ML20207E436
ML20207E436 | |
Person / Time | |
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Site: | Fort Calhoun ![]() |
Issue date: | 12/10/1986 |
From: | Terc N, Yandell L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
To: | |
Shared Package | |
ML20207E402 | List: |
References | |
RTR-NUREG-0737, RTR-NUREG-737 50-285-86-20, NUDOCS 8701020179 | |
Download: ML20207E436 (59) | |
See also: IR 05000285/1986020
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t > APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-285/86-20 License: DPR-40 Docket: '50-285 Licensee: Omaha Public Power District 1623 Harney Street Omaha, Nebraska 68102 Facility Name: Fort Calhoun Station Inspection At: Ft.'Calhoun,_ Nebraska Inspection Conduct % July 7-1 , 1986 ( * # Inspector: N. M. Terc, Emergency Preparedness Speci ist, Date Regional Team Leader Accompanying personnel: J. B. Baird, NRC C. A. Hackney, NRC G. R. Bryan, PNL, Comex P. J. Hof, PNL M. K. Lindell, PNL A. K. Loposer, PNL, Comex - J. V. Ramsdell, PNL G. A.-Stoetzel, PNL Approved: - (1AA 12 /o/E LY A. Yandell, Chief, Emergency Preparedness Dat6 ' and Safeguards Programs Section Inspection Summary Inspection Conducted July 7-11, 1986 (Report 50-285/86-20) Areas Inspected: An announced appraisal of the Emergency Response Facilities (ERFs) was conducted at Fort Calhoun Station on July 7-11, 1986, using IE Inspection Procedure 82212 to determine if the licensee had successfully implemented the requirements in Supplement 1 to NUREG-0737 and the regulations. The appraisal included the Technical Support Center (TSC), Control Room (CR) 870102 PDR A % $$$$$as PDR O
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hi response, Operational Support Center (OSC), Emergency Operations Facility (E0F), 3; ' - and the emergency data acquisition systems as well as the instrumentation, 3
-supplies,- and equipment for these facilities. ~ . [ /Results: Within the ERFs inspected, no violations or deviations were : identified.1However, the NRC inspectors found 14 deficiencies and 7 unresolved
p , : items. -In addition, thel inspectors identified 22 improvement items that were J. _-
.related orally to the licensee but were not included in this report. - 4 5 + t s * - .m
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' - s - : 7 ym . m * ,!g' E i?. - :o _ , ; r - i - 4 , , P 1., - - * fr, '2 0 - r ;j. s, ' j 1 3 ' @[ ' ~ k ' .b,; ; . v , - ' ~. [- , Table of Contents - For the Detailed ERF Evaluation 3: - ' , - . ' Page Number- A .!, - [1.0 . Technical Support-Center (TSC) . . . . . . . . . . . . . . . . . . 8 t ' E , : 1.1 : Physical Faci li ties. . . . . . . . . . . . . . . . . . . . . . . . . - 8 1:- ., _ 1.1.1 . Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - :;1.1.1.11 sSize. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 , . 1.1.1.2 Layout. . . . . . '. . . . . . . . . . . . . . . . . . . . . . 8 11.1.1.3 Location. . . . . . . . . . . . . . . . . . . . . . . . . . . 9 ' 11.1.1.4 Structure 1 . ........................ 9 * -- 1.1.1.5 Habitability / Environment. . . . . . . . . . . . . . . . . . 10 4 ^1.1.1.6 Display Interface . . . . . . . . . . . . . . . . . . . . . 11 1 1.1.2 L-Radiological Equipment and Supplies. . . . . . . . . . . . . . 12 ' 1.1.2.1 Radiation Protection. . . . . . . . . . . . . . . . . . . . .L12
- , T1.1.2.2 : Personnel Dosimeters. . . . . . . . . . . . . . . . . . . . 13
, -1.1.2.3 , Protective _ Supplies . . . . . . . . . . . . . . . . . . . . 13 .4 ~1.1.3 Non-Radiological Equipment and Supplies. . . . . . . . . . . . 14 ' 1.1.3.1 Communications. . . . . . . . . . . . . . . . . . . . . . . 14
0 1.'1. 3. 2 : Records /Drawi ngs. . . . . . . . . . . . . . . . . . . . . . 15
1.1.3.3
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, -Support Supplies. . . . . . . . . . . . . . . . . . . . . . 16 1.1.3.4 Powe r Supply. . . - . . . . . . . . . . . . . . . . . . . . . 16 , 4 1.2 -Information Management . . . . . . . . . . . . . . . . . . . . . 17 1. 2.1 Variables Provided . . . . . . . . . . . . . . . . . . . . . . 17 11.2.1.1 Reg. Guide 1.97 Variables . . . . . . . . . . . . . . . . . 17 .1.2.1.2 Other-Variables . . . . . . . . . . . . . . . . . . . . . . 19
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:1.2.1.3 Relationship to Functional Needs. . . . . . . . . . . . . . 20
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1.2.2 ' Data Acquisition . . . . . . . . . . . . . . . . . . . . . . . 20 ' 1.2.2.1 Data Collection Methods . . . . . . . . . . . . . . . . . . 20 1.2.2.2 Time Resolution . . . . . . . . . . . . . . . . . . . . . . 21
J 1.2.2.3 Isolation . . . . . . . . . . . . . . . . . . . . . . . . . 22
1.2.3 Data Communications. . . . . . . . . . . . . . . . . . . . . . 23 1.2.3.1 Capacity. . . . . . . . . . . . . . . . . . . . . . . . . . 23 1.2.3.2 Error Detection . . . . . . . . . . . . . . . . . . . . . . 24 1.2.3.3 Transmission between ERFs . . . . . . . . . . . . . . . . . 24 1.2.4 Data Analysis. . . . . . . . . . . . . . . . . . . . . . . . . 25 1.2.4.1 Reactor Technical Support . . . . . . . . . . . . . . . . . 25 1.2.4.2 Dose Assessment . . . . . . . . . . . . . . . . . . . .~. . 26
- 1.2.4.3 . Central Processor Capability. . . . . . . . . . . . . . . . 31
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:1.2.5 Data Storage . . . . . . . . . . . . . . . . . . . . . . . . . 32 1.2.5.1 Storage Capabilities. . . . . . . . . . . . . . . . . . . . 32 ' 1.2.6 System Reliability and Validity. . . . . . . . . . . . . . . . 32 1.2.6.1 Validation and Verification . . . . . . . . . . . . . . . . 32 1.2.6.2 Computer Based Systems. . . . . . . . . . . . . . . . . . . 33 1.2.6.3 . Manual Systems. . . . . . . . . . . . . . . . . . . . . . . 33 1.2.7 On Shift Dose Assessment . . . . . . . . . . . . . . . . . . . 34 -1.2.7.1 Dose Assessment Proficiency . . . . . . . . . . . . . . . . 34 1.2.7.2 Dose Assessment Technical Adequacy. . . . . . . . . . . . . 34 1.3 Functional Capabilities and Walkthroughs . . . . . . . . . . . . 35 1.3.1 Operations. . . . . . . . . . . . . . . . . . . . . . . . . . 35 1.3.1.1 Organization. . . . . . . . . . . . . . . . . . . . . . . . . 35 1.3.1.2 Staffing. . . . . . . . . . . . . . . . . . . . . . . . . . 35 1.3.1.3 Acti vati on. . . . . . . . . . . . . . . . . . . . . . . . . 36 1.3.1.4 Communication Interfaces. . . . . . . . . . . . . . . . . . 36 '1.3.1.5 Of fsite Interfaces. . . . . . . . . . . . . . . . . . . . . 36 1.3.1.6 Transfer of Responsibilities. . . . . . . . . . . . . . . . 37 1.3.2 Control'~ Room Support . . . . . . . . . . . . . . . . . . . . . 38 1.3.2.1 ' Technical Support . . . . . . . . . . . . . . . . . . . . . 38 1.3.2.2 Wal kth roughs. . . . . . . . . . . . . . . . . . . . . . . . 38 .2.0 -Operational Support Center (OSC) . . . . . . . . . . . . . . 2 . 39 2.1 Physical . Facili ties. . . . . . . . . . . . . . . . . . . . . . . 39 2.1.1 Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 2.1.1.1 Location. . . . . . . . . . . . . . . . . . . . . . . . . . 39 P.1.1.2 Alternate OSC Location. . . . . . . . . . . . . . . . . . . 39 2.1.1.3 Size, Layout, and Environment . . . . . . . . . . . . . . . 39 2.1.1.4' Display Interface . . . . . . . . . . . . . . . . . . . . . 40 -2.1.2 Radiological Equipment and Supplies. . . . . . . . . . . . . . 40 -2.1.2.1 Radiation Monitoring. . . . . . . . . . . . . . . . . . . . 40 2.1.2.2 Personnel Dosimeters. . . . . . . . . . . . . . . . . . . . 41 2.1.2.3 Protective Supplies . . . . . . . . . . . . . . . . . . . . 42 2.1.3 Non-Radiological Equipment and Supplies. . . . . . . . . . . . 42 2.1.3.1 Communications. . . . . . . . . . . . . . . . . . . . . . . 42 -2.1.3.2 Support Supplies. . . . . . . . . . . . . . . . . . . . . . 42 2.2 Functional Capabilities and Walkthroughs . . . . . . . . . . . . 43 2.2.1 Operations . . . . . . . . . . . ...............43 2.2.1.1 Staffing. . . . . . . . . . . . . . . . . . . . . . . . . . 43 2.2.1.2 Acti vati o n. . . . . . . . . . . . . . . . . . . . . . . . . 43 2.2.1.3 Onsite Interface. . . . . . . . . . . . . . . . . . . . . . 44 . -.
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5 htL!: t ; i ~ . . L 2.2.2 '0SC Functions. . .:. ... . . . . . . . . . . . . . . . . . . . . 44' C' - '2.212.1? Coordination, Assignment, Proficiency, and Walkthroughs . . . "44 d ,, L3.0t'Emergench Operations Facility (EOF). . . . . .-. . . . . . . . . 44 3.1 Physical; Facilities. . . . . . . . . . . . . . . . . . . . . . . 44 3.1.1 Design.. . . . ._. . . . . . . . . . . . . . . . . . . . . . . . 44 L 3.1.1.1 ' Size. . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 13.1;1.2. Layout. . .-. . . . ._. . . . . . . . . . . . . . . . . . . . 45 .3.1.1.3 . Location. . . ... . . . . . . . . . . . . . . . . . . . . . 45 3.1.1.4 -Structure . . . . . . . . . . . . . . . . . . . . . . . . . 46 3.1.1. 5 - Habitability / Environment. . . . . . . . . . . . . . . . . . 46 , 3.1.1.6 Display Interface . . . . . . . . . . . . . . . . . . . . . 46 ~ ~3.1.2 ' Rad'iological Equipment and Supplies. . . . . . . . . . . . . . 47 3.1.2.1 -Radiation Monitoring. . . . . . . . . . . . . . . . . . . . 47 3.1.2.2' Personnel Dosimeters. . . . . . . . . . . . . . . . . . . . 47 . 3.1.2.3 Protective Supplies . . . . . . . . . . . . . . . . . . . . 47 L 3.1.3' Non-Radiological Equipment and Supplies. . . . . . . . . . . . 48 L 3.1. 3.1 - Communications . . . . . . . . . . . . . . . . . . . . . . 48 3.1.3.2 ' Records / Drawings . . . . . . . . . . . . . . . . . . . . . 49 , 3.1.3.3 . Support Supplies . . . . . . . . . . . . . . . . . . . . . 49 ~ 3.2Information Management . . . . . . . . . . . . . . . . . . . . . 49 . 3.2.1 Variables. . . . . . . . . . . . . . . . . . . . . . . . . . . 49 3.2.1.1- Reg Guide 1.97, Revision 2, Variables . . . . . . . . . . . 49 3.2.1.2 'Other Variables . . . . . . . . . . . . . . . . . . . . . . . 50 3.2.1.3 Relationship to Functional Needs . . . . . . . . . . . . . 50 3.2.2' Data Acquisition . . . . . . . . . . . . . . . . . . . . . . . 50 - .3.2.2.1- Data Collection Methods . . . . . . . . . . . ... . . . . . 50 J3.2.2.2 -Time Resolution . . . . . . . . . . . . . . . . . . . . . . 50 5 3.2.2.3. Isolation . . . . . . . . . . . . . . . . . . . . . . . . . 50 3.2.3 ' Data Communications. . . . . . . . . . . . . . . . . . . . . . 50 3.2.3.1 Capacity. . . . . . . . . . . . . . . . . . . . . . . . . . 50 3.2.3.2 Error Detection . . . . . . . . . . . . . . . . . . . . . . 50 3.2.-3.3 Transmission between ERFs . . . . . . . . . . . . . . . . . 50 -3.2.4 Data Analysis. . . . . . . . . . . . . . . . . . . . . . . . . 50 3.2.4.1 Reactor Technical Support . . . . . . . . . . . . . . . . . 51 3.2.4.2 Dose Assessment . . . . . . . . . . . . . . . . . . . . . . 51 3.2.4.3 Central Processor Capability. . . . . . . . . . . . . . . . 51 3.2.5 Data Storage . . . . . . . . . . . . . . . . . . . . . . . . . 51 3.2.5.1 Storage Capabilities. . . . . . . . . . . . . . . . . . . . 51 3.2.6 System Reliability . . . . . . . . . . . . . . . . . . . . . . 51 .
