IR 05000321/1990006
| ML20034C184 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/18/1990 |
| From: | Conlon T, Hunt M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20034C181 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-321-90-06, 50-321-90-6, 50-366-90-06, 50-366-90-6, NUDOCS 9005020148 | |
| Download: ML20034C184 (9) | |
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- NUCLEAR REGULATORY COMMISSION -
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UNITED ST ATES
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REGION 11
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101 MAfuETTA STREET.N.W.
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t-ATLANTA, GEORGI A 30323
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s-Report Nos.:
50-321/90-06 and 50-366/90-06 Licensee: Georgia: Power-Company _
P. 0. Box 1295 Birmingham,LAL 35201-Docket Nos.: 50-321 and 50-366-License Nos.: L DPR-57 and NPF-5 Facility Name:
Hatch 1 and 2 March '-9'and 15,-23,;1990 Inspection Conducted:
Inspector:
jMb h')/3) T'b -
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M. D. Hunt
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Date Signed Approved by:_
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T. E. Conlon, Chief Date Signed-l Plant Systems Section
Engineering Branch Division of Reactor Safety
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SUMMARY
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Scope:
This routine, unannounced inspection was conducted in the areas' of the licensee's conformance to Regulatory. Guide (RG) 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants-To' Access Plant and Environs Conditions During and Following an Accident and review of 1,1censee. actions on previous
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inspection findings, s
Results:
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In the areas inspected, violations or deviations were not. identified..
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No major weaknesses were identified in the areas inspected.
The licensee was
. responsive to requests and concerns expressed by the inspector. The technical i
personnel contacted by the inspector were knowledgeable and thorough in
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furnishing information to support this inspection.
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Based on -observations and examination of records, drawings and installed I
equipment, it was concluded by the inspector that the licensee meets the i
requirements of Regulatory Guide 1.97 at the present time.
Future modifica-tions will be required to comply with the_NRC approved-flux monitoring system.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
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- P. E. Fornel," Manager.- Maintenance
- G. A. Goode, Manager - Engineering Support-H. C. Nix,. General Manager - Nuclear P.lant
- S.-B. Tipps, Manager - Nuclear Safety.and Compliance.
- T. Moore, Assistant General Manager - Plant Support
- D C.- Wilson, Project Engineer
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Other licensee employees - contacted during this inspection included '
craftsmen, engineers, operators, mechanics. security force members,
-i technicians, and administrative personnel.
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Other Organizations I
- D. A. Brock,.I&C/EQ' Engineering Support, Southern Company Services.
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- Y W. Yee, Elect / Control Systems Group Supervision, Bechtel Power Corp.
=1 NRC Resident Inspectors John Menning j
- Attended exit' interview y
Acronyms and initialism used throughout this report are listed in the last paragraph.
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2.
Inspection of Licensee's Implementation - of Multiplant Action A-17:
Instrumentation. for Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Regulatory Guide 1.97)
(25587)
Criterion 13. " Instrumentation and Control," of Appendix A to 10 CFR'
Part S0 includes a requirement that instrumentation be provided-to monitor-variables and systems - over their anticipated ranges.for accident l
Conditions as appropriate to ensure adequate safety.
Regulatory-i Guide 1.97 (RG 1.97) describes a method acceptable to the NRC staff for.
complying with the Commission's regulations to provide instrumentation to i
monitor plant variables and systems-during and following an accident.
The purpose of this inspection was to verify that the licensee has an
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instrumentation system for assessing variables and systems during and
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following an accident, as discussed in Regulatory Guide (RG) 1.97.
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accident conditions, it is necessary-that the operating personnel have:
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t (1) information that permits-the operator to take preplanned-actions to accomplish a safe plant shutdown; (2) determine.whether the reactor tripped, Engineered. Safety-Feature Systems '(ESFS) actuated, and that other manually initiated safety (3) provide information -to operators that will-systems-importan
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intended functions; 'and enable them to determine the potential for causing a gross breach of; the barriers to radiation release and to.. determine if a gross breach of-barrier has occurred.;
For-this' reason multiple instruments with overlapping ' ranges may be necessary. -The required instrumentation must-be capable of. surviving the accident environment for the length of time its-
- operability is required.
