating the work, activities were accomplished 1 using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, radiological ,

controls were implemented, and fire prevention controls were implemente l Portions of the following maintenance activities were observed during the inspection period:

Disassembly of the incore detectors Removal / Installation of seal table ferrules-l Installation of the containment storage platform l Installation of additional pipe / valve supports in response to l revised center of gravity calculations L Training provided to boilermakers for the removal of

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Westinghouse steam generator mechanical plugs Preventive maintenance on 4160v breakers including the reactor trip breakers Balancing of the 23 fan cooling unit Installation of split ring canopy seal on the reactor vessel head i Replacement of defective reactor vessel thermocouple connectors Sludge lancing of the 21/22 steam generators

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Change out of Type 410 bolting material used as retaining block studs on accumulator check valves (Reference Information Notice 88-85: Broken Retaining Block Studs on Anchor Darling Check Valves) 2SI-6-1, 2, 3, and In order for the licensee to isolate check valves 2SI-6-2 and 2SI-6-1, the operators realigned the residual heat removal system from Loop B cold leg injection to the reactor vessel injection flow path. The inspectors monitored the flow path transfer noting the precautions, procedure review and anticipated results by the operators. This evolution occurred while the plant was at mid loop operation. The licensee appropriately assigned '

a full time operator to monitor RHR pump pressure and flow and RCS level because this flow path did not have annunciator capabilities for RHR low  ;

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No violations or deviations were identifie . Surveillance (61726, 60710)

The inspector witnessed portions of surveillance testing of safety-related systems and components. The inspection included verifying that the tests were scheduled and performed within Technical Specification requirements by observing that procedures were being followed by qualified operators, that Limiting Conditions for Operation (LCOs) were not violated, that system and equipment restoration was completed, and that test results were acceptable to test and Technical Specification requirement Portions of the following surveillance were observed / reviewed during the inspection period:

SP 2083 Response to Safeguard Signal Test, Rev. 13 SP 2036 Turbine Overspeed Trip Test and Setpoint Verification, Rev. 13 SP 2092A Safety Injection Check Valve Test (Reactor Vessel Head Off), Rev. 9 SP 2615B Bus 26 Voltage Restoration, Rev. 3 SP 2098 21, 22 Station Battery Load Test, Rev.10 SP 215 Main Steam Safety Valve Test (Hot), Rev. 1 SP 2624 E. H. Overspeed Trip Setpoint Verification, Rev. 2 SP 2646 Steam Flush of Atmospheric Steam Dump Valves, Rev. O Prior to the performance of SP 2098, the licensee replaced cell 17/18 with one from storage. Cell 17/18 exhibited the greatest cover bowing and cracking as discussed in Inspection Reports No. 50-282/89003; i No. 50-306/89003(DRP). Cell 17/18 was transferred to the licensee's

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service facility where it will undergo destructive examinatio The inspectors will continue to monitor the licensee's activities in this are No violations or deviations were identifie _ _ _ _ _ _ _ _ _ _ -

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. ' Facility Modifications (37701, 60710)

As noted in Inspection Reports No. 50-282/89003; 50-306/89003(DRP), the Unit 2 steam Generator Feedwater Control Systen, is being replaced with one of a digital desig Amendment Nos. 87 and 80 to Facility Operating Licenses No. DPR-42 and DPR-60- Elimination of Steam /Feedwater Mismatch Flow and Low Feedwater Reactor Trips, was approved via NRC letter dated April 3, 1989, contingent upon undergoing an acceptable audit of the verification and validation program dealing with the separation requirements of IEEE 27 During the course of the outage, the inspectors witnessed the modifications to the protection racks, main control boards and the installation of additional pressure, temperature and flow transmitter Installations were monitored for compliance to wiring diagram QC verification and equipment operability testing was also monitore Individuals installing the equipment and performing checkout tests were familiar with the installation requirements and the logic behind the particular tests. During the course of the installation, the inspector noted that all field landed cables were attached to terminal strips with lugs applied with a calibrated crimper. The licensee was asked for information supporting this practice and whether point to point wiring completely internal to the protection cabinets also had to be installed with lugs applied with a calibrated crimper. The licensee has contacted the architect engineer who will contact the equipment vendor concerning this questicn. The inspectors noted the licensee has made extensive efforts to minimize the distractions of noise and sight to the control room operators and noted also the expedition with which the main control boards were modifie The inspectors will continue to monitor the testing and system performance of the steam generator feedwater cortrol syste Additionally, the inspectors monitored the installation of a new divider plate and manways on the 22 component cooling heat exchanger (CCHX).