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: 3. 2. 6. 3 ' . Manual Sys tems . . . . . . . . - . - . . . . . . . . . . . . . . 51 ' - 3.31 Functional Capabilities and Walkthroughs . .. . . . . . . . . . . 51 ;. ' :
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' 3. 3.1 ' Operati ons . . . . . . . . . . . . . . . . . . . . . . . . . . 51 3.~3.1.11 1. Organ i zati on. . . . . . . . . . . . . . . . . . . . . . . . 51 73.3.1.2 ; Staffing. . . . .-. . . . . . . . . . . . . . . . . . . . . 52 ug , ' .3:3.1.3 : Activation. . . . . . . . . . . . . . . . . . . . . . . . . 52 ' 3.3.1-4. Communication Interfaces. . . . . . . . . . . . . . . . . . 52 ' . ;3.3.1.5 Offsite Interfaces. . . . . . . . . . . . . . . . . . . . . 52 - : 3. 3.1. 6 Transfer of Responsibilities. . . . . . . . . . . . . . . . 52 j;
g0, 3.3.2 TSC Support. . . . . . . . . . . . . . . . . . . . . . . . . . 53 4? ;c '3.3.2.1 : Technical Support . . . . . . . . . . . . . . . . . . . . . 53
T' 3.3.2.2 Logistic Support. . . . . . . . . . . . . . . . . . . . . . 53 4, - - 3.3.2.3 Implementation'of Mitigating Actions. . . . . . . . . . . . 53 ,, [ . 3.3.3 EOF Functions. . . . . . . . . . . . . . . . . . . . . . . . . 54 3.3.3.~1 Notification / Communication. . . . . . . . . . . . . . . . . 54
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. ' ~ 3.'3.3.2 Dose Assessment . . . . . . . . . . . . . . . . . . . . . . 55 3.3.3.3- Protective Action Decisionmaking. . . . . . . . . . . . . . 55 ' , 3.3.3.4 Coordination of Radiological and Environmental Assessment . 55 > 3.3.3.5~ Wa l kth rough s . . . . . . . . . . . . . . . . . . . . . . . . 56
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~ 5.0~-Exit Interview . . . . . . . . . . . . . . . . . . . . . . . . . 57 is - - Acronyms and Initialisms
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----- s. ' (r s; 7 1 1 ' : DETAILS t . l - , .1.. hersonsContacted- ' *R.'. Andrews,: Division Manager, Nuclear Production *D. Bloemendaal, Consultant - Hydronuclear-Services ' . ' '*C, Brunnert, Superviser, Operations, Quality Assurance - D. ; Burns. Electrical Maintenance Foreman *A._Christensen, Health. Physicist- .M. Christensen, Health Physicist- S. Crites, Senior Designer. - M. Ellis,-I & C Supervisor ' *J. Fisicaro,. Supervisor, Nuclear & Regulatory Industry = *F. Franco, Manager,-Emergency Preparedness , ~ *S. _Gambhir, Section Manager W. Gartner, Senior Electrical Engineer *J. Gasper, Manager, Administrative Services G. Gates, Plant Manager . ~ S. Gebers, Radiation Protection Training Instructor S. Hahn,- Communications Engineer T. Heng,-Senior Engineer *K.-Holthaus, Manager, Reactor & Computer Technical Services *R. Jaworski, 'Section Manager, Technical Services B. Johnston, Programmer Analyst' . J. Kecy, Supervisor, Reactor Performance Analyst *T. McIvor, Supervisor, Technical Services > *R. Mehaffey, Supervisor, Electrical Engineering J. Mixan,.I & C Technician *K. Morris, Division Manager, Quality Assurance *D.- Munderloh, Senior Engineer, Nuclear Affairs *C. Norris, Supervisor, Radiation Services A. Richards, Manager, Quality Assurance
' r *G Hrh, h nk:r, Chemistry & Health Physics
F. Rutar, Senior Engineer, Technical Services 'J. Spilker, Operations Supervisor *J. Tesarek, Reactor Engineer M. Wilson, Assistant Analyst M. Yttri, Prograrcer Analyst * Denotes those present during exit interview. -
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] p' ' fl.04 Technical ~SupportCenter(TSC) ' l 1.1 Physical Facilities P ' ; 1.~ 1.1 : Desian 11.1.1.1t TSizi ; & 1 * ' TheNRCLinspectorsreviewedtheTSCsizeagainst10CFR50.47(b)(8) , V and Supplement 1 to NUREG-0737, item 8.2.1. The inspectors examined the conceptual design, the. Emergency Plan, and EPIPs, and toured the , -facility.' _ - 'The NRC inspectors determined that the TSC was sited in a building > located immediately north of the Auxiliary Building. There were W , three rooms within the building comprising the functional TSC. These rooms (room 113, 4 persons assigned, 200 square feet; rooms 107/115, ' 19 persons assigned, 1000 square feet;_ room 117, one person assigned, i 100 square feet): averaged just over.50' square feet per person of assign'ed ' occupancy:for-licensee and NRC personnel. In addition, there were ; areas for records; storage (designated by EPIP-TSC-1-1 as part of the . .OSC), sanitary' facilities ~, and building support (HVAC and electrical). Workstations provided adequate floor space and horizontal workspace . 'to support the' tasks performed by each of the individuals assigned to ~. the1TSC. Each piece of. operational equipment (CRTs, PCs, printers) iwas accessible for corrective or preventive maintenance or H replacement. The NRC inspectors noted that' circulation space within .the TSC was. limited when the area-just inside the door leading to room 114.was'being;used as a briefing area for OSC teams. : + , Ba' sed on the above, the NRC inspectors concluded that the size of the - TSC appeared to be adequate. 7 . . . , 1.1.1.2 Layout . The'NRC'inspectorsreviewedtheTSClayoutagainst10CFR50.47(b)(8) and Supplement 1~to NUREG-0737, items 8.2.1.c and k. The NRC inspectors examined the conceptual design, the Emergency Plan, and EPIPs, and toured the facility. The NRC. inspectors determined that the layout of the TSC provided for - a single room (designated as rooms 107 and 115 in Procedure EPIP-TSC-1-1) which constituted the focal location for ' emergency response activities. This room had areas for emergency -assessment and radiological assessment that were separated from each other by a management / communication area. This arrangement provided physical separation of the two assessment areas without impeding visual access of plant status boards by radiological assessment personnel. Other TSC personnel (NRC team and the Procedure / Training Supervisor) were located in close proximity to this room on a . corridor that loops around rooms 107/115. The location of doors to
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, _- .c , h - g yx _.,I . ~ c C x'~ :roomsi107/115.al16wed personnel.to move about' freely without- h@ ' ~ Ldisrupting; activities inLunrelated work areas.- Each individual or- pCW > 'l team was located within an_ adequate distance of other. individuals - ? L swith'whom interaction ~'aust'take place.i - W, ~ , .._ ...;.- - v .: Based.onithe above, the NRC inspectors concluded that th'e TSC layout ' - s - b g' , . appeared to be adequate. , , Iq. 3 ' m 1.1.1'.3 Location- ., 2 y , , . .. # ' , ;The'inspectorsireviewed the TSC location with' respect to the control - T- room, the OSC, and the' EOF.,:The TSC was located within the site, / protected area'on,the north side-of the' plant. 'The south wall of the ~ ~ ~ # qe (TSC.was shared ~as the' north' wall of theLauxiliary building. A m ' "In ~ . , ,. machine shop was'ontthe east. side of the TSC. The distance from the ' ' y~$ , , TSC.to ,the_ control-room could.be. traversed in approximately < - 2 minutes. -A major part of;the OSC was contiguous.to_the TSC; the , e" . _. ,, remainder of the OSC was: contiguous to the; control room. Although
L~ 'these!1ocations' precluded face-to-face interaction between the TSC,
s ' 'the control. room,fand the_ EOF. located 17 miles distant, adequate- T - iinteraction was available through_ installed communications systems. 4 j Face-to-face interaction between the TSC and part of the OSC was -available. - - - ~ Based on the above,;the inspectors concluded that the TSC location " _ appeared to be adequate. f L1.1.1.4.-Structure ~ , The NRC inspectors reviewed the TSC structure using the documentation :provided in Omaha Public Power District letter dated June 1, 1981, ~ , ' which forwarded, as Attachment 1, " Design Description for Technical e Support Center,. Fort Calhoun Station, Unit No. 1," (Sections 2.1 and 4.1),'and in the plant microfilm library, microfilm CART 1176, page , s on frame 851, " Gilbert Associates Design Calculations." ', . Additionally, the NRC inspectors examined the TSC structural. ~ features. - The NRC inspectors determined that the TSC building was designed to ' meet the criteria of NUREG-0696; i.e., to withstand the most adverse conditions reasonably expected during the design life of the plant including adequate capabilities for (1) earthquakes, (2) high winds - --(other than tornadoes), and (3) floods. The TSC need not meet 4 seismic Category I criteria or be qualified as an engineered safety feature. - The NRC inspectors noted that the TSC was composed of the protected . ' area and the equipment area, and that the protected area was manned during TSC activation. The TSC building was constructed of a heavy e concrete mat with 1 foot thick reinforced concrete walls and -4 , .?
p% J p. qyw ; ~ , , , - ' <' ' . - at ,- %,? ". y' ; e( 4lj f[. j o ; ~ .; , 'y; 4 ' , _i> ' . < QQfl. * '? - g, ' y' ^' -- c - - , . - ;10 : , - y. .; g_ ' > - 6:: k. - m .,.. ' x - ,- ~ , l' ~ : ceiling'.' Portable flood barriers with inflatable boots, provided to J .q ; protect al1, exterior doors of(the:TSC to~a height of 4 feet above - ~- a ground 11evelfagainst flooding,1were designed for.a~100 year recurrence p. m'f J ; frequency.-LThe building was constructed in accordance with the ' T 9, -Uniform Building Code,;1976. n .. ";' 4 - - CompuOrs?locatedin'theTSCwereprotectedagainstlossofoff-site " - power.by an; uninterrupted power. supply (UPS) which incorporated a m &. , '1 ' - 160-cell; storage' battery and a dedicated diesel generator. A< . , , . _ _ - -. Based on the above, the inspectors concluded that the TSC structure - ; appeared'to be adequate. ' &[[F %m v "f -1.1.1.52eHabitability/ Environment- . . _ . .. c * ~' LThe NRC. inspector's. reviewed the design Enalysis of the TSC and held - ' discussions with the licensee representatives to determine if the TSC. radiological controls. net.the requirements of-10 CFR 50.47(b)(8), - 110 CFR:50.47(b)(11), and item 8.2.1 f. of Supplement 1 to NUREG-0737. The5NRCinspectorsdeterminedthattheTSChadbeendesignedand ~ - . . constructed _to provideLhabitability which meets the 10 CFR Part 50, ' - Appendix A, Genera 1' Design Criterion 19 for facility control' rooms. , . ~This is equivalent to the NUREG-0737, Supplement 1, requirement - , # ,- specified,in' item 8.2.1 f., which requires that TSC radiation ~ -protection be provided which assures'that' radiation exposure not > lexceed 5 rem who.le body, or its equivalent to any part of the body, ' .for the duration of an accident. The licensee's design evaluation ~ . . indicated that' doses to'the whole body-and thyroid of emergency ' workers in the TSC'would not exceed 3.1 rem and 24.5 rem, e respectively, during a:30-day period following an accident. This is , ' achieved by 18 inches of concrete shielding by the walls and ceiling o , ' and a filtered ventilation system. The protection analysis was based , .on exposure from inleakage, plume immersion, direct radiation from '# the containment, and' buildup on the charcoal filter as a result of a -loss of coolant accident (LOCA) described in Regulatory Guide 1.4. . I+ - The_NRC inspectors noted that protection from airborne radioactivity - . 'was provided by the TSC ventilation system which maintained the N nonequipment' area of the TSC at a positive pressure and passed the building air through a prefilter, high efficiency particulate air ~ . , (HEPA) filter, and charcoal filter. The HEPA and charcoal filter removal efficiencies were initially tested in June 1981 and after ; activation of the deluge system in June 1985. The NRC inspectors noted that a special: procedure had been provided to verify flow rate through-the unit and a monthly preventive maintenance schedule had
,
been set up. However, licensee representatives stated that no i routine filter efficiency testing progrgm had been implemented, since . ~ there were no technical specification requirements for this testing.
r l'
1
L, r
-w - - --_.. -._._._._._ ,_ , _ _ . _ . _.,. _ ._ _ _ ..~ ..._,__, - . ,_--..- _ ,_ . _ - - - - . - - 4
, .
11 >The NRC inspectors considered this to be a programmatic deficiency, since the design analysis for meeting TSC habitability criteria depended on achieving iodine removal efficiencies per Regulatory Guide 1.52. Based.on the above, the NRC inspectors concluded that the following deficiency was identified in this area: The licensee failed to establish a routine testing program for the TSC atmospheric filtration system to adequately verify that the design criterion filtration efficiency was being maintained. (285/8620-01) 1.1.'1.6 Display Interface The NRC inspectors reviewed the TSC display interface against the guidance in NUREG-0696. The NRC inspectors examined the emergency plan and EPIPs, the system description, and the operating instructions and training guide for the Emergency Response Facility Computer Systern/ Safety Parameter Display System (ERFCS/SPDS), and inspected the facility and operated the Cathode Ray Tubes (CRTs). The NRC inspectors determined that Room 115 contained pre-formatted status boards for plant and radiological data, a chart of the nuclides, a plant electrical system diagram, a plume EPZ map, an organization chart, and an emergency classification matrix. There were four unformatted display areas: a tackboard, two chalkboards, and a marker board. In a nearby area there was a microfilm reader and printer. The NRC inspectors noted that each of the status boards in the TSC was readily visible and understandable to those personnel who needed to monitor the information, that each display could be updated in a timely manner, and that the number of display boards was adequate for the needs of the TSC. The NRC inspectors also noted that the TSC contained five terminals for the ERFCS/SPDS. This computer system provided a wide range of plant, radiological, and meteorological variables that could be accessed by personnel in the TSC. Data points in this system were labeled by variable name and units of measurement. Key variables relevant to TSC personnel were available on special display pages that were readily accessible even though they had not been integrated within the overall SPDS display menu. The inspectors determined that existing reference materials available within the TSC reference library (0 PPD OI-ERFCS-1-1) were well written and complete. However, they would be suitable for the expert user or computer programmer but not for the novice or infrequent user.
- - - ~ - - - - 7;q'; y- >- - 42 . Ew:. m ' > < , ' # - W _ y . - - . k ls - - 12 ; , , ye ' . , Tl[ , , . ~ ?The NRC inspectors determined that the TSC display interface was acceptable. .The data ~ required to fill.out the emergency assessment . ~ ;and radiological assessment-status boards were consolidated on two ~ . ' 1 * - . readily accessible display pages whose.page number was keyed to the 1 ' . iforms: control number of the manual' data sheets. ~ ? ~ Based'on the above,Lthe NRC inspectors concluded that the TSC display , . interface'. appeared to be adequate. * T 1.1.2 -Rid'iologicaliEquipment and Supplies' ' 1~ o 21 .1~. 2.11 [Ridiation' Protiection - x 7 The NRC: inspectors' reviewed the TSC radiation monitoring i instrumentation, inventory, examined fixed and portable instruments, " 4 3 - and' reviewed ten' selected; instrument calibration records ~to determine V ' ' - whether the radiological monitoring capabilities met the requirements .of 10 CFR 50.47(b)(8), 10 CFR-50.47(b)(11), Section IV.E.1 of Appendix E;to;Part 50, and item 8.2.1 f. of Supplement 1 to NUREG-0737. * . , - The'NRC: inspectors noted that an area radiation monitor and a - , __ particulate, noble gas, and iodine air monitor system were located in 1 the TSC/OSC area to continuously monitor radiation and airborne A radioactivity; levels in the TSC and OSC part of the TSC during an p' . accident.' In addition, 1 low-range and 1 high-range portable dose rate monitors, 3 high volume air samplers, 2 sample counters and a ' , ~ _ radiciodine air sample counter would be provided for TSC/OSC * ^ ' habitability ~ surveys and monitoring in plant activities during an ' * .. , accident. 'In reviewing the monitoring responsibilities of the TSC .and OSC, the NRC= inspectors concluded that the number of dose rate ' ' instruments.provided was not adequate to support simultaneous TSC/0SC , habitability surveys and onsite and in plant monitoring actions. s c 3 ' The NRC inspectors observed that various radiological' supplies and J . equipment were maintained in the TSC. The NRC inspectors conducted a review of the surveillance test (ST-RM-3) for operability and a- inventory of emergency plan supplies and equipment, and a sample of
<
Eten calibration procedures and records for radiological instrumentation. The NRC inspectors concluded that calibration
'l procedures and records were adequate. In addition, the NRC :
inspectors noted that calibration stickers and performance of ' operability tests for emergency equipment indicated that radiation '
i
protection' equipment was being calibrated and maintained as stated in
4
the Plan and procedures.
-
Based on the above, the NRC inspectors concluded that the following
,
deficiency was identified in this area:
. I i
-
!
! - - _ , _ . , _ _ - , _ _ . - - - . , , - , _ _ , . - _ - _ _ _ _ . _ - - - - - - _ _ _ _ _ - _ - - - - ;
Y 5 j;! ' ~ .: x h w . l ~ .. . ' ' - O - ~ ' -' .; '{_ " }Lb r. j y p- aw y , jg , * , _ ' "% = -1v t > . s ~c e ' - * - ; m :13: Kgg %' > 7. ' . u. ', . - , * n -r . "n 7 & 7f ~^ > *? kThe number 'of' portable dose rate survey instruments provided 4 - lwere insufficient-to support the radiation monitoring functions , y ,W ' (assigned =to-the TSC and OSC. 1(284/8620-02) " << V:g. ; ~ 7. . . ~ 1 _ 3., Personnel' Dosimeters + . p ' ,( 1 { 1 ; c ,2. 2 . , . ' ' The NRC inspectors reviewed the TSC radiation dosimeter inventory and- ~ . ~.