It 'is desirable that components continue: to function following seismic events.
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l As a result, five types of variables have been specified that serve as-
' j guides in defining criteria and the selection of accident-monitoring-instrumentation.
The types are: Type A'
.Those variables that provide information needed to permit the control. room operating personnel to take specified manual actions for which no automatic control is-provided-and that are required for safety systems-to accomplish their functions for design basis accident events; Type B - Those-variables that provide l,
information to indicate whether plant safety functions 'are being
accomplished; Type C - Those variables-that provide information to '
indicate the potential for barriers being> breached or the actual breach of
barriers to fission product release; Type 'D.- Those variable that provide
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information to indicate operation of individual safety _ systems and1other j
systems important to safety; Type -E - Those variables to be monitored in -
j determining the magnitude of the release of radioactive material and for i
continuously assessing such release.
The design and qualification criteria are separated into the separate l
categories that provide a graded approach to requirements. depending on
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the importance to safety of the measurement of-a specific-variable.
Category 1 provides the most stringent requirement and.is intended for key variables.
Category 2 provides less~ stringent requirements and generally applies to instrumentation designed for indicating systems operating
status.
Category 3 is intended to provide requirements that will ensure the high quality off-the-shelf instrumentation is obtained and applies to backup and diagnostic instrumentation.
A key variable-is the single l
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accomplishment of a safety function (Types B and C), or the operation of a
safety system (Type D), or radioactive material release (Type E). Type A
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variables are plant specific and depend-on the operations - that the designer chooses for planned manual actions.
Inspection of Categories 1 and 2 equipment was performed as described below.
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Categories 1 and 2 Instruments for Units 1 and 2 The instrumentation listed in the Table was examined to verify that the design and qualification criteria of RG 1.97 had been satisfied.
The instrumentation was inspected by reviewing drawings, procedures,
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i data sheets, and 'other documentation and performing' walkdowns' for l
visual observation 'of selected installed equipment including CR -
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indicators and recorders. The following areas were inspected:,
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(1) Equipment Qualification - The EQ Master Equipment List' and the-l Q-List were reviewed for confirmation that the: licensee had,
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addressed environmental qualification requirements _for. Class IE
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equipment.
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't (2). Redundancy - Walkdowns were' performed to verify by visual.
i observation that. selected. instruments were installed as specified and that separation: requirements were met.;
In-addition, drawings. for all listed Category 1 instrumentation-
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were reviewed to verify redundancy and channel separation..
J (3) Power Sources Drawings were' reviewed to verify E the
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instrumentation'is energized-from a safety-related' power source.
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(4) Display and Recording' - Walkdowns werei performed to. verify b,v
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visual observation that the specified display 'and ' recording.
instruments were installed.
Drawings: were reviewed : to ' verify
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there was'at least one recorder in a redundant channel ~and two
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indicators, one per division-. (channel) fori each measure.d variable.
(5) Range - Walkdowns: were performed L o ' verify the actual-range of t
the indicator / recorders was as specified in-RG 1.97 or the SER.
Review of calibration procedures verified sensitivity and overlapping requirements of RG 1.97: for instruments. measuring the same variable.
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Interfaces - The drawings and Q-List were-viewed to-verify that-safety-related isolation ' devices were usedo when required to-isolate the circuits from nonsafety systems.
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(7) Direct Measurement - Drawings were reviewed to verify that the parameters are directly measured by the sensors.
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(8) Services, Testing, and Calibration - The maintenance program for~
performing ' calibration and surveillances was reviewed and ;
discussed with the licensee.
Calibration and surveillance
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procedures and -the latest calibration completion date for each instrument were reviewed-to verify the instruments have a valid calibration.