Initial attempts to perform a hydrostatic test were unsuccessful due to leakage at the north end of the heat exchanger. The licensee determined that the welding of the new divider plate caused the shell to shrink which prevented the head from obtaining the proper fit-up. After correcting the interference problem, the licensee completed a successful hydrostatic test on the 22 CCH No violations or deviations were identifie . Refueling (60710)

The Unit 2 refueling and maintenance outage has progressed smoothly and according to a well planned schedule. Early delays associated with test equipment problems for the main steam safety valves were quickly minimize _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _

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Other activities which caused minor schedule delays included a damaged dummy fue1 assembly used for fuel handling equipment checkout and the need for a second hydrostatic test on the 22 CCH The first major task accomplished by the licensee was the 100 percent' l eddy current test of the 21 and 22 steam generators (SG). The contractor i provided a detailed explanation of.the equipment, processes and data interpretation using the MIZ-18' system. With the improved equipment the

. licensee was able to obtain' reliable data. Eddy current testing prompted .l the plugging of 3 tubes in the 21 SG and 5 tubes in the 22 S ].

As a result of a failure of a Westinghouse mechanically expanded steam

. generator tube plug at another plant, the licensee performed a safety evaluation which justified leaving in place those tube plugs from a questionable heat of such plugs. The plug failure mechanism was attributed to improper heat treatment which permitted circumferential primary water stress corrosion cracking (PWSCC). The licensee has a total of 24 such plugs installed on the hot leg side of the Unit 2 steam generator Major points in the safety evaluation include: visual inspection of subject tube plugs, lower hot leg temperatures than those of the plant where the failure occurred, and results of an examination of tube plugs pulled from a similar facility which yielded no indications of circumferential PWSC Based on the above, the licensee has determined =

that the probability of analyzed accident occurring has not increased nor could this situation result in a previously unanalyzed accident. Further detailed information may be found in the licensee's letter to the T

director of Nuclear Reactor Regulation dated April 11, 198 During the course of the SG work, Unit 2 was operated at mid-loop. (See Paragraph 8.8). The inspectors confirmed that the approved operating procedures were followed. Licensed operators were cognizant of plant conditions, necessary responses to perturbations and the time frame within which these actions needed to occu Major evolutions conducted by the licensee, such as the performance of Surveillance Procedure SP 2083, Response to Safeguards Signal Test, Rev.13, and the installation of the reactor vessel head, were preceded by p anning meetings with participants including radiation protection technicians. The results of such planning permitted those evolutions to be completed in an expedient manne While performing checks of equipment to be used in conjunction with the core shuffle and fuel loading / unloading, the licensee discovered that the dummy fuel assembly used for testing fuel handling equipment had a bent dummy control rod, which prevented the assembly from being removed from the fuel transfer cart through normal methods. The licensee was able to extract the dummy fuel assembly with " grappling" hooks. Close examination revealed tnat the dummy control 'od had not been fully inserted into the dummy fuel assembly and had struck the overhang above the transfer tube i while being raised and lowered in the refueling cavity. After extracting L

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the dummy control rod, the licensee satisfactorily completed the fuel transfer equipment checkou .l

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'An unrelated incident occurred while moving a new fuel assembly to the transfer cart. During this incident, a new fuel element was picked up by its holddown springs rather than the top nozzle. The assembly was lowered back into the rack and correctly connected. Subsequent examination of the assembly and discussions with the fuel vendor determined that the assembly was undamaged and suitable for us During the course of the outage, the inspectors verified that: the licensee closely monitored appropriate jobs for radiation controls; the requirements specified in Operations Procedure C1.4, Power Operation, Rev. 20 and 2C1.3, Unit 2 Shutdown Procedure, Rev. 19, were complied with when coasting down and removing the reactor from operation; and the requirements specified in Operations Procedure DS, Reactor Refueling Operation, Rev. 20 were complied with during the course of fuel transfe The inspectors monitored core shuffle activities from the spent fuel pool, the refueling cavity and the control roo '