, t ..3 E '
; examined. selected self-reading dosimeter. calibration stickers to : determine-if the radiological monitoring capabilities met the K M%,- 7_~OA~ , requirements of 10 CFR 50.47(b)(8), 10 CFR 50.47(b)(11),' tSection.IV.E.1;of Appendix E to Part 50, and item 8.2.1 f. of h+ .-' _ D:/. S.upplement 1.to NUREG-0737.' j s , ' __ _ nThe NRC inspectors-noted that 20 thermoluminescent dosimeters (TLDs) ' 1- :QV ' were provided to-supplement licensee's preassigned dosimeters. The <.f~ , NRC inspectors also noted that 20 (0-500 mrem) and 20 (0-50 rem) - % < 4 self-reading dosimeters were provided for the TSC/OSC. staff, with 5 ' ^~t% ..(0-100) ren,and 6 (0-500) arem dosimeters provided for the,TSC staff Cf i and rescue squad _ members, respectively. -The NRC inspectors observed 'that-during the walkthrough in the TSC, a log'of self-reading V ,, - .. dosimeter-issuance and personnel doses was maintained. ~However,'the ~ , - '%: e NRC inspectors reviewed.the self-reading dosimeter inventory for the i , TSC and determined that with the potential TSC/OSC occupancy of over + , 50 people, the' dosimeters available in the TSC/OSC were not , 'c sufficient in number. The NRC, inspectors suggested self-reading . e' > ' dosimeters with.an intermediate range (e.g.,-0-5 rem) which could ; accumulate up to'the design doses with good accuracy of' readout and ' - m . - % - without being recharged.' .- ; 'Q. The NRC inspectors' spot checked a sample of self-reading dosimeters n - 'and noted that calibration stickers and the surveillance checklist verified that the dosimeters were calibrated and maintained as stated .in the Plan and procedures. ' ' Based on the above, the NRC' inspectors concluded that the following deficiency was identified in this area:
-
* . The number of self reading pocket dosimeters was insufficient to
- support radiation monitoring functions assigned to the TSC and
- , , OSC.' (285/8620-03) .- n 1.1.2.3 . Protective Supplies-
n / , The NRC inspectors reviewed the TSC dedicated radiological protective
fsupplies inventory and. examined the contents of the TSC closet and ~
' 4 - ' emergency kits to determine if the radiological control capabilities
met the requirements of 10 CFR 50.47(b)(8), 10 CFR 50.47(b)(11),
i:
' . ' Section IV.E.1 of Appendix E to Part 50, and item 8.2.1 f. of
- Supplement 1 to NUREG-0737. ,
t '
!
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F - 4 ,--,~,,-n-+,.~..,,,,_nn _ , , , ,-,,n,, , . - - - - , , ,-.,m .n r----------..,,,--.,m---,e.,nn v,
$ 5+: , _ . - , . - < , , ;,t w . - , . * $ _ }w Y' * . l ' g, ^ i [ < , 74 y g.s. i f ' . . f:e .. ' e- > u . , ; , ' . . ' . . . '- _ ' ?The NRC inspectors: determined thatlthe licensee had provided . ' ' " ; 7~ (respiratory protective equipment,Jprotective clothing, potassium- > vc .. . iodide tablets, and other protective and support supplies for the TSC .and.OSC. ' "? .The NRC inspectors reviewed the surveillance test (ST-RM-3) for -inventory of. emergency plan supplies and equipment and verified that $% ~ ' protective-supplies were maintained as stated in the Plan ~and , , procedures. ~ . 3 ' Based on the above, t.he NRC inspectors concluded that protective 7fv g supplies were adequate.- r. ~ . .y 1.1.-3 .Non-Radiological ~ Equipment and Supplies J .. . . . . s .i '1;1.3.1 Communications i .. ! ' , ' , The NRC inspectors compared licensee communication equipment with - -NUREG-0737, Supplement 1, item 8.2.1.g, which requires that the TSC provide reliable voice and data communications with the control room : 'and EOF and reliable voice communications with the OSC, NRC ^ Operations. Centers,'and state and local operations centers. In , , ' addition, the NRC inspectors reviewed Section H of the Plan. - JThe NRC inspectors-determined that the licensee had installed a local- private automatic branch' exchange.(PABX) in the TSC. The PABX could , be controlled by an attendant's console in the TSC and allowed 1 , emergency response personnel to communicate with the control room, < 0SC, EOF, and offsite' agencies. 'The NRC inspectors noted that the PABX system had a battery power source, and that it was linked to the security emergency generator in the event of a loss of offsite power. ! l Additionally, the PA8X' consisted ofLa. redundant computer system that Linterna11y checks the system.' The PABX system had telephone trunks going to offsite communication systems. The NRC inspectors ~ - determined that there.were 28 telephone lines available from the TSC, , of which two:had'been made available to the NRC for their use during " an emergency. ' Additionally, the NRC inspectors determined that the licensee had installed the NRC ENS telephone and licensee owned microwave communications.and a dedicated Conference Operations Network (COP) system for communicating with the control room, E0F,
i- and offsite agencies.
The NRC-inspectors reviewed the licensee's Conference Health Physics . network and noted that this communication system was dedicated to . coordinating radiological information between the TSC, E0F, and ~
E
, offsite agencies. > , The NRC inspectors noted that the licensee verified full use of the
,
:TSC communication system during the recent emergency exercise . conducted June 25, 1986, and that after the exercise, the emergency . M " P .,-t- W y - w,,,,-,y.y-w,-m - ' ' ""-7TNP 58-*
K , +p ~. ; x ~ , l wy - - h%W ;,?gi ",; ' q% w, . : +AQ ' "2 4i, g M' , I # '^ '& a .a . l) ' * , ' % ',. '15 ' p 7f\p ' , -l ; ' p; , - - - ' 9. , :%. p ( ' . " ipreparedness staff informed.the communications department of any , 1 , c (correctiveactionsidentifiedbytheNRC. Corrective' action on such '. _m , -items was sinitiated by'the communication department and verified by. Ly Ethe. emergency. preparedness group. .The NRC inspectors determined that . . N- 'the' licensee performed communication tests -as required in their . p- , ~ proceduresf and:that communication. tests were being conducted monthly. ! 19 C ' ' In" addition,. dose. assessment computer program and hardware were y i-y > z Echecked ~ quarterly:or whenever program changes were made. . . _ , , i ,The:TSC emergency communications equipment had either battery power generator backup,;or both, , J[ Based on the above, the NRC inspectors concluded that this area . W 4 appeared to-be adequate. ' , , ' ' e 1.1,3.2 . Records / Drawings' ' ' ' JTheNRClinspectbrsreviewedTSC.recordsanddrawingsagainstthe s , 1 -requirements lof;Supplementl'ofNUREG-0737. In addition, the NRC r ,- linspectors reviewed the following FCS/0 PPD procedures: G-47, " 5 ' , Control Room: Drawings";;G-62, " Control of Vendor Manuals"; and >< ,R. , <GSEP A-9,." Document Control Procedures". The inspectors observed the . . - , document storage area in-the TSC.- ' + . . ; - . , , g;; The NRC inspectors determined that the master. file of documents was ! ' , , ' maintained by the111censee's main office in Omaha and that another ' copy wa's" obtainable at several locations in FCS,~(e.g., TSC Rooms 115 ' > - , and-118). .The NRC inspectors'noted that record indexing and status . # information was available from an on-line computer system, and that , ~ -selected. drawings were available_in the TSC P&ID book or were stored on film. Vendor's manuals were separately maintained, but were ' readily available:to the TSC. By procedure, the FCS control room : drawing stick was maintained current within 25 hours. The NRC inspectors noted that other drawings, including the TSC files, were
',
maintained current'within 2 weeks by batch corrections to the film file. 'When a drawing'was-reproduced in the TSC, it was marked "for "' . information." . Administrative procedures required the person using
o the drawing to check status against the computer index prior to use.
The NRC inspectors pulled one drawing and checked it against the , > index system; two vendor's manuals requested were also made ' . available.
, Based on'the above, the NRC inspectors concluded that TSC plant - records (including drawings and vendors manuals) essential for evaluation of plant status under accident conditions appeared to be . adequate. ! , q.,
,
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' ' ~ - L ' tr? ,,:il ;1.3.3 ;Suppostl Supplies. ~ ~
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' . . . . . . - .. 7 m > sThe NRC inspectors _ reviewed support supplies maintained in the TSC M jagainst;the= requirements of Supplement-1 of NUREG-0737,Litems 8.;2.1.h - and'i and,specifically.against the inventory listing provided within , , @:4 ' Surveillance Test ST-RM-3, " Emergency Plan Radiation Instruments and t~ ,I[ . m ' Equipment."- . e . ' , . 1 ~ - - LThe NRC.insp~ectors' determined that the inventory listing provided for > 7_ ; calculators, pens, pencils, grease pencils, paper, clip boards, spare 1 ~ ;c1. ' ;"D" cell batteries, . masking tape, flashlights, and first aid kits. A' . M isurveillance, test was required.to be performed monthly. The &iW,, * ' . inspectors; reviewed the previous.6-month tests and concluded that the system was adequate. to maintain the . inventory. u. :a < ! 7 ,1 ,' 1The NRC inspectors reviewed other TSC supplies and noted that ;isopleths were readily available, as were means for data trending, an . , 7 , , adequate supply of computer paper, and a library of reference ~@<' , ; - material which included RERP/EPIP. sets, a Technical Data Book, the 7 y ' .INPO Emergency Resources Manual, 1982,.a copy of the Updated Safety ; , - , . . Analysis Report (USAR), and a set of Technical Specifications. ,, m , , Based on the above, the NRC inspectors concluded that support ,_ supplies appeared to be adequate. , [ l1.1.'3.41:.PowerSupply ~ ' 'The NRC inspectors reviewed the TSC power supplies against the ,~ . > . requirements of 10 CFR 50.47 B.8, 10 CFR 50.E IV E 9, Supplement 1 of <- ' :NUREG-0737, item 8.2.1.'a,.and the guidance of NUREG-0696,. Item 2.8. * In addition, the NRC inspectors reviewed applicable sections of the - 'USAR, the Emergency Plan,.and station drawing 4778 293 206-001 . ' - Electrical, TSC One Line Diagram" and walked down the electrical v < system in company with the OPPD Supervisor of Electrical Generation ' and the FCS Electrical Maintenance Foreman. ' ?The NRC inspectors determined thati the TSC was powered from both a !- normal and an emergency power supply and that normal power consisted of a double end feed from the 161 KV line to a single 13.8KV/480V transformer which. fed through a series of switchgears to the TSC load center (MPP-14). ' Emergency power was provided from a dedicated ; diesel- generator. Power to MPP-14 was selected by an Automatic , _ Bus Transfer (ABT). The NRC inspectors noted that upon loss of normal power, the diesel would receive an auto start and sequence - o. 4 signal; HVAC load breakers would fail open on undervoltage, while the " .others would remain closed on the dead bus. On completion of the
g , start sequencing, the ABT switch would shift to the diesel output and
power would be restored to the TSC and HVAC breakers could be closed manually. Since the ERF computer was provided with an UPS, it would ' survive the dead bus transfer to emergency power. However, during
- . power sequencing, the TSC terminals and displays would lose power.
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r * 17 ' . J c Based.on the above,.the NRC inspectors concluded that the TSC was 1provided with adequate normal and emergency power alternatives to - function- during a power l emergency. . !1.2. Information~ Management . , 1.2.1 -Variables Provided 1. 2.~ 1.1 - Rec. Guide 1.97 Variables- The NRC inspectors conducted interviews and held discussions with . licensee personnel, conducted a TSC walk-through which included 'several computer terminal demonstrations, and reviewed pertinent - documentation, including OPPD letter dated April 1, 1985, " Fort . .Calhoun Station Compliance with Regulatory Guide 1.97, Revision 2," .and NRC. letter dated June 18, 1986, "Conformance to Regulatory Guide 1.97, Revision 2." The NRC inspectors determined that, with the following exceptions, 'all applicable Regulatory Guide 1.97 variables were available in the .TSC on the SPDS and/or the ERFCS: * The. Post Accident Sampling System (PASS) ion chromatograph was = the primary means of obtaining the RCS soluble boron -concentration, in lieu of a boronometer. * Analysis of primary coolant (gamma spectrum) was measured by PASS gross gamma detectors and gamma spectrum analysis of reactor coolant samples, in lieu of a radiation monitor on the letdown line. The Condensate Storage Tank Water Level was available on demand in the ontrol room, but had to be passed to the TSC from the control room by telephone or messenger. * Containment sump water temperature was not available. Justification was submitted in the OPPD letter of April 1, 1985, but was deemed insufficient by the NRC reply letter dated June 18, 1986, item 3.3.10. * Airborne radiohalogens and particulates, plant environs radiation, and plant environs radioactivity variables were obtained using portable instruments with results passed to the TSC via telephone or messenger. All variables listed under " Accident Sampling - Primary Coolant and Sump, and Containment Air" were obtained by PASS / grab sample, with the exception of dissolved oxygen and oxygen content. The licensee stated that this exception regarding dissolved oxygen and oxygen content had been accepted by the NRC
E .
- , ... , ,_ l -
...
18 l 1 cincident to the-NRC review of NUREG 0737, Item II.B.3. However, the. letter granting the exception was not made available to the NRC inspectors prior to departing following the exit meeting. The NRC inspectors noted that the'.following Regulatory Guide 1.97 variables, although available.in the TSC on the SPDS and/or the ERFCS, were not in full compliance with Regulatory Guide 1.97 for the . reasons stated: Coolant level-in the reactor deviated from Regulatory Guide 1.97 - in that the installed level measuring instrument did not cover the full range of the reactor vessel. * .- ' Accumulator tank level and pressure instruments were not in L " compliance with Regulatory Guide 1.97 in range and environmental qualification. * Instrumentation for Component Cooling Water temperature and flow to the ESF system was not environmentally qualified. Justification for this deviation was submitted to the NRC by the licensee, but was not accepted. * Wind direction instrumentation'provided by the licensee did not -meet the accuracy requirements of Regulatory Guide 1.97, nor had the licensee provided justification for this deviation. Based on the above, the NRC inspectors identified the following unresolved items in this area: *- Accumulator tank level and pressure instruments were not environmentally qualified as required by Regulatory Guide 1.97. The level instrument range was stated by the licensee as 0-100 percent,.but did not state whether this was percent of tank volume or percent of instrument tap height. The pressure instrument range was stated to be 0 - 300 PSIG, which was not in compliance with Regulatory Guide 1.97. (285/8620-04) * Instrumentation to measure containment sump water temperature was not available to satisfy the intent of Regulatory Guide 1.97. (285/8620-05) Instrumentation for measuring temperature and flow of Component Cooling Water to ESF system components was not environmentally qualified in accordance with the provisions of 10 CFR 50.49 and Regulatory Guide 1.97. (285/8620-06) * An accuracy of 15 percent for wind direction measuring instrumentation was not available. (285/8620-07) !
- gm ', ' , .. Q]8 , , . ;." s , + l' 'Qy [ ,9 _e - . y ' . - m >- , i k ' ' % ,>; r l . , - ._ , - _ < , c yg ' , < . , . . ,. , - , - - L;. , # * ' .- The instrument forf measuring coolant level in the reactor did not have'a range from.the top of the core to the' top of the ' ^ ; : reactor vessel. 'A continuous. display of this1information was - . , ' ' . . .not'availableiin accordance ,with Regulatory Guide 1.97. This . matter >1s~being' addressed hyithe NRC as part of their review of NUREG 0737/ Item II.F.2. '(285/8620-08) * ' The licensee deviated frori Regulatory?G uide 1.97 with respect to post-accident sampling' capability inithat,the PASS did not have , the capability for dissolved oxygen or oxygen content, as required by NUREG 0737, Itom II.B.3. (285/8620-09) f2.1.2 Other Variables ' The NRC-inspectors reviewed Regulatory Guide 1.97 variables against i those available in the TSC as reported in 0 PPD _ letter of April 1, L 1985, .." Fort Calhoun Station Compliance with Regulatory Guide 1.97, - -- . -Revision 2," and-as~ responded to in NRC letter of June 18, 1986, "Conformance to Regulatory Guide 1.97, Revision 2." 2The'NRC inspectors held discussions with two staff engineers, keyed - toward the' availability of~other variables in the TSC which could be
, -substituted for_any of the Regulatory Guide 1.97 variables not
.- available. The.NRC inspectors determined that four such variables X 'were not;available.- The licensee stated that two of these variables, " ' .. dissolved oxygen and oxygen content, could be obtained using
a ; chemistry procedures. However, the NRC inspectors determined that ' "-
' there was no alternate variable which could adequately substitute for -the third unavailable variable,- that is, containment sump water
'
temperature. The fourth' variable, coolant level in the reactor, was . measured indirectly by ths reactor vessel level monitoring system which used the~ heated junction thermocouples and the SPDS. The NRC~ inspectors datermined1that five other variables were
, available'in the TSC, but did not meet Regulatory Guide 1.37
requirements for the. reasons described in Section 1.2.1.1 of this report. However, these variables did serve as substitutes for the
L '
. corresponding Regulatory Guide 1.97 requirements. These variables were: Accumulator Tank level, Accumulator Tank pressure, Component
- Cooling Water flow to ESF components, Component Cooling Water
l
, - temperature to ESF components, and Wind direction. 4 The NRC' inspectors noted that another variable available in the TSC
u . was the offsite monitoring information frc,a field monitoring teams
- via radio. Additionally, the TSC had access to the National Weather
Service (NWS) via telephone to the Omaha airport, Eppley Field.