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t Category 1 Instruments
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All instrument numbers are common to both Units.1 and 2 unless prefixed by the number 1 or 2.
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t A Variables-Instrument Number
Variables-Channel or Division-
RHR Service Water Flow EII-FT1 N007 A&B
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EII-FT R602-A&B j
I N600 A&B-He/02 Content in Drywell P33 H,R R601-A&B:
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P33 H
g P33 H,T N604 A&B-J P33 0,T~N601 A&B.
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-P33 0 R:R602 A&B-P33 0
,R R603.A&B
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RPV Pressure B21-PT N090 A&D B21-L/PR R623'A&B j
RPV Level B21-LT N091 A,B,C,80-B21-LT R604 A&B i
.821-L/PR R623 A&B.
l Drywell Atmosphere Temperature IT47-TE N001 B&L.
i 1T47-TE N002,4,6,8&9:
1T47-TR R611 i
IT47-TE N001 J,K,&M
IT47-TE N003,5,7&l0
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IT47-TR R612 2T47-TE N001A, B&L
2T47-TE N002,4,8&9 2T47-TR R626 2T47-TE N001,J,K,&M 2T47-TE N003,,5,7,&l0
.q 2T47-TR R627 Suppression Pool Temperature 148-TE N301A To N311A T48-TR R647
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T48-TE N009A&C
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IT47-TR R611
2T47-TR R626 T48-TE 009B&D
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. IT47-1R R612 2T47-TR R627 f
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Diesel Generator Output Voltage Current Power D/G 1A 1R11-VM-R676 1R43-AM-R653 1R43-WM-R601A D/G IB 1R11-VM-R677 1R43-AM-R654 1R43-WM-R601B D/G 1B 2R43-RM-R910 2R43-AM-R906 2R43-WM-R901 D/G 1C 1R11-VM.R678 1R43-AM-R655 1R43-WM-R601C D/G 2A 2R43-VM-R904 2R43-AM-R903 2R43-WM-R907 D/G 2C 2R43 VM-RE918 2R43-AM-R914 2R43-SM-R915 Note D/G 1B is the swing diesel.
Those instruments prefixed with 1 or 2 are i
located on the Main Control Panel B Variables
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Instrument Number Variables Channel or Division
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RCS Soluble Boron Concentration 2P33-BOE N054 (Thissystemiscommonforbothunits)
2P33-BIT R078 Coolant Level in Reactor B21-LT N085 A&B B21-LR R615 B21-LI 610 B21-LT N027 B21-L1 R605 Bel-LT R038 A&B
$PDS Computer Drywell Pressure T48-PT N020 A&B TA8-L/PR R607 A&B T<3-PT N023 A&B T48-PI R631 A&B T48-PT N003 A&B T48-PR R601 A&B Drywell Sump Level Gil-LT N047 A&B Gil-LIS N001 & N002 Gil-LT N015 A&B G11-LIS N008 & N009 Primary Containment Pressure T48-PT N008 A&B'
T48-PI R632 A&B T48-PR R6084 R609
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C Variables Suppression Pool Water Level T48-LT N021 A&B
T48-L/PR R607 A&B
T48-LT N010 A&B
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T48-LI R622 A&B
D Variables
Condensate Storage Tank Level P11-LT N001
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IP11-LI R900 J
2P11-LI R601 e
RCIC Flow E51-FT N003 E51-FI R613
HPCI Flow E41-FT N008 E41-FI R613
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Core Spray System Flow E21-FT N002 A&B E21-FI R601 A&B-SLCS Flow C41-PT N004
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C41-PI R600 SLCS Storage Tank Level C41-LT N001 Col-LI R601 RHR System Flow Ell-FT N015 A&B
Ell-FI R603 A&B Ell-FR R608 A&B
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RHR Heat Exchanger Outlet Temp.