While monitoring outage activities, the inspectors questioned the operability of the door interlocks associated with the shield building airlock. The licensee stated that the interlocks were not operating properly. Further questioning by the inspector revealed that the licensee has not established a surveillance or preventative maintenance procedures for the shield building airlock doors or the auxiliary building special ventilation system doors. Failure to have established surveillance or preventative maintenance procedures for these doors is contrary to TS 6.5.c. which requires maintenance and test procedures to be developed to satisfy the routine testing of engineered safeguards and equipment as required by the facility license and technical specifications, and is therefore identified as a Violation (282/89008-01; 306/89008-01(DRP)). Modifications to improve the shield building airlock door interlocks have been pending for several year . Management Meeting (30702)

The inspectors attended the March 16, 1989 Safety Audit Committee (SAC)

meeting held at the licensee's facilit During the course of the meeting the new resident inspector was introduced. The inspectors then conveyed their perspective.s on various safety issues, including those which the licensee will encounter as the plant ages and moves towards longer fuel cycle On April 11, 1989 management oersonnel from Northern States Power Co. (listed above) met with Re; anal Administrator A. B. Davis and other NRC personnel (also listed abo /e) to discuss the company's " Pursuit of Excellence Program." This program is about one and one-half years old, having been instituted in 1987 to aid in improving plant performanc The latest iteration of the program was discussed by the plant staf Subjects reviewed included: personnel issues, procurement, technical support, maintenance, operations and training, administration, and management. A few highlights from the meeting are as follows:

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Review System (PMRS) will relate year end cash bonuses and annual E raises to company profits and individual business unit performanc

All six operating crews are making week-long visits to other operating nuclear plant :

^. Management meetings are being held with the staffs of other 2-loop Westinghouse plant Improvements in procurement practices have been made in response to the recent findings of the NRC Vendor Inspection Branc Configuration Management and SSFI efforts are under wa Maintenance backlog remains lo A significant increase in number of licensed operators is planne Improvements are being implemented in the area of surveillance testing to reduce the number of citations and the number of tests

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The Error Reduction Task Force continues to be active in evaluating plant events and in recommending corrective action In response to Mr. Eliason's request for feedback from the NRC on matters of interest to Prairie Island, Mr. Davis discussed the recent Vendor Inspection Branch findings, EQ follow-up inspections, Independent Plant Evaluations, Event V reviews, the Tech. Spec. improvement program, and ,

results of maintenance team inspection ' Temporarv Instructions (Closed) TI 2515/101: Loss of Decay Heat Removal (Generic Letter No. 88-17)

Recent loss of decay heat removal events in the nuclear power industry have prompted the NRC to issue two Generic Letters (GLs),

GL 87-12, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled." and GL 88-17, " Loss of Decay Heat Removal." The purpose of this TI is to review the licensee's " expeditious actions" requested in the latter G l l

05,.01 Genera The licensee's submitted response dated January 6, 1989, to GL 88-17 was reviewed. The response comprises both the

" expeditious actions" and the " programmed responses" requested by the GL. Comments on the specific areas identified in the GL follo .02 Trainin The resident inspectors 'ttended one of the day-long training sessions given to eaa rperating crew. This training covered the Diablo Canyon evera, the Westinghouse WCAP safety analyses of reduced inventory operation, and the newly revised operating procedure for reduced inventory operatio i

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05.03 Containment' Closur ~

Normal and emergency procedures fo closing the containment If the need should arise.during reduced a

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-inventory operation are available in the control' room and were '

reviewed by.the inspector. A log sheet of current containment boundary bre' aches is maintained in the control room. This log sheet describes each opening, when.the cpening was created, when it was closed, and includes the work request number.. A special containment penetration for steam generator. test cabling had been' designed and installe .04' Reactor Coolant Temperature. A total of three independent core exit thermocouple ("in-cores") are temporarily jumpered during reduced inventory operation to provide continuous indication on the Inadequate Core Cooling Monitor (ICCM) panel in the control roo These indications are logged by the control room operator every half hou '.05 RCS Water Level Indication Two indications'of RCS water level are available in the control room when in reduced inventory operation. High and low level alarms are available-on one of the indicators (A respanned refueling water level indicator). Both indications are- recorded on a control room log sheet every half-