I National Weather Advisory Service (NAWAS) information was available
to the control room, and could be relayed to the TSC. The NRC inspectors noted that the TSC also had access to emergency vendor l . s , . . - - - - - -
^ ^ ' T ;Q Y ' },
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;p y v_ ;- - p ' d - - , . .' , -20 nk.}. uy: _', . ' ' : , . , > a. - * 4 .; g.g x g,, ' pg [assistactebytelephone,-andthatacopy~of-the'InstituteforNuclear
4f j';" C ," *
. Power _ Operations?(INPO) Emergency Resources. Manual'was'available for reference'.- Evacuation time estimates were contained in the plan, 1 SectioniJ,; Table J-4.- 4 f ." y 4 . , . . .. T . Ba' sed lontheiabdveilthsNRCinspectors.concludedthattheadditional , - , - fvariablesjprovided appeared 'to be adequate.
< {1 ,11d2.1.3%?RelationshiptoFunctional'Needs~
' ., e s ' , 3 . . . . .c . - . . ' sThe~NRC. inspectors reviewed the: adequacy of:TSC variable information . 3 against:the requirements of Supplement 1 to NUREG-0737, . ' * , c. 1 . . .- -items 8.2.1.a,;h; and'k. . e p . ' : - ,. . ' ' ' 7The NRC inspectors reviewed'0 PPD letter dated April.1, 1985, the g' .. slicensee's-report of compliance-with Regulatory Guide 1.97, ? . ' ' Revision ; 2,1 requirements, the NRC letter of June 18, 1986, which . responded to the;0 PPD compliance report, and the availability of Y , : Regulatory Guide 1.97 and other variables within the TSC as - ' y' ' documented in Sections 1- 2.1.1-and '1.2.1.2 of this report. .The NRC . , < + -inspectors: observed operation of the SPDS and ERFCS terminal display / units, held discussions =with two staff engineers assigned cognizance iover the Regulatory Guide' issue, and observed walkthroughs documented 1; , ' Lin Sections 1.3.~2.2'and 3.3.3.5 of this report. m ' . . . .With the; exception of-the unresolved items documented in~ - , Section 1.2.1.1.of this report, the variables available within the t' _ :TSC' appeared to be adequate to determine reactor system integrity,' ? , ' . heat' removal capabilities, containment integrity, vital auxiliary Lc o isystem; status,. liquid and gaseous radiological waste status,' spent l fuel and in plant radiation levels, radioactive release path path rinformation' andlother information impacting the offsite dose , _ projection process. During the review no shortfall of information n 'was noted. , , _ Based on.the above, the NRC inspectors concluded that this area of :c ' relationship to functional nee 4 < ppeared to be adequate. ~ - , o , 21.' 2. 2 : Data' Acquisition- # , . ' * 2.2.1
'
' Data Co11ection Method
,- ' ;
'
. LThe NRC inspectors determined that the ERFCS was a distributed
u computer system purchased from Energy Incorporated (EI). This system
'.-
wa found to be configured as two pairs. Each pair consisted of one . computer operating in the active mode and one computer operating in '* , the ' hot' standby mode. One pair was located in the TSC and was -referred to_as the HOST computer system. The second pair was located behind the control room and was referred to as the Data Acquisition
i- , _ Subsystem (DAS). . t
4+bw www m _ - , - m-m-e-<.e-. -a w-- ,-,e-o...e,---,-me--- -,---+,.---=.wy -a r -e- e v m--r-+-~=~ <- =*r -- =m*w r- *- w t- -'--nw ~7e ""
g , , =@ % . k ? f, v.l , ? bl 'Q- y - - n; ,< , _ - ~ - pg , p y ^~ ' - gl n. 4' 21' , e_ . . . . ; - a ' C ; ;C~ 9 ^ - V ' "," "pn ' P- *~ p.; ' - td db W J,L q ' 4# . <~ ' , ' , )The:HOSTcomputersystemcontrolledtheDAScceputersystemwhichin turn' acquired plant sensor data (analog and digital) and Qualified ~ ,. - < : Safety. Parameter Disp 1ay; System (QSPDS) dual computer-system . ud , . < sdata:(class 1E). The HOST computer system consisted of two . $y .-MDDCOMP'7870 computers and.various peripherals. -The DAS computer . .M ? ; system consisted of two MDDCOMP 7821 computers with three MODCOMP
V" ^
MDDACS~III.multiplexers.for sensor input / output:and signal
[1 -
conditioning.$ The DAS computers were used to convert raw data into
@'. floating point engineering units. 3 >
g'i , @ ,, . . A4 .J :The'NRC' ins @pectors noted that the HOST-computer system received Me ~ meteorological data input as_well~as input from the DAS commiter y Mg ' ' ' systems. .All'operatorfinput/ output was controlled by the HOST ERFCS. .'fy ; TheHOST.wasialsojpsedforalldataarchiving. m. . _ .7,w .. g The'NRC inspectors? evaluated the111censee's compliance with & ' itemL8.2.1h of Supplement 1 to!NUREG-0737 in regards to the - : methodology:by which sensor data inputs from onsite and offsite ' locations:were'receivedt *f., ~ . > LTheNRCinspectors\reviewedtheHabareandSoftwareDescription - Manuals provided to OPPD'by EI:esi.ed August-1985, ERFCS System ' l Description ~II-10, Revision'2,iQSPDS Communications (QSP) Software s ' .. Design Document, LRevision L 0,- and ; Interface Requirements for . QSPDS/ERFCS Data, Communications, Revision 3. In addition, the NRC y inspectors conducted interviews and held discussiotis with members of
n '
the licensee: engineering a'nd computer programming' staff. & @ % b; .: .l' :s .h ( OL r- . The Meinspectors determined that data were collected by the active j . ' .ERFCS HOST damputer by serial input from the. meteorological tower computer system and the active DAS computer system. The DAS received '
- ^ '
' . . ~ ~approximately:400 inputs ea'ch 'from analog and digital sources. This , numberuiricluded the QSPDS ' class 1E inputs. The system was expandable
- pN t& .'
to over 2000 inputs each from analog a'nd digital sources. All inputs
M '
. were direct except for the serial ASCII input from the dual QSPDS (a t subset:of the Inadequate Core Cooling Instrumentation (ICCI) computer
'
~y system). The HOST received the DAS input over one of two high speed fiber optic' links. . Based on the above, the NRC inspectors concluded that data collection methods. appeared to be adequate. " ~ _ - 1i 2.'2. 2 Time Resolution
n
g, ,The NRC inspectors evaluated licensee's compliance with item 8.2.1.h <" of Supplement 1 to NUREG-0737 in regards to the time resolution of # sensoi' data which are'available to the ERFCS. The NRC inspectors reviewed the OPPD ERFCS Software Overview prepared by EI; OPPD number 1283. In addition, the NRC inspectors conducted an interview
7 ,
and held; discussions with a system program analyst and received a - demonstration by Sirin the TSC. .
p I
"
g __ : - - , s 2.- _'w . , OM [ ' ' . 't 3. , , , 37 []( d;" < ~ ~N _.- ,s.. ' 4 " 33$ %' . . 22- l g9 ~j'; < ' , ~ gyh a , _ , -2 G S n :%dw% + 1 The NRC inspectors determined that the maximum scan rate of user , m ~? ? , selected-signals was 0.5-seconds / scan. Normal' sampling was t &4- U ~one'second/ scan for digitalfsignals and l'to 60 seconds / scan for Yf >>s- " ianalog: signals.(depending-on sensor type).~. QSPDS signals were . scanned'at-the: rat'e of'3' seconds /sc'an. The history and transient ' ' , A. -- . _ : data 1 files.1 stored by the. system for'all points:was as follows: - , , ~ u. ~ . s (Tygg Duration. Scan Rate- Data Type (analog) , ' . a ,. . . #3 ' , n General ( .22 hr :30 sec' raw binary ,, x 7 Purpose' c m ~ , , ' '~ M. x ,%.[, Prevent- : 3, min 1 0.5 sec' raw binary ~ ' ~ s30 sec engineering units ' U ~ , ;2 hr +. . - m. f . - - ! , s ,N Postevente 3 min:- . :0.5-sec raw binary. , s ^ ' , , /30~ min. . 10 sec engineering units % ,- a, y" , 24.hr 'l min engineering units , x4 -< ; - 114 day :10 min - ' engineering units - , . . , %< ' ' , , ' ; Based on the above,-the NRC inspectors concluded that time > resolution methods. appeared to be adeq'uate.' ' _ " ' '", D1. 212. 3' Isolation : -p : ' , . ' . . , _1 - The NRC in_spectors. reviewed the isolation provided between class 1E , , systems: and the downstream non :1E ERFCS against the requirements and ; ' ~ standards.of GDC-24:and Supplement 1 of NUREG-0737, item 8.4.lc. The .NRC inspectors examined the TSC, operated the ERFCS terminals and ' ' displays,1 inspected the computer mainframe room, and' reviewed ~ n - -applicable sections of the USAR and several letters on the general - i subject'o_f= isolation. Key letters are i_dentified below. ' *' 'The ERFCS installation used several different devices to achieve isolation-between the safety systems and the downstream nonsafety related equipment. These isolations were evaluated by the licensee
,
Lin OPPD letter to the NRC dated December 7, 1984. By letter dated " June.7, 1985, the NRC forwarded the safety evaluation report which Lapproved-the use of certain isolators, including fiberoptic cables, '
,
- butnoted that insufficient information had been provided to evaluate
. - .nine particular type isolation devices used by OPPD. NRC directed
the licensee to submit additional information for confirmatory review. By a series of letters and phone calls which culminated in a . letter dated January 7, 1986, OPPD acknowledged the need to replace
. I
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~ . s , T ;23 4 S' ' . , , . . ,; w _ 32 ,- ; u ^ '?* ' , I those,isold.iondevicesandrequestedextensionofthecommitmentto (O'l W , .the end-of:the 1987 refueling' outage. NRC granted that request by - . - (letter' dated January,9,- 1986. , - ' ' ~ 6 :-Based on . 2 the Ebove, the>following unresolved item was identified: w ' , - ~ ^*' ,. , !Isdlat' ion between safety systems'and downstream nonsafety '_' related equipment will not be completed until the end of the 1987 refuelingioutage. -(285/8620-10)~ - N _ fl. 2.~ 3 Data Communications C , 3 * , . , '1.2.3.11 Cap'acity; . ' ,' 4The NRC--inspectors evaluated licensee's compliance with item 8.2.1.h - ' of. Supplement I to NUREG-0737 in regards to the data handling ' capability of the data communication links. The NRC inspectors -reviewed.the:ERFCS System Description II-10, Revision 2 and Interface "? a _ Requirements for?QSPDS/ERFCS Data Communications, Revision 3. In 9 ? : addition,s.the inspectors conducted interviews, held discussions with ( ~' - - ' Laembers of.the licensee engineering and programming staff, and examined the hardware associated with the_ERFCS in the TSC, CR, and ;- -the DAS behind the CR. , ~ :The:NRC inspectors determined that transmission of information was~- , ' _ ' accomplished serially (except for the meteorological data which was . time multiplexed and transmitted as a frequency) and was compatible , through a RS-232C. serial connection'at the user interfaces. A table of transmission rates between equipment at various locations was as -follows: .To/From Rate Equipment- ERFCS HOST (kilobaud) e Typ_e CRT monitors Within TSC 9.6 Hardwire w/ function
F ,
boxes Matrix Printer. Within TSC 4.8 Hardwire _CRT monitors' CR 9.6 Hardwire w/ function boxes ' Matrix Printer CR 4.8 Hardwire ' Line Printer CR 19.2 Hardwire
- CRT monitors E0F 4.8 Microwave w/ function boxes and MODEM . DAS Behind CR 2,000 Fiberoptic _ . . _ . _ - . _ _ . . . _ . _ _ . _ _ - . _ _ _ _ . . _ _ _
- p- . . g ;e ,. ' [ <; -24. L i). " ' . DAS to QSPDS- . 19.2 Fiberoptic . ' Meteorological. -TOWER NA Fiberoptic e~ 4 Based on the above, the NRC inspectors concluded that the capacity of the data communication lines appeared to be adequate. , 1.2.3.2 Error Detection . .The NRC inspectors evaluated data communications error detection - capabilities against the requirements of item 8.2.1.h of Supplement 1 to'NUREG-0737. cThe"NRC~ inspectors reviewed the QSPDS Communications (QSP) Software Design Document, Revision 1.0 and ERFCS System Description II-10, Revision 2. The NRC. inspectors conducted interviews and held L discussicns with members of the licensee engineering and programming staff. The NRC inspectors determined.that QSPDS data were transmitted with a parity' check and a packet checksum (an exclusive OR of all ASCII data bytes). DAS data were transmitted to/from the ERFCS HOST through a communications controller in packets using MODULO 16 checksum. No -error detection was performed between the user interface color monitors and' function boxes. The CRT screens were updated at a three second rate, so errors could be detected with reasonable quickness. Based on the'above, the NRC inspectors concluded that data communications error detection methods appeared to be adequate. 1.2.3.3 Transmission Between ERFs The NRC inspectors evaluated licensee's compliance with item 8.2.1.g of Supplement 1 to NUREG-0737 in regards to transmission of information between ERFs. The NRC inspectors reviewed the ERFCS System Description II-10, Revision 2 and the undated ERFCS training instruction handout sheets. Additionally,'the NRC inspectors interviewed and held discussions with the licensee engineering programming staff. The inspectors also examined the CR, DAS location, TSC, and EOF to make observations and get hands-on experience. The NRC inspectors determined that data trar.4 mission between ERFs was under the control rf the ERFCS active HOST computer system located in the TSC. Two pairs of data links were available so that loss of one link did not prevent data from being received by the active or hot. standby HOST if switchover occurred because of a failure,
f
_. - , , ' - % ;( . z,; < , . , , *l 'w : , a c ,,. ., - > . , ~ ,77 25 * .l; f :, - ll - ^ s > ~q 3. ' ' - , , E 4 ,g - . . _ . .. q. - 4%~ SjN " s The inspector'^noted s that the CR had four hardwired serial links,to ^the TSC for full duplex operation with two operator consoles, one M M _' - - cline printer,.and.one dot matrix alarm printer. The TSC had six hardwired serial-links'to five operator consoles.and-one line M. ., , Y _.o. ' printer. _ ;In addition, the NRC inspectors noted.that'the EOF utilized three [' W ' channels ~over a microwave link for two operator consoles and one TI ~ 6 1' ' Omni-800 dot matrix printer, and that multiple operator consoles.in i' ' ioachtERF location ~ enhanced' reliability and provided additional . 4 resource tools for personnel during emergency conditions. .s s :: Additionally,: the two CR and four of the TSC operator consoles.