Ell-TE N027 A&B IT47-TR R611, R612 2&47-TR R626, R627
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Discussion and Comments
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t During) the NRC's Aprilaudit it was noted that certain Post Accident Monitoring (P 24-26, 1989 Detailed Control Room Design Review (DCRDR instrumentation was not specifically highlighted on the main control room panels at Plant Hatch, Georgia Power Company (GPC) was requested to resolve this item considering methods used by other nuclear plants-As
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the result of GPC's review, the decision was made to highlight only the PAM
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instrumentation listed in the Units 1 and 2 technical specifications.
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review of these instruments and comparison of the requirements of the operating procedures with the PAM instruments highlighted determined that ample instrumentation was identified for the operator to discern
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which is intended for use under accident conditions.
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The neutron flux monitoring 1,ystem instrumentation was determined to be
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acceptable on an interim basis at the time the RG 1.97 Safety Evaluation Report was issued.
NRR and the BWR hners Group are currently negotiating to resolve what-instrumentation =is ' acceptable that will meet the requirements of RG 1.97 and 10 CFR 50.49.
The licensee advised the inspector that they are awaiting the decision regarding acceptable equipment before committing to an approach to meet the full intent of RG 1.97 in this area.
Since an acceptable flux monitoring system is still in question, the presently installed system is still in an interim condition that is acceptable as stated in the SER.
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Actionor,PreviousInspectionf_indings(92701)
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(Closed)
Violation, 50-34/366/89-02-01, inadequate Administrative--
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Procedures. The licensee issued a response dated July 7,1989.
Procedure i
No. DI-ADM-05-1087N Upgraded Procedure Develonent and Processing was revised on 4/4/@ to insure that any variance between a procedure and a
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vendor manual is documented.
Procedure 52PM..R-22-001-05, 4160 Yolt AC Switchgear and Associated Electrical Components Preventive Maintenance was revised to address vendor recommended checks of bus connections and the main disconnecting contacts and stuports.
Procedure No. DI-ADM-05-1087N has now been cancelled and replaceo by Administrative Procedure 10AC-MGR-003-03, Preparation and Control of Procedures.
This procedure requires justification for any variation between a procedure and a related vendor manual. Thit violation is closed.
Exit Interview The inspection scope and results were summarized on March 23. 1990, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report.
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Acronyms and Initialism
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AM Ammeter AT Analyzer Transmitter AR Analyzer Recoeder CR Control Room D/G Diesel Generator EDG Emergency Diesel benerator EQ Environmental Qualitication EQ DOC PAC Environmental Qualfiication Documentation Package ERDADS Emergency Response Data Acquisition Display System FSAR Final Safety Analysis Report H,RM Hydrogen Indicator I
HR High Range Radiation Monitor HT Heat
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H.T Hydrogen Transmitter Ikl Inspector Follewup-Item FI Flow Indicator FIC Flow Indicating Controller FR Flow Recorder FT Flow Transmitter HPCI High Pressure Core Injection L1 Level 7ndicator
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LIS Level' Indicating S' witch L/PR Level / Pressure Recorder LR Level Recorder LT tevel Transmitter NCR Nonconformance Report NOV Noncited Violation ND Neutron Detector NI Neutron Indicator NRC U. S. Nuclear Regulatory Connission
0xygen Indicator 0T 0xygen Transmitter i
P Post Accident Monitoring PI Pressure Indicator PR Pressure Recorder PT Pressure Transmitter PTM Pressure Trip Module RCIC Reactor Core Isolation Cooling R1 Radiation Indicator RC Reactor Coolant RCS Reactor Coolant System RG Regulatory Guide RM Radiation Monitor RHR Residual Heat Removal RPV Reactor Pressure Vessel RVP Reactor Vessel Pressure RR Radiation Recorder
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SER Safety Evaluation Report SIS Safety injection System SPDS Safety Parameter Display System TE Temperature Element TI Temperature Indicator TR Temperature Recorder VM Yoltmeter WM Wattmeter