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hour. As noted in licensee's January 6, 1989 response to GL 88-17, these indications are not totally independent in the sense that they share a common tap on the RCS. Modifications are slated to. correct:

this as part of the programmed enhancements actions required by GL 88-17.-

'05.06 RCS Perturbation Procedures and controls have been established to avoid operations which might lead to perturbations of the RCS when in reduced inventory operatio Along this line and in conjunction with the current Unit'2 outage, systems which have potential for disturbing the RCS if work is performed on them were identified and listed by the plant planning organizatio Then individual outage work requests were reviewed in a planning meeting attended by representatives of operations, maintenance, and engineering. The result of- these efforts was a summary of the various jobs which can be performed at the various stages of the outage and still not perturb the RCS during operation with reduced inventor .07 RCS Inventor The plant operating procedure requires both a

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safety injection pump and the charging system to be available during reduced inventory operation. The SI pump is required to be available to inject directly into the reactor vessel. Times required to initiate water addition to prevent core uncovery and flow rates needed to accomplish this are provided in the procedur Requirements are also included for an RCS vent path when emergency injection is require .08 Hot Leg Flow Path The use of nozzle dams has been temporarily suspended until the question of hot leg venting can be further evaluate a

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• % 105.09 Loop Stop Valve Prairie Island does not have_these valve (Closed) TI 2500/17, Inspection Guidance for Heat Shrinkable Tubing l

4 Resident Inspectors were asked to provide information on the' extent l of' deficient electrical splices. involving heat shrinkable tubing (Raychem splices). The licensee has taken the following actions in response to the Information Notice on this subject, IN 86-5 Minimum acceptance criteria for splices were established based'on Raychem and utility testing results. Safety-related splices for

. Unit 1 and' Unit 2 were then examined by plant Quality Engineering personnel. Splices not meeting the acceptance criteria were-replaced with Raychem splices meeting current-Raychem standard A total of 123 such splices were replaced in Unit 1, and 129 were replaced-in Unit 2. (See also' Inspection' Reports No. 50-282/86012(DRS); No. 50-306/86014(DRS)).

1 Information Notice Followup (92701)

(Closed)-282/87023-IN: Loss of Decay Heat Removal During Low Reactor Coolant Levei Operation This subject is summarized under TI 2515/101, paragraph 9, in this repor (Closed) 282/88902-IN (IN 87-59): Potential RHR' Pump Loss The subject of this IN is the adequacy of recirculation' flow for RHR pumps under very low flow or dead-headed conditions. The licensee reviewed this question with the pump manufacturer and has concluded that

.the existing design and the current operating practices are acceptabl The safetyLinjection system was also reviewed by the licensee and found to'be acceptabl (Closed) 282/88903-IN (IN 88-01): Safety Injection Pipe Failure This subject was investigated by licensee and results reported to the NRC in the response to NRC Bulletin No. 88-0 (Closed) 282/88051-IN: Failures of Main Steam Isolation Valves (MSIVs)

The experience related in this IN is not applicable to Prairie Island because the MSIVs are of a different design than those discusse (Closed) 282/88055-IN: Potential Problems Caused by Single Failure of an Engineered Safety Feature Swing Bus Swing buses are not used at Prairie Islan (Closed) 282/88067-IN: PWR Auxiliary Feedwater Pump Turbine Overspeed Trip Failure

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' Licensee review indicates that the current plant actions of inspecting the overspeed tappet ball and testing the overspeed trip at each refueling outage is responsive to the concerns expressed in the I . Exit (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 at the conclusion of the April on April 17, 1989. The inspectors discussed the purpose and scope of the inspection and the findings. The inspectors also discussed the likely information content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any documents or processes as proprietar Attachments: Notice of Violation Section 12 - Access Control -

Personnel (UNCLASSIFIED )

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