?8 '4"
contained one megabyte of memory each to hold graphic backgrounds for , typical user response tiseiof 1 to 2 seconds.' W- Based on the above, the NRC inspectors-concluded that the methods for :g ; & ' transmission between ERFs appeared to be' adequate. s ~IJ2.4 " Data Analysis h'. fl.2.4.11 ~ Reactor Technical' Support im .The:NRC inspectors reviewed TSC reactor technical support against the , requirement of 10 CFR-50.47(b)(4), 10 CFR 50 Appendix E, , paragraph IV.E.8, and item 8.~2.1.h'of Supplement 1 of NVREG-0737. The tinspectors' examined the TSC, reviewed FCS 01-ERFCS-1 "ERF Computer * System"L(an operators ~ manual for that system),-operated the TSC ERFCS ; equipment for approximately 2 hours,-and observed the-walkthroughs a " reported.in Section11.3.2.2 of this report. '
, The NRC inspectors observed that the primary source of reactor plant
data in the TSC was the ERFCS~which included the SPDS. Using that - s system, the operator-may call up top and'sub level SPDS displays and ' other-ERFCS non SPDS data which may be displayed as real time information or in graphic form, usually as time trended data, but in selected instances, with one variable as a function of another. - Boundary _and design values were displayed to assist the operator in l ' detection of deviations or excesses. For selected' systems, system ~s tatus diagrams were available and included position and operating
m status of key valves, motors, and pumps. [
* !Although the system was just declared operational on April 30, 1986,
' -the TSC user staff appeared to be conversant with system operation. Based on the above, the NRC inspectors concluded that reactor technical support within the TSC appeared to be adequate. ;
'
, i 4 i 1 - l . 1
'
% 4 3 1'-
' . . W " . : 26 ' a; 4 - 1. 2.'4. 2 TDos'e Assessment , ~ The'NRClinspe-tors reviewed licensee's dose assessment' capability
W_ -
against requitasents in-10 CFR 50.47(b)(9) and Appendix E, Paragraph'IV.E.2, Supplement 1 to NUREG-0737,- item 8.2.1.h,-and , Regulatory Guide 1.97. .-The NRC inspectors also reviewed the ' Fort Calhoun. Station meteorological system for acceptability as part .of;the. emergency response-facilities required by Section 50.47(b)(8) ~ , .and-Appendix E, paragraph IV.E.2 of.10 CFR 50 and Regulatory -Guide 1.97, Revision 2. Criteria for evaluation were based on ; specifications set forth in Section 2.4 and Table 2 of Regulatory e ' . Guide 1.97,-Regulatory Guide 1.23, and ANSI /ANS 2.5-1984, "American ' ' , National: Standard for Determining Meteorological Information at 4 Nuclear Power Stations." ' LThe'NRC inspectors;i~nterviewed licensee staff responsible for development ofLthe dose assessment program and.its associated -training. ;In addition, the inspectors reviewed FCS Emergency Plan < . Implementing Procedures (EPIPs) and other references describing the licensee's dose assessment program,'such as: EOF-6 "Onsite/Offsite " - ' .. Dose' Assessment,":Section I of the emergency plan, the Emergency ' - Assessment'of Gaseous and Effluents (EAGLE) computer code .(August 1984), the-EAGLE Program User's Guide and EAGLE Program Technical Manual, both dated December 1984. The NRC inspectors checked equipment used to perform dose assessment-and had the , licensee perform sample calculations using the EAGLE Program. In - ' conducting the meteorological portion of the review,'the inspectors .~ examined the' meteorological measurement site, examined the instrument installation, reviewed instrumentation maintenance and calibration > . records, interviewed an I&C technician responsible'for instrument . ' maintenance and calibration, interviewed the data analyst who - processes-the meteorological data and maintains the meteorological ' . * ' data' base, and monitored the performance of the instrumentation during~several periods. 'The NRC inspectors determined that.the licensee had a computer . program designated EAGLE and'a manual method for calculating doses. The manual method was the primary method used in the control room and was available as a backup method in the TSC and EOF. For several windspeeds, the manual method was based on graphs of dose rate versus gross counts per minute (cpm) readings on process monitors. These graphs were generated using the EAGLE program. Inputs into the manual method included information pertaining to: meterological data, stack flow rate, process monitor readings, and duration of release. 'Results of a sample problem calculated using the manual
, -
method and the EAGLE Program were adequate. The NRC inspectors noted that the EAGLE program was written for the licensee by a consultant and the user could calculate doses using _. _ _ _ _ . . _ . _ _ _ _ _ . _ - _ _ -- _ , . _ _ _ . _ . _ - . . _ _ . _ ._. __.
._.-_... - . + ~ - ., - .- -. -- - - _ - .9F" w $- 5 :{.g q y 4% . r C/ M. _ yi!J 4 y - ' ( _( , '! + . .- c:I o '27. ~ . .
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i g-M 7 P. .. , "' 7 ; ?a either a straight.line Gaussian atmospheric dispersion model or a 4 ; - s -segmented plume. atmospheric dispersion model. The program was run on ;the' licensee's mainframe' IBM computer located in the corporate office &c * "f * Efrom a Tektronix 4105; terminal in the TSC. Major features of the 1model included. ground-level-or' elevated release, plume rise, building ' u, s
g# %g N 2' wake effects, dry-deposition, radioactive decay during plume. transit,
& . W a default-radionuclide' mix, a tabular' option (e.g,, comparison of A p 7 dose. projections with field team readings), and-a graphics option
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> J(e.g/,: plume trajectory plots). We_ ' + . q s , :The NRC inspectors noted that during the TSC-walkthrough, the 4 ' licensee performed dose calculations every 15 minutes per ~ 7',c , ' procedure EOF-6,Susing the straight-line Gaussian option. Interviews ;yE -~l- + .with: licensee s_taff revealed that the more sophisticated segmented ' plume option'was never used during drills or exercises because- g ' results were.not.in agreement with models used by the state of ~ S" < Nebraska. "The NRC inspectors noted that the segmented plume model 1 'would'be more appropriate for continuing assessments in the TSC and
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, .E0F. fThesinspectors also.noted-that the tabular and graphics option . ' .of'the-EAGLE program were never used because requirements to perform ~ ~ dose calculations every 15 minutes:(even..if conditions remained
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+ , stable) did notiallow time enough to use these options.
p
V4 The' inspectors determined that there was no s'pecific procedure " w , dedicated to source term determination. .Such procedures were ' embodied in TSC-8, " Core Damage Assessment Procedure," and in E0F-6, *
y, ,# 7
- - :"Onsite/Offsite Dose Assessment.": The core" damage' assessment
h '
:b, procedure was' based on Combustion Engineering Owner.'s Group (CE0G) Task"467- Report, '.' Comprehensive Procedure for Core Damage %* , . Assessment," and employed four basic methods: . hydrogen in ' J"_ containment,: PASS, radiation in containment, and core exit . ' thermocouples. ' * 'a NheNRC'inspectorsnotedthattheEAGLEprogramconsideredthree ~ - 4 , ' release points: ' Auxiliary Building stack (which included Auxiliary 4 Building, ventilation exhaust, containment atmosphere purge exhaust, , containment hydrogen purge exhaust, waste gas released from Radwaste , , . Dispos'al System, and condenser off gas), Condenser / Main Steam, and "' i Containment Leakage. Gross source release rates were divided within - - itbe model into noble gases, iodines, and particulates. ot^ j- For. ventilation stack releases the following monitors were specified: ' ; Noble Gas: RM-062 (if on scale) " > RM-052 (if RM-062 out of service) RM-063 L,M,H (if RM-060 off scale high or out of -W service) . ' . _ For airborne releases from the main steam line, RM-064 was used. For
," ' .
' assessment'of airborne releases from containment design leakage, the highest reading area radiation monitor was used. s 4 ' . . - . . _ _ _ _ . _ . . . . . _ - . _ _ _ _ , - _ . - . _ . .,.m ,, . . _ - . __ __,_.._.-.m__,-__,-__.,,.,_,. _ . _ _ _ _ _ ,
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' . , . by ', The:NRClinspectorsnoted'thatthe'monitordatareadings-incountsper 3,f ~ iminutes-(cpe) were entered into the EAGLE Program. Monitor readings he , * ,were obtained:on the ERFCS. The EAGLE code then converted the. cpm - reading,to a' concentration in microcuries per cubic - >" " : centimeter (pCi/cc) based on current calibration data. -A release . ' irateLin Ci/sec'was_then determined based on stack flow rate. The nuclide six considered in the release'was predetermined for the EAGLE , y code and considered the following nuclides: - : Kr- (83m,- 85m, 85, 87,' 88,2 89) . ' - 'Xe- (131m, 133m, 133, 135m, 135, 137,' 138) .a - . > .I;-:(131,132;,133,-134,135)' , .Ce .144 , ' .- Co .60- Cs< 134, 137, . . .Mn- 54.- - -Ru-1106 4 ETe- 132- : . . 4The:-inspectors'noted that the EAGLE program had the option to allow , input'of specific nuclide mixes based on an isotopic analysis of a . ' stack grab sample or based on PASS (i.e. containment atmosphere) ' results... Isotopic-dependent release rates from a set of several ' ' simulated reactor ~ accident scenarios were stored as an established data file in the EAGLE-program. These scenarios were: Loss of _ .' Coolant Accident,' Steam Generator Tube Rupture Accident, Contaminated ' -Fuel Handling Accident, Spent Fuel Pool Handling Accident, Hain Steam JC. Line Rupture Accident, Control Element Assembly Ejection Accident, : and,-Gas Decay. Tank Rupture Accident. The'way the EAGLE program was ' currently set-up did not allow usage of these seven scenarios. _ s' ' .The NRC inspectors noted that section 3(a) of Fort Calhoun * , procedure EPIP-EOF-06-H directed the user in making dose assessments, to treat- all atmospheric releases as ground-level releases. This ' . precaution is consistent with NRC guidance for treatment of releases L from stacks when the stack height is less than 2.5' times the height . of ' adjacent buildings. The EAGLE code treated releases from the ventilation stack as elevated, rather than ground level under low wind speed conditions. Therefore, the NRC inspectors concluded that the EAGLE code was inconsistent with EOF-06-H 3(a) and NRC guidance . with respect to determination of release heights. ' The NRC. inspectors determined that the atmospheric model computed ' , < concentrations assuming that: (a) the diffusion in the vertical -direction was unlimited, or (b) diffusion in the vertical direction was limited and'the plume was uniformly mixed. The EAGLE code made a step change from one assumption to the other. The NRC inspectors m concluded that there would be an increase in dose estimates when this change occurred and consequently the dose just prior to the change would be underestimated. f . . '1 . - . . , . . - __ - .. . . . . _ . _ . - _ _ . . . _ . . _ - . _ _ _ _ _ _ . _ _ . _ _ _ _
.. . 29 The NRC inspectors noted several hardware problems with the EAGLE program primarily associated with printouts of information and information transfer to the states. For example, the licensee had no way to provide a screen printout just in the TSC or E0F without sending the information to the states of Nebraska and Iowa. In addition, there was no provision to print out tabular or graphic outputs. .. The NRC inspectors reviewed the technical basis for the EAGLE program and concluded that the basic equations used to calculate whole-body, . thyroid inhalation doses, and the dose conversion factors used in the EAGLE program were adequate. The inspectors noted that the radiological dose calculations took into account the radioactive decay during plume transit but did not consider radioactive decay between the time of reactor shutdown and the time of release. The inspectors concluded that this was a conservative assumption but would over-estimate dose if the radioactive material remained bottled up in containment for some time prior to release. The EAGLE program ]ll also failed to consider daughter ingrowth after reactor shutdown. The NRC inspectors also noted that the EAGLE program used a " structure shielding factor" which would result in a non-conservative result and would tend to underestimate radiation doses for individuals not sheltered. ' The NRC inspectors determined that the EAGLE system did not have the capability to calculate ingestion pathway doses. The licensee stated that the States performed these calculations, and that ingestion pathway data could be inputted into its code from routine operations. : The NRC inspectors noted that a systematic verification of the EAGLE a program had not been completed. The licensee had compared hand calculations to EAGLE results for the straight-line gaussian options. . Thyroid inhalation dose comparisons were acceptable; however, whole body dose comparisons differed by 20 percent. The inspectors noted that verification of the segmented plume model had not been initiated. .. The NRC inspectors determined that the licensee had compared EAGLE program results (straight-line Gaussian model) with the models used by the state of Nebraska and Iowa. Dose comparison were generally : within a factor of 2-3 except for low wind speed scenario where the EAGLE program used an elevated release. The NRC inspectors noted that there were no records from which ' meteorological instrument system performance could be readily '- determined. Meteorological data submitted for assessment of the consequences of routine releases included data from offsite sources for periods when the onsite data are unavailable. The inspectors also noted that, while hourly average meteorological data were - m
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' ,f , . ' ~ ' 30 . 7 ' - routinely' available in th'e control room, all other readily available ~ s . displays of meteorological data presented only instantaneous values , .' -which were of little value in dose assessments, or for making } , protective' action recommendations. # ' 'The'NRC' inspectors' determined that in the dose assessment procedures - ' there were no' warnings that the meteorological data displayed on (dials in the control room and on computer weather displays were instantaneous values, and as such were inadequate for use'in dose , ' assessment or for taking protective actions. Regulatory Guide 1.23 and ANSI /ANS-2.5-1984 indicate that meteorological data should be averaged for periods of at least 15 minutes. Further, 15-minute averages of meteorological data have been accepted as appropriate for use in emergency response dose assessments. -The inspectors noted that the operations' staff in the CR and the ERF - computer were the only sources of meteorological information listed in;the procedures, and that there were no provisions in place for obtaining meteorological data from non-0 PPD sources or for selecting default values in the event of failure of the Fort Calhoun meteorological system. Finally, the NRC inspectors noted that there , was no mention of the availability of or procedures for obtaining historical meteorological data or data trends. The NRC inspectors concluded that EOF-06 dealt only with dose assessment during a release and disregarded potential releases. As a result, the EPIPs did not provide for dose assessment and prompt protective actions when a release may be imminent but not yet in progress. Based on the above, the inspectors concluded that the following dose assessment deficiencies were identified in the TSC: - Meteorological data, appropriately averaged for use in dose assessment, were not available in the Control Room, TSC, or E0F. (285/8620-11) * The dispersion model utilized by the licensee inconsistently treated releases from the Auxiliary Building stack as elevated .in some atmospheric conditions and ground-level in others. (285/8620-12) * The licensee failed to use a sophisticated realistic model in dose assessment calculations (i.e. the segmented plume model) and made no use of tabular and graphi: options available through the EAGLE program. In addition, the EAGLE program verification was not completed. (285/8620-13) , * Readily available records to evaluate historical data recovery performance of meteorological instrumentation were lacking. (285/8620-14)
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- l- ' ' :31 - l ' l. ~ ; .y . -' ' Procedures for dose' assessment failed to consider' scenarios < c , ' :where a release of radioactivity to the environment was imminent
@ '
(285/8620-15). , , '@ L*~ : Procedures lacked adequate guidance for obtaining meteorological ' ' ' ' .f idata in the event of a partial ^or: complete failure of the ERFCS _ or of the Fort Calhoun meteorological system. (285/8620-16)~ . .. ~' _, 1.224'3 . Central Processor Capability , i The'NRC' inspectors evaluated central processor capability against the' , 1 requirements:in item 8.2.l.h of Supplement 1 to NUREG-0737. -, ' 'The NRC' inspectors reviewed the Hardware and Software Description < - Manuals dated August 1985 and ERFCS System Description 11-10 Rev. 2, * , conducted interviews, and held discussions with licensee engineering - e and computer programming staff. . 4 The NRC' inspectors determined that each of the two central processor . equivalents (the'ERFCS HOST M00 COMP Classic 7870 computer systems) .had two megabytes of interleaved memory each. Additionally, the use of common shadow memory (redundant data memory) assured no loss of . current data when a single computer failed. Direct memory access to '. .the OAS M00 COMP 7821 computers minimized central processor unit (CPU) overhead. A 64 bit floating point unit enhanced throughput for , i 1 making calculations. The two DAS computers had one megabyte of ' Laemory each which comprised the intelligent frontends to the HOST - computer. All computers used 16 bit data bases. The NRC inspectors noted that approximately 60 percent of the plant ' processor signals had been moved over to the ERFCS. A simulated test during an emergency indicated that'the QSPDS requirements were ;approximately a 50 percent load on the active CPU. Moving all of the -plant process signals.over to the ERFCS was anticipated to increase ~the CPU load to a 60 percent duty cycle. The NRC inspectors determined that each HOST CPU had a 67 megabyte 3 .and a.256 megabyte dual' ported disk for increased reliability by parallel history file storage, that two 9-track 800/1600 bpi magnetic tape transports provided the potential for future archiving on industry standard magnetic tape, and that three line printers provided hard copy output for program development and a local _. teleprinter provided a log of system status. " < ~ System software operated under MAX IV with FLIC, as the vendor supplied application program which the licensee modified for system enhancements. MAXNET provided communications between the TSC HOST and DAS computer systems.
~ jf% - , %x < 'J ' , . . 1 ( p ', ' , , % sT ' ' . J 32 j . i j= . ; ' , .The_ CPU operating system software operated with a combination of i *- ' _ ^ hardware and software multi-level priority interrupts. The software Lpriorities were user programmable. - Based on the above,.the"NRC. inspectors concluded that the central - , ; processor analysis capability appeared to be adequate. ' :1.2.5 Data Storace , -1.2.5.1 -Storaos capabilities The NRC' inspectors evaluated data' storage capabilities against the requirements of' item 8.2.1.h of Supplement 1 to NUREG-0737. 2 .. - . , . < The NRC inspectors reviewed the OPPD ERFCS System Software Overview m y and ERFCS System Description II-10 Rev. 2, conducted interviews, and - held discussions with members of the licensee engineering and : programming staff. ' . . The NRC. inspectors determined that the historical data files on disk ' required approximately 20 megabytes of storage space, and that the ' addition of transient data and general purpose data files increased ' . j the storage requirements to approximately 60 megabytes.- This was ; compatible with the 67 and 256 megabyte disks ~available at each CPU _ for. data storage.and' application-program requirements. . Based on the above, the NRC inspectors concluded that the data , storage capabilities appeared.to be adequate. l
. 1. 2. 6 - ' System Reliability and Validity !
1.2.6.1 Validation and Verification The NRC inspectors evaluated verification of the models, system , 1- reliability and validity against the guidance of Section 1.5 of NUREG-0696. . The NRC inspectors reviewed the results of the Site '. , Demonstration Test and the Operational Availability Demonstration
t'
Test. ~
>
The NRC inspectors determined that minor problems had been discovered
F by the licensee and corrected prior to the conclusion of the hardware
and software site demonstration tests. Revisions to hardware '
-
configuration documentation were lagging behind the "as is" ' configuration, although the required information was obtainable. Standing Order 0-32A established procedures for maintaining software
-
- '
documentation. The verification test was conducted with the vendor acting as the demonstrator and the licensee acting as the witness.
. Based on the above, the inspectors concluded that verification of the
computer system models and validation appeared to be adequate.
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71 LThe'NRC-inspectors. evaluated the computer based system against the p
, , * . guidance of-Section 1.5 of NUREG-0696, and the requirements of
U
'10 CFR-50.47b(8) and (9). * , < .The NRC:. inspectors held interviews and discussions with licensee , ~ Instrumentation and Control Group and Reactor and Technical Services _ : personnel. Additionally, the NRC inspectors reviewed procedure OPPD-1-907, :' / 4 . Revision 9, " Operational Acceptance Demonstration (OAD) Test." - The NRC inspectors _ determined that an operational reliability test :was conducted over a 6-month period (June through November 1984). ~1 The methodology for calculating the unavailability resulted in a ' ^ 99 percent to-100 percent availability. Independent calculations by
m .. -
, the inspectors (which;did not subtract out hold time) resulted in a
f *, v 98 percent to_99 percent availability.
,
' -
- The NRC inspectors noted.that formal maintenance logs on the ERFCS ^ have-not been kept since the conclusion of the OAD test, and subsequently, system availability had not been calculated and recorded. ' The inspectors noted'that the ERFCS received power through a UPS unit ' ' .with battery backup and should normal power be lost, the batteries ,- - would take over.until the backup diesel generator came on-line. Based on the above,~the NRC inspectors concluded'that the following deficiencies were identified in computer cased systems: -* * Formal maintenance logs on the ERFCS have not been kept since ; the OAD test concluded in November 1984. (285/8620-17) * System unavailability calculations have not been made and recorded since November 1984. (285/8620-18)- 1.2.6.3 Nanual Systems LThe NRC inspectors reviewed the reliability of manual systems against the guidance in NUREG-0814. The inspectors examined the emergency plan and EPIPs, examined the data forms stocked in the ERFs, observed a walkthrough, conducted interviews, and held discussions with licensee personnel. The NRC inspectors determined that the data entering the TSC were , recorded by a control room data collector on standardized forms and transmitted via the control room communicator to the TSC where they were recorded on the same standardized forms. These forms are identical (FC-194) or nearly identical (FC-195) to the principal status boards. These forms, and other message traffic, were logged by a clerical assistant. Technical data were monitored by technically qualified personnel (e.g., engineers) to identify
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v
34 discrepancies between variables at a given time and within variables over time. Data were retransmitted to the E0F using the same standardized forms on both ends of the link. Based on the above, the NRC inspectors concluded that the reliability of manual systems appeared to be adequate. 1.2.7 Onshift Dose Assessment 1.2.7.1 Dose Assessment Proficiency The NRC inspectors reviewed dose assessment proficiency against the requirements in 10 CFR 50.47(b)(9). The NRC inspectors interviewed licensee staff in charge of developing and performing dose assessment and reviewed procedure EPIP-EOF-6, , "0nsite/Offsite Dose Assessment" to determine dose assessment proficiency. The NRC inspectors determined that the licensee had a computer program (EAGLE) and a backup manual method to compute doses (see Section 1.2.4.2). The manual method was available in the control room, TSC, and EOF. The EAGLE program could be run in the TSC and EOF. The inspectors noted that through the use of each method, the staff was capable of calculating doses at the site boundary within 15 minutes. These doses would be used to determine EALs. The NRC inspectors noted that procedures OSC-3, " Notification of Unusual Event Actions"; OSC-4, " Alert Event Actions"; EPIP-0SC-5, " Site Area Emergency Actions"; 05C-6, " General Emergency Actions"; and EOF-6 included dose assessment instructions for backshifts. The inspectors determined that the shift chemist or shift health physicist would be tasked to perform the initial dose assessment in the control room using the manual method described in procedure E0F-6. The inspectors noted that the manual method included dose rate graphs which were developed using the EAGLE system. Comparable dose projection results could be obtained with either method. The inspectors commented on these methods in Section 1.2.4.2. Based on the above, the inspectors concluded that the licensee's program in this area appeared to be adequate. 1.2.7.2 Dose Assessment Technical Adequacy The NRC inspectors reviewed dose assessment technical adequacy against the requirements in 10 CFR 50.47(b)(9). The NRC inspectors interviewed licensee staff in charge of this area and reviewed procedures EOF-6 and OSC-1, " Emergency Classification."
p m - - - _ _ , , . } ym y,n r - , - -- . .- s , . - ; ,- :!j ' ' 35 m , , ' , , ! s :); , y The NRC inspectors determined that the shift chemist or shift health- # , ' physicist ~would perform the initial dose assessment in the control %, * ,' :reen'(see Section.1.2.7.1). :The licensee indicated that dose s . assessment.would.be given a high priority during an emergency and L ,' # . ,, either tho' shift chemist or Health Physicist would be available. * !The. inspectors.noted that both the EAGLE computer system and the " . . Lmanual dose calculation method would provide doses at the site boundary upon which EALs were based.
'
~ : Based'on the above,'the inspectors concluded that the licensee's N fd 1 program in;this area appeared-to be adequate. xr . l.3- Functional Capabilities-'and Walkthroughs - 37 , ' ' - 1.3.1 - ;0perations ! ' -
. ,
. - 1.3.1.1- Ornanization , :The' NRC inspectors' compared licensee organization and staffing l l requirements with the regulatory requirements in 10 CFR 50.47(b)(2) ~ fand NUREG-0737, Supplement 1, item 8.2.1.a. The NRC inspectors
'
. reviewed Sections H, B, and M of the Plan. . . The' inspectors determined that the TSC staff interface with other emergency response facilities was described on Figure B-2 and that # i ' E Figure M-1 listed the emergency staffing requirements for all ERFs. Emergency organizational elements listed in Section H.1.4 to the TSC :
.
included the following titles: Site Director, Technical Support, Manager,' Technical Support ~ Supervisor, I&C Support Coordinator, t 1 Engineering Mechanical Support Coordinator, Core Physics Supervisor, ' Thermal' Hydraulics Coordinator, Procedure Training Supervisor, Security Admin _istrative Supervisor, Health Physics Chemistry I - Supervisor, and Radwaste Coordinator. Each functional group had a " primary and an alternate person assigned. Based on the above, the NRC inspectors concluded that the , organization for operations appeared to be adequate. See
- - Section 3.3.3.5 for further discussion.
p . 1.3.1.2 Stafflng .
The NRC inspectors compared licensee staffing requirements with regulatory requirements in 10 CFR 50.47(b)(2) and NUREG-0737, Supplement 1, item 8.2.1.a.
.
The NRC inspectors noted that the TSC was activated during the
t June 25, 1986, annual exercise. The licensee staffed the TSC with
- persons listed in Section H of the Plan. The TSC staffing
,- commitments were found to be adequate. L i a
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4 : . - - : v . , s . > g+ , S3 h f" ' . Based on the'above the'NRC inspectors concluded that staffing for , ; ' ,g. the!TSC appeared-to be adequate.' N < . fl.3.1.3 JActivation , ^-, ' The_NRC inspectors reviewed NUREG-0737, Supplement 1, item 8.4.1.i f and Table 2 to ascertain the adequacy of licensee's activation and - timely personnel' response-to the TSC. lThe NRC11nspectors'noted 'that"the TSC was' activated during the ; . June 25, 1986, anrual exercise, and that emergency response personnel
e were called out unannounced and responded to the TSC at the alert
. : emergency class.' The site director assumed command of the emergency m'" ., 'within 1. hour and 9 minutes after the declaration of the alert. The ,, ' : TSC had been fully. staffed prior to the Site Director assuming .E control.- _ ' ' ' Based on the above, the NRC inspectors concluded that the activation .of the emergency organization in the TSC appeared to be adequate. ~ F , 1.3.1.4 . Communication Interfaces The NRC inspectors reviewed TSC communication. interfaces against the i requirements of 10 CFR 50,' Appendix.E, paragraph IV.E.9. 10 CFR ' ;50.47(b)(5), NUREG-0737, Supplement 1, item 8.2.1.g, and selected - ' sections of section H of.the Plan. # ' 'The NRC inspectors determined that the TSC had a private automatic branch exchange (PABX) system located in Omaha that provided telephone service from the TSC.to the OSC, EOF, and control room. Additionally, the inspectors noted that the plant had a PABX system I, and'each emergency facility has at least one of the plant PABX telephones. Both systems had back up battery supply and emergency diosal power. In addition, the NRC inspectors noted that each communication system could cross communicate or operate independently, and that there existed a Conference Operations Network (C0F) (green phone) dedicated system for communicating with the control room, TSC,: EOF, and offsite State and county agencies. * The NRC. inspectors determined that there were written procedures (e.g. EPIP-OSC-15-1, EPIP-OSC-14-4, and EPIP-TSC-2) describing the ' use_of telephones and communication / notification checklists. 2 Based on the above, the NRC inspectors concluded that communication , interfaces appeared'to be adequate. 1.3.1.5 ~ Offsite Interfaces + , The NRC inspectors reviewed offsite interfaces against the requirements in 10 CFR 50.47(b)(5), 10 CFR 50, Appendix E.IV.E.9 and NUREG-0737, Supplement 1. In addition, the inspectors reviewed
v
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37 section F of the Plan and procedure TSC-2 which describes communications systems available for offsite interfaces from the TSC and E0F. The NRC inspectors determined that the TSC communicators could contact the NRC via the Emergency Network System (ENS) or commercial telephone, and that communications and interfaces with offsite agencies could be established by the dedicated C0P network terminettes (computer terminals) to Logan, Iowa; Des Moines, Iowa; and Lincoln, Nebraska. Additionally, material could be sent by the facsimile transmittal system to the E0F, Lincoln, Nebraska State E0C, Des Moines State E0C, and Logan, Iowa (Harrison County EOC). Further, the NRC inspectors noted that the Conference Health Physics Network was a dedicated system for the TSC; EOF; Logan, Iowa; Des Moines, Iowa; and Lincoln (State E0C); and to the Nebraska EOF room or communications mobile van. The NRC inspectors noted that the licensee used these communication systems to interface with offsite agencies in making protective action recommendations and other transmittals of information. Based on the above, the NRC inspectors concluded that offsite communication interfaces appeared to be adequate. 1.3.1.6 Transfer of Responsibilities The NRC inspectors reviewed the transfer of responsibilities in their emergency response organization against the requirements of NUREG-0737, Supplement 1, item 8.2.1.a and reviewed procedures RR-18, OSC-14, and EOF 13. The NRC inspectors determined that the shift supervisor was automatically designated as the Site Director, and that upon arriving onsite, the plant manager or designated alternate was briefed by the shift supervisor according to procedure EOF-13 in order to transfer responsibilities for emergency direction. In addition, the NRC inspectors determined that key emergency personnel and emergency workers were informed of responsibility transfers by announcements over the plant address system. The site director notified offsite agencies of protective action recommendations via telephone and verbal announcements over the TSC address system, and the coordination of offsite monitoring was accomplished via status boards, radio, and telephone. Based on the above, the NRC inspectors concluded that the transfer of responsibilities appeared to be adequate.
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51.3.'2.1 I; Technical Support yy' + ~ ,, .The'NRC inspectors reviewed.TSC technical support against the
- ,
; requirements of item 8.2.1.a'of Supplement 1 to NUREG-0737.- =, _ 1 E' ;The NRC inspectors held discussions'with the plant manager, operating y , s ' personnel, HP supervisor, Land the technical assessment staff,
4 . inspected the TSC, observed
' ~ ,' , _ , , facility staffing requiremen.the TSCprocedures ts and walkthrough, and ' inspected for status board the y logging, data transfer, communications, and core damage assessment.. ' ' c - Based on'the satisfactory results of.the above, the NRC inspectors - - -concluded that.TSC technical support of.the control room appeared to . be adequate. 1. 3. 2. 2 - Walkthrouchs The NitC inspectors observed the TSC walkthroughs to determine whether ' TSC personnel appeared capable of performing their assigned- t, ' functions, and whether the facility and equipment were adequate. -The NRC inspectors met with the utility representatives and developed ' - 'a listing of functional requirements-by facility. The licensee ' developed various scenarios which would demonstrate those , . ' capabilities during the walkthrough, provided controllers, and ran J the 2 hour walkthrough with the NRC. inspectors acting as observers. ' The NRC inspectors suggested that since the OSC staff was integrated .into the TSC spaces under the FCS organization, a combined TSC/OSC walkthrough should be' held. The combined TSC/OSC scenario ' demonstrated the following TSC/OSC functional capabilities: dose : projections, PARS, notification, classification, habitability,and ' technical support,'plus evacuation / accountability decision making and retrieval of meteorological data from non-0 PPD sources. The NRC inspectors determined that both the TSC and the EOF simulated the , , release and transmission of PARS without obtaining signed . authorization from the site director or recovery manager. In the case of the TSC, the PARS had not been reviewed by the site director. The NRC inspectors. identified the following deficiency in the TSC walkthrough: * * The licensee failed to obtain site director or recovery manager signature authorization prior to transmitting protective action recommendations as required by 10 CFR 50 Appendix E, paragraph IV.A.2.C and the guidance of NUREG-0654 II.B.4. (This is a repeat item from earlier exercises and most recently identified in NRC Inspection Report 50-285/86-11. The deficiency applies to both the TSC and EOF). (285/8620-19) i) . e
; - - . . - t- g - 39 2.0 Operational Support Cer.ter (OSC) 2.1 Physical Facilities ' ~ 2.1.1 Design 2.1.1.1 Location The NRC inspectors reviewed the location of the OSC against the requirements in 10 CFR 50.47(b)(8) and NUREG-0737, Supplement 1, item 8.3.1.b. t The NRC inspectors determined that the OSC functions were performed at two locations: the Shift Supervisor's office near the control room, which served as an OSC for the Operations Support Manager and shift personnel, and the Technical Support Center Building, from which the remainder of OSC functions were carried out. Additional 0SC support personnel could be dispatched from a nearby assembly warehouse located outside the protected area but located on the licensee's owner controlled area. Based on the above, the NRC inspectors concluded that the location of the OSC appeared to be adequate. 2.1.1.2 Alternate OSC Location The NRC inspectors noted that the licensee did not have a designated alternate OSC, and the primary OSC functions were carried out in facilities that had a protection factor of five, and charcoal and high efficiency particulate filtering systems. The inspectors determined that the OSC would be radiologically monitored during an emergency for personnel habitability. Support personnel assembled at the warehouse would be relocated to the North Omaha station cafeteria approximately 17 miles sou% of the Ft. Calhoun Station. Based on the above, the NRC inspectors concluded that these OSC provisions appeared to be adequate. 2.1.1.3 Size, Layout, and Environment The NRC inspectors reviewed the size, layout, and environment of the OSC against the guidance in NUREG-0696. The NRC inspectors examined the conceptual design, the Emergency Plan, and EPIPs, observed a walkthrough and inspected the facility. The NRC inspectors determined that the OSC consisted of a series of rooms that included the Shif t Supervisor's office (approximately 6 operators, 200 sq. feet), Room 114 (9 persons, 250 sq. feet),
i l
- _ - - - - - - , - _ , . . _ . , _ - . - _ _. _ . - _ . _ - - . - - ,_ -
p . 40 s Room 112 (11 persons, 200 square feet), Room 101 (2 persons, 100 square feet), Room 116 (1 person 150 square feet), and an assembly area of 300 square feet for additional support personnel. The inspectors noted that the size and layout of the OSC areas were sufficient for their function as an assembly area for certain inplant personnel. The inspectors further noted that since emergency tasks, including pre-dispatch briefings, were conducted elsewhere, the normal office environmental conditions were sufficient. In addition, during emergency conditions, the Shift Superv'.or's office will be located within the CR envelope, while the remainder of the OSC is within the TSC protective envelope. Based on the above, the NRC inspectors concluded that the size, layout, and environment of the OSC appeared to be adequate. 2.1.1.4 Display Interface The NRC inspectors reviewed the OSC display interface against the requirements of 10 CFR 50.47(b)(11). The NRC inspectors examined the Emergency Plan and EPIPs and inspected the facility. The NRC inspectors determined that there were no status boards within the OSC but that plant and radiological conditions were made available to OSC teams when called to the TSC for briefings prior to dispatch for performance of diverse emergency tasks. Based on the above, the NRC inspectors concluded that the OSC display interface appeared to be adequate. 2.1.2 Radiological Equipment and Supplies 2.1.2.1 Radiation Monitoring The NRC inspectors reviewed the OSC radiation monitoring instrumentation inventory, examined fixed and portable instruments and reviewed ten selected instrument calibration records to determine if radiological monitoring capabilities met the requirements of 10 CFR 50.47(b)(8), 10 CFR 50.47(b)(11), and Section IV.E.1 of Appendix E to Part 50. The NRC inspectors noted that the OSC was divided into two general areas. One area was the shift supervisor's office in the control room and the other area was in the TSC. The monitoring capabilities of the TSC area were reviewed in Section 1.1.2.1 of this report. This section will be limited to the control room part of the OSC. The NRC inspectors determined that an area radiation monitor was located just outside of the shift supervisor's office and an iodine
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L ' ~ air monitor was located in the control room area. These two monitors ;would continuously measure radiation and airborne radioa:tivity levels in the OSC and control room during the course of an accident. ,In addition, the inspectors noted that one high-range portable dose
' * rate monitor, one high volume air sampler and a-sample counter would
be.provided for OSC/CR~ room habitability surveys and contamination control. activities during an accident. In reviewing'the monitoring responsibilities of the OSC, the NRC inspectors concluded that
, .- radiological equipment appeared to be adequate to support OSC/ control
' p roce radiation protection activities. :The NRC inspectors reviewed the surveillance test (ST-RM-3) records y' for' operability checks and inventory of energency plan supplies and equipment, which was conducted on a monthly basis, and ten selected - calibration procedures and records for radiological instrumentation. In addition, observations of calibration stickers and performance of operability tests for emergency equipment by the NRC inspector verified that equipment was calibrated and maintained as stated in - the Plan and procedures. Based on the above, the NRC inspectors concluded that radiation monitoring in the OSC appeared to be adequate. '2.1.2.2 Personnel Dosimeters The NRC inspectors reviewed the OSC/CR radiation dosimeter inventory and examined selected self-reading dosimeter calibration stickers to determine if the radiological monitoring capabilities met the , requirements of 10 CFR 50.47(b)(8), 10 CFR 50.47(b)(11), ' Section IV.E.1 of Appendix E to Part 50, and item 8.3.1 f. of Supplement 1 to NUREG-0737. The NRC inspectors noted that 20 TLDs were provided to supplement the licensee's preassigned dosimeters. The NRC inspector also noted that 20 (0-500 mrem), 20 (0-50 rem), and 5 (0-100 rem) self-reading dosimeters were provided for the-0SC/CR staff. A log of self-reading . dosimeter issuance and personnel doses was maintained during accident conditions. In reviewing the self-reading dosimeter inventory, the NRC inspector noted that the number of dosimeters available in the TSC were adequate for the emergency response staff. The NRC inspectors suggested intermediate range self-reading dosimeters (e.g., 0-5 rem) which could accumulate up to the design doses with good accuracy of readout without needing recharging. (See Section 1.1.2.2.) The NRC inspector sampled the calibration stickers of self-reading dosimeters, reviewed the surveillance checklist, and verified that dosimeters were calibrated and maintained in accordance with the emergency plan and EPIPs.
sm y , ,. --- -- *'* g; %, ~ _ i
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b)E ' ' , 42 s- , ' " , ! Based on the above, the NRC. inspectors concluded that personnel e dosimetry in'the'CR and OSC appeared to be adequate. " , , 2.1.2.3 Protective Supplies .r .. . . . . ~ The NRC inspectors reviewed the OSC/CR dedicated radiological a ^ protective supplies inventory and examined the contents of the OSC/CR storage cabinet to determine if the radiological control capabilities
J. . set the requirements of 10 CFR 50.47(b)(8), 10 CFR 50.47(b)(11), and V' 'Section IV.E.1 of Appendix E to Part 50.
.w . , < The NRC inspectors noted that the licensee had provided respiratory
A. protective equipment, protective clothing, potassium iodide tablets,
,V and other protective and support supplies for the OSC and control roon.. I 1- The NRC inspectors also reviewed the surveillance test (ST-RM-3) for * . inventory of emergency plan supplies and equipment, which was conducted monthly, and verified that the protective supplies were maintained as stated in the Plan-and procedures. ' Based on the above, the NRC inspectors concluded that protective 1 supplies appeared to be adequate for supporting radiological protection activities in the OSC and CR. t' ' ' LNon-Radiological Equipment and Supplies _2.1.3 M 2.1.3.1 Communications . ' The NRC inspectors reviewed licensee's communication systems against the regulatory requirements in 10 CFR 50, Appendix E, Section IV.E.9 and NUREG-0737, Supplement 1, item 8.3.1.c. In addition, the inspectors reviewed section H-3.2, " Communication Failures," of the emergency plan, t The NRC inspectors determined that the OSC had an inplant intercommunication system for exchanging information among the TSC, OSC, and CR; local extension' telephones; and direct dial telephones , for offsite calls. The NRC inspectors noted that using these communication systems, OSC personne1'could communicate with inplant emergency teams. The NRC inspectors noted that the licensee had used all three communication systems during exercises and the systems
.
appeared to have performed satisfactorily. Based on the above, the NRC inspectors concluded that communication systems appeared to be adequate. 2.1.3.2 Support Supplies The NRC inspectors reviewed tho OSC non-radiological support supplies against the requirements in 10 CFR SU:47(b)(11) and 10 CFR 50, Appendix E. Section IV.E.1. - , - - - - -
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e ,; ;;; :- , 43. y' :y; -:: . , .,, , y ': Lw>~ _ _ . . . . u . ' ,c ' _ , , The' NRC inspectors examined the Emergency Plan and EPIPs and
n,'m: 11nspected the: facility.
,y - W y / < " The'NRC inspectors determined that plant reference materials and ' 9~ , other job performance aids were available within the OSC. Damage < ' ' -w : control equipment was maintained in the Tool Room, with spare' parts.
n _ and equipment in the store room,
em ,. , , ._ , s. _ M -Based on the above, the NRC inspectors concluded that OSC
y non-radiological support supplies appeared to be adequate.
4 - I2.'2.FunctionalCapabilitiesandWalkthroughs c .2.2.1- Operations J . "2.2.1.1 lStaffino- , ; . , < ' 1- * The NRC inspectors reviewed licensee staffing against the requirements in 10 CFR 50, Appendix E, Sections IV.E.4 and 5, and ' iNUREG-0737, Supplement 1, item 8.3.1.a. In addition, the inspectors . ,; ' reviewed sections H and M of the emergency plan. ' 'The NRC inspectors determine'd that section H-3.4 contained a list of , personnel who,would report to the OSC in the event of an emergency.
_
- The NRC> inspectors noted that records of past drill exercise ' critiques . reflected that additional OSC personnel were needed, and ~ ' * .that corrective action'was taken. The NRC inspectors found that , operations, health physics, chemistry, and maintenance personnel were ' - available in the OSC, that they were responsible for performing in plant monitoring, first-aid, fire brigade, damage control, - ' ' post-accident analysis, and search and rescue, and that each OSC - functional group had a supervisor who reported to the Site Director. ' . ' ' ; Based on the above, the NRC inspectors concluded that OSC staffing . appeared to be' adequate. . 2.2.1.2 Activation : The NRC inspectors reviewed the licensee's activation scheme against u' , regulatory requirements in 10 CFR 50.47(b)(2) and NUREG-0737, Supplement 1, item 8.3.1.a. ' ' The NRC inspectors determined that licenste's procedure OSC-2
"
provided written instructions for the activation of the OSC at an ' emergency class of a Notification of Unusual Event. The inspectors ' noted that during the recent June 25, 1986, annual emergency exercise, the OSC was activated and became functional in -z ~ approximately 1 hour and 45 minutes, and that since the licensee does not have an alternate OSC, transfer of responsibilities to alternate OSC staff would not apply. ' -- - _ . _
. , / y, O - ,: 4 . 44
y,
i' The following deficiency was identified in the activation scheme:
E~ *
The licensee has not demonstrated that the OSC can be made
% _ operational within about I hour. (285/8620-20) h '2.2.1.3' Onsite Interface
The NRC inspectors reviewed the onsite interface with regulatory
i~ requirements in 10 CFR Appendix E.IV.E.9 and NUREG-0737,
Supplement 1, items 8.3.1.a and b, and reviewed section M of the emergency plan. The NRC inspector determined that OSC personnel were , located in an area adjacent to the control room and reported to supervisors co-located in the TSC, who in turn reported directly to the emergency director. Based on the above, the NRC inspectors concluded that the onsite , interface appeared to be adequate. 2.2.2 OSC Functions ' 2.2.2.1 Coordination, Assianment, Proficiency, and Walkthroughs The NRC' inspectors reviewed coordination, assignment, proficiency, and walkthroughs against the requirements of Supplement 1 of NUREG-0737, items 8.3.1.b and c. ,, The NRC inspectors reviewed a list of deficiencies from the last exercise report and noted that the OSC provided adequate support during the simulated emergency. The inspectors attended a combined walkthrough of the TSC and OSC. Results were reported in Section 1.3.2.2. Based on the above, the NRC inspectors concluded that provisions in this area were adequate. 3.0 Emergency Operations Facility (EOF) 3.1 Physical Facilities 3.1.1 Desian 3.1.1.1 Size The NRC inspectors reviewed the EOF size against the requirements of 10 CFR 50.47(b)(8) and Supplement 1 to NUREG 0737, items 8.4.1.c and k. The inspectors examined the conceptual design, the emergency plan, and EPIPs, and inspected the facility. The NRC inspectors determined that the EOF was approximately 6000 sq. feet in size, and that it included space for the licensee emergency -
, . _ _ _ _-. - - - - . - . - - - - - - - --- - - -- y :, , f, , ' ' x % w.- . -
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- ' ! ; * - g' f , . ' . response.. staff;. federal.: state,.and local personnel; a' press briefing - ; "' . . ;" room; records; storage; sanitary facilities; and building support : ' , 3(HVAC and electrical). .The1 inspectors noted that procedure EOF-1-3 J , e listed 45 persons assigned to the EOF, which yielded over,100 sq. feet >
y'gg 3' -* , per person at full: assigned occupancy. Workstations provided ample
: floor space'and horizontal workspace to support the tasks performed by ' ^ ~'* ' each individual in the E0F. .-In addition, .the inspectors noted that: - * each piece of operational _ equipment (CRTs, PCs, printers) was readily - , . accessible for corrective or protective maintenance or replacement. . -[ :3 $ " l> _. : Based on' the above, the NRC inspectors concluded that the size of the m3g .. EOF: site. appeared to be adequate.
s" ,
1, - ' . -3.1;1.'2 . Layout- - f' ' i The'NRClinspectors reviewed the EOF layout against 10 CFR.- 50.47(b)(8) and Supplement 1_to NUREG-0737, item 3.4.1.k.
'/
' .c 'The.NRC inspectors examined the conceptual design of the EOF, the ~ . emergency plan, and EPIPs, inspected the facility, and determined that _ ' < the layout of the EOF provided a single room to be the focal location * ;for emergency' response activities. '
1- '
e ' The-NRC: inspectors determined that the EOF had' separate areas for * / emergency. assessment and radiological' assessment that.were separated
f7
from each other by the control center in such a manner that it provided physical and acoustic separation without significantly, , " . ~ impeding visua1 Lor physical access. Other teams were located in. separate rooms across'a corridor that~1 oops around the core formed by * Rooms 7 and 13. The location of doors in the EOF. focal location , (Room 13) allowed personnel to move 'about freely without disrupting activities in unrelated work areas.- Distances among individuals were- , ' , close enough to allow for adequate interaction. Based on the above, the NRC inspectors concluded that the EOF layout ., appeared to be adequate. ~ 3.1.1.3 Location The NRC inspectors reviewed the EOF location against 10 CFR ' , ,50.47(b)(8) and Supplement 1 to NUREG-0737, item 8.4.1.b and Table 1. ' The inspectors examined the conceptual design, the emergency plan, and EPIPs, and inspected the facility. The NRC inspectors determined that the EOF was located 17 miles from the Ft. Calhoun Station, at the North Omaha Station, and noted that
l
' provisions had been made for housing representatives from offsite agencies (federal, State, and local) during emergencies. Based on the above, the NRC inspectors concluded that the location of the EOF appeared to be adequate. !
w..
' s ps "
.
46 4
y
" 3.1.1.4 Structure ;. The NRC inspectors reviewed the EOF structure against Supplement 1 to NUREG-0737, items 8.4.1.b and d, and Table 1, examined the conceptual design, the emergency plan, and EPIPs, and inspected the facility.
lL The NRC inspectors determined that the EOF was designed to meet the
r . requirements of the Omaha Municipal Building Code, which provided for
[ the adoption of the National Building Code-1976 Edition. This means
the building was designed to meet winds and floods with 100 year
{[ , ,
frequency of recurrence.
4 'f Based on the above, the NRC inspectors concluded that the structure
of the EOF appeared to be adequate. 3.1.1.5 Habitability / Environment The NRC inspectors reviewed E0F habitability against requirements in 10 CFR 50 and item 8.4.1.b of Supplement 1 to NUREG 0737. ' Since the EOF is 17 miles from-the plant site, radiation protection habitability features were not required or provided. Based on the above, the NRC inspectors concluded that habitability provisions for the E0F appeared to be adequate. i 3.1.1.6 Display Interface The NRC inspectors reviewed the EOF display interface against the guidance in NUREG-0696. .The inspectors reviewed the emergency plan and EPIPs, examined the system description, operating instructions, and training guide for the ERFCS/SPDS, toured the facility, and operated the CRTs. The NRC inspectors determined that Room 13'in the E0F contained preformatted status boards for plant and radiological data, four large EPZ maps, and an emergency classification matrix. The inspectors noted three unformatted display areas (i.e. a marker ' , board, chalk board, and flip chart), a microfilm reader / printer, EPZ maps, protective action sector charts, and emergency response status charts. The inspectors also noted that display boards in the EOF were readily visible, adequate in number, understandable to those personnel who needed to monitor the information, and could be updated in a timely manner. The NRC inspectors determined that the E0F contained two terminals for the ERFCS/SPDS computer system, which provided a wide range of t
g& , We , '% gy ~ ; - - ' ' *. ' . .. ' .. n.' - , - , - * q.n o , ' 1 pn.9? - , . ', . ' ' . . " - ' s j .jin - _ -, ' , , ai4 2.g' r z s - 's G fi' ', [ ~ . , 7 , . . , . 0 > ^ Jr , f , ._ ~ V f.W 4, *- h hi: "! > ' - ' "l " ' - - . w%ww3 < . . + ~ Q' : g& . -; : ? ' . __[ , . . _ . _ _ . _ plant,sradiological and meteorological variables accessible to EOF .; f y. %b , i - v ; personnel. :Referfto;Section 1.1.1.6 of this; report for further MW -- ^ discussion. ( M4 1 . , ._ ?" %.' . , 7 Based--on'the'above,8the NRC_ inspectors concluded that the EOF display ? interface ~ appeared to be' acceptable. i A ,1.1 f- .&y , - - - m J Radiologicil Eau'ipment and Supplies % [ '13.1;2 [ hh[ , ,w= N 'I3.1[2'.1 . Ra'iationMonItorina d , , 4 > * ~ - , h ;The NRC: inspectors determined that a high-range dose rate survey JE7 . instrument, a portable air sampler, and a sample counting kit were - t #j % ' > .+ - ; maintained in the EOFfforfradiation monitoring contingencies. ' . e L :The'NRC inspector also reviewed the surveillance-test.(ST-RM-3) for ;9(m " ' s , inventory lof emergency plan equipment, which was being conducted , Leonthly,:and verified that the protective' equipment was maintained as p, ? _ ,. stated in the Plan and~ procedures, a + n;.= . , ' 7M : Based on.the above,~the NRC inspectors concluded that radiation , monitoring in the EOF appeared to be adequate. p . - 3.1.'2.2 Personnel Dosimeters ~ ' The NRC inspectors ~ determined that 12 TLDs and 12'(0-500 mrem) and 12 L(0-50 rea),self-reading dosimeters were maintained in the EOF for , radiation monitoring contingencies. t ,a . . + i ' - tThe NRC1 inspectors also. reviewed the surveillance test (ST-RM-3) for inventory.of emergency' plan equipment, which was conducted monthly, and verified that the personnel monitoring equipment was maintained as stated in tha' Plan and procedures. . - . - Based on.the'above, the NRC inspectors concluded that personnel ! . dosimetry in the' EOF appeared to be adequate. 3.1.2.3 ProtectiveSboplies .
6 '
< The NRC inspectors determined that respiratory protection equipment, protective clothing, potassium iodide tablets, and other protective and support supplies were being maintained in thec .0F for radiation
} ,
; control contingencies.. The NRC inspectors also reviewed the surveillance test (ST-RM-3) for
, '
, inventory of emergency plan supplies, which was conducted monthly, and , . verified that the protective supplies were maintained as stated in the Plan and procedures. '
p h. ^
Based on the above, the NRC inspectors concluded that protective
i
supplies in the EOF appeared to~be adequate. n. .
, I;
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s m ,; c , y 7,1 . , ,x , * . A * 1 . a , 4g , y * ,g , .<- . 3.1. 3' Non-Radiological Equipment and Supplies yk l3;1.3.1: Communications- ~ ' ~ ' . . :The_NRC: inspectors reviewed licensee ~ EOF communication links with . emergency _ response facilities and offsite agencies against the ( ' 1 regulatory requirements in 10 CFR 50.47(b)(5), 10 CFR 50 Appendix E, ' Section IV.E.9, and NUREG-0737, Supplement 1,' item 8.4.1.f. In -addition, the inspectors reviewed EOF communication section H-2.2.1 4 x of the emergency plan. . ' l ' i The NRC inspectors determined that the EOF communication system a ' consisted of_a' dial telephone' system pfovided by#a telephone system located in Omaha, Nebraska. 'This.sy' stem provided an intracompany . 't telephone system with access to the public telephone network. Three * a - -lines from the Ft. Calhoun dial-telephone system'(PABX) in the EOF ,' -allowed uninterrupted private communications with all Ft. Calhoun _ _; e - , : station service areas. The.NRC~ inspectors noted that dedicated 3 : telephone: lines were.available from the EOF to the Nebraska' State EOF, ~ .p ' -Washington County. Emergency Operations' Center (E0C), Iowa State EOC, b c LPottawattamie County EOC, and Harrison County EOC, and that - radiocommunication means were in-place using Ft. Calhoun's ultra high ifrequency-radio repeaters. The NRC inspectors determined that the -C ^ -radio. system could be used in conjunction-with portable radios to - ; communicate with-offsite radiological. monitoring teams. Additional ~ . ' ? radio equipment was available in-the E0F for the Nebraska State Civil Defense emergency response team. * ~ < - iThe NRC inspectors noted that an~ emergency notification system (ENS) i ' C dedicated'line was available-for.NRC and licensee emergency response b. communicators. Additionally, a commercial-line was available for the ; - NRC and licensee to exchange radiologi al data and information. Two .'"' < - facsimile' machines were installed in the EOF to transmit information to. compatible offsite agencies receiving equipment; e.g., the NRC, , - ' , ~the states of. Iowa and Nebraska, and Logan, Iowa, in Harrison County. - - The inspectors found that the Media Release Center located at the s , - ' Omaha / Douglas County Civic Center could communicate with the EOF via % 'a-dedicated telephone line located in the. EOF media response area, 'and that a dose assessment computer terminal available in the EOF ( ,7 l , would allow transmitting meteorological, radiological, and dose - -assessment data to the TSC, state EOCs, and Iowa's forward operation ] ,1ocation. ' < The NRC! inspectors noted that the licensee had used drill and . _" ' exercise critiques to identify communication deficiencies and had taken corrective action. The results of previous drill and exercise Y' ' . deficiencies had resulted in telephones being added, system changes, or telephones being relocated. ' ,
,
. r* -3 g w ~ vy, , --ee.e-ere w . - - p ,,,m.- ---g-%-1.- .>w.. 9 ----. -w- r- ,,. - - .~ -- - e. m,-- y- -9,4---my-, - .. -- -
Wa, ,w g > ~ > o - ..y p. , Q L , \, ., - q :- ., , F ,q. , ~ ' M gJ ' . , ' % ;w - v $$% Js ,49. $ di.% yk d,( . - ' " ' ' % .. . _ _ . c2 ~ ; d. G fN:..v% ,~ . .. . . . , N. ,, Finally,the inspectors determined that communications equipment in d%4 . -g: the EOF had backup battery, power or an-alternate power source in the g- . g % - . event of a loss;of power to,the-EOF. ., . - ' _ .
f J, y Based on the above, the NRC inspectors concluded that communications
W , Mac ;: , systems 11n'the EOF. appeared to be adequate. ,, =. yll pp.M . ' ' f)( % ' z3.L 3.2 [ Recons /Orawings - ' - - , ,, . . . . ~ . 1R efer to'Section 1.1.3.2 of this report.' , . . c f$i ~ ,13.1I3.3 Support Supplies #
krL j?t
, .N vic The NRCcinspectors1 reviewed support supplies maintained in the E0F - :. . ., - 7'F "' Lagainst'the inventory listing provided within the Surveillance test ' LST-RM-3,7" Emergency Plan. Radiation Instruments and Equipment." g ' . l' ' ,4 -The(inspectorsdet'erminedthatthe.inventoryincludedplantand. area w , > # , . maps, calculators, pens, pencils, grease pencils, paper, flashlights, _ ' ' q 2, ~ flashlight batteries,-and masking tupe. The Surveillance Test was - N'. O ~ B required to be performed monthly. The NRC inspectors reviewed the previous 6 months of. tests and determined that the system was adequate Y,- - sto maintain the" EOF inventory. Fg n. . . . , h ,The NRC. inspectors reviewed'other EOF supplies and noted that %bj$ fy.3 z> * . ' .isopleths were:readily available for use, as'were means for data - ?m 4, . trending, computer.-paper,'and a library of reference material which o' " #_ , included the~RERP and EPIP-(5 sets), the USAR, state of Nebraska h sEmergency Plan, Estate of Iowa Emergency Plan, Pottawattamie, Harrison, . g .and Washington County' Emergency Plans, FEMA Region VII Emergency W ! Response Plan,-a complete set of operating manuals, P&ID books, ?? J 3 ,. ; ' ,. ' Technical Specifications, and the INPO Emergency Resources Manual, ' . , , . 21984. . 3 . / y , 4 , N ~'" i -sasedontheIbove,ftheinspectorsconcludedthtEOFsupport
^; JP l supplies'hppyaredtobeadeque%.
.' ~' ' ,. " 3.27,,_Information Nadagement~ :% 'O:*7 M y. . . R - 13.2.1 Variables-Provided. .. '
h, .&;'n ~'
S.
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.' G 3. 2.1.11 Regulatory Guide'1.97, Revision 2, Variables
d .' - JThe NRC inspectors conducted interviews, held discussions with
b ,.. licensee personnel, conducted E0F walkthroughs, including several
i '
it W computer terminal demonstrations and pertinent documentation- specifically, OPPD letter dated April 2,1985, " Fort Calhoun Station '
h.;7 , ' M// ' Compliance with Regulatory Guide 1.97, Revision 2," and NRC letter
dated June-18, 1986,."Conformance [of OPPD] to Regulatory Guide 1.97,
l; 4 3 .k L 1
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' . . , - - i? ; 3*
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. '
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* 1 50 Revision 2." Based on the above, the NRC inspectors concluded that unresolved items ~ identified in Section 1.2.1.1 would also apply to -this section. 3.2.1.2 Other Veriables , Refer to Section 1.2.1.2 of this report. 3<2.1.3' Relationship to Functional Needs Refer to Section 1.2.1.3 of this report. 3.2.2 Data Acquisition :3.2.2.1 Data # Collection Methods . Refer to Section 1.2.2.1 of this report. 3.2.2.2 Time Resolution The NRC inspectors evaluated time resolution of sensor data available 'to the ERFCS, against the requirements in item 8.4.1.g of Supplement 1 to NUREG-0737. The NRC inspectors reviewed the OPPD ERFCS System Software Overview, conducted interviews and held discussions with a system program analyst, and received a hands-on demonstration of ' system capabilities in the EOF. , Refer to Section 1.2.2.2 of this report for further details. 3.2.2.3 Isolation Refer to Section 1.2.2.3 of this report. 4 E 3. 2. 3 - Data Communications 3.2.3.1 Capacity - Refer to Section 1.2.3.1 of this report. 3.2.3.2 Error Detection s b +s Refer to Section~1.2.3.2 of this report. 3.2.3.3 Transmission Between ERFs . Refer to Section 1.2.3.3 of this report. 3.2.4 Data Analysis -
p ---
_ ,-._ , __ __
44;,; . . ,
a .
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~ < 51 *
l;
' ~ , 3.2.4.1, Reactor Technical Support Refer to Section 1.2._4.1 of this report.
L -3.2.'4.2 . Dose Assessment
- Refer to Section 1.2.4.2 of this report. ' - 3.2.4.3 Central Processor Capability
-
Refer to Section 1.2.4.3 of this report.
l _ .
- 3.2.5 . Data Storage 3.2.5.1 = Storage Capabilities _ ' ' Refer to-Section 1.2.5.1 of this report. 3. 2. 6 - System Reliability and Validity , ~3.2.6.1 Validation and Verification - ' Refer to Section 1.2.6.1 of this report. 3.2.6.2 Computer Based Systems Refer to Section 1.2.6.2 of this report. , 3.2.6.3 . Manual Systems ; Refer to Section 1.2.6.3 of this report. - : 3.3 Functional Capabilities and Walkthroughs ~3.3.1 Operations 3.3.1.1 Organization
L-
-The NRC inspectors compared licensee organization commitments with regulatory requirements in 10 CFR 50.47(b)(2) and NUREG-0737, Supplement 1, item 8.4.1.i. The inspectors reviewed Section M of the emergency plan and held discussions with selected licensee staff members. The NRC inspectors determined that the E0F organization was the same , as indicated in the Plan and procedures. During the recent exercise conducted June 25, 1986, the EOF was staffed in approximately 1 hour .with the organization indicated in the Plan. Based on the above, the NRC inspectors concluded that the emergency response organization of the E0F appeared to be adequate.
,
' '~ , , i;3f. y [. ; , ; s. - p , . . - *' , M. $g 7#; :,2 . . 52 -i. , ^ ll - l p * _' m- . .. , r ' rQ2 i J3.3.1.2 ' Staffing ~ ' " ; , , '* " The.NRCinspectorscomparedlicensedstaffingcommitmentswith , ,?N -regulatory requirements in'10 CFR 50.47(b)(2) and NUREG-0737,- - . ) Supplement 1, and reviewed section H-2.4, of the emergency plan. , N i,,! 1 <. ' ,s ' # _ . S .The NRC inspectors. determined 1that-the' licensee had 21 positions ; ~-* ' including: Recovery Manager, Recovery _ Operations-Coordinator; JW^F ' ~ Administrative Logistics Manager,' Emergency Coordinator,' and Dose + , - T- Assessment' Coordinator.. ' , - n - . . . . ~ * ? Based on the'above, the.NRC in'spectors concluded that the staffing of hf y *- . , , (the EOF emergency. response organization appeared to be adequate. 7 . 3.3.1.3 ' Activation- h? f" . The NRC inspectors compared licensee activation commitments made in a ~1 . - letter dated March 1, 1982, with regulatory requirements in 10 CFR . < < 50.47(b)(2) and NUREG-0737,' Supplement 1, item 8.4.1.i. :The NRC . inspectors determined from a review of the licensee EOF activation . ' - , 4 ,' ~ .. irecords pertaining to the annual exercise conducted June 25, 1986,
%% '
t that the' EOF.was activated and functional within 1 hour and 9 minutes .after the declaration of the alert emergency class. 4' y' j . . Based on the above, ths NRC inspectors concluded that the activation @f -time of the emergency response organization appeared to be adequate. '3'.3.1.4' ~ Communication Interfaces. - t.y' . ' Refer.to Section 1.3.1.4 of this report. , ,. *- - 3.'3.1.5: 0'ffsite Interfaces . , , _ Refer to Section 1.3.1.5 of this report. 13.3.1.6 ' Transfer of Responsibilities . .The NRC inspectors reviewed the transfer of responsibilities in their , emergency response organization'against the requirements of Supplement 1 to NUREG-0737, items 8.4.1'.a and b, and Table 1. . , The inspectors reviewed the emergency plan and EPIPs and determined ? ~' 'that the transfers of responsibilities from the Shift Supervisor (CR) -to the Site Director (TSC) and from the Site Director to the Recovery ;
'
, Manager (E0F) were adequately addressed in procedures E0F-13 and E0F-14. ' . ' The NRC inspectors noted that the transfer of responsibilities was l ~ ' accomplished when the Site Director had reported to the TSC following
'
declaration of an Alert (a Site Area Emergency). Procedure E0F-2 +.
d
.([ l , - l , , . - - ..
_ _ .__ . -_ __ . .. _ - 5 . %_ h .: , ' 1y ' , . + J~ ^ ' * 53 ' f y . - - c g , y. , - T , ,_ ; c" . . ,W^ . instructed the user to' notify inplant personnel of.the transfer from m >the: Shift Supervisor to-the Site Director by means~of.;an' announcement- 'fif"r , , a ' Lover the'GAITRONICS~ system, but'did not explicitly indicate how , 't c. 7 u . " > ~offsite agencies.would be notified of' this. critical . transfer of: . responsibility." RR-10 directed the- Recovery Manager to make:a formal ' ' L ' , d.% q' announcement, to issue a memorandum, Land to brief the EOF staff, . ^ ' ' W, including offsite representatives. - , ' Based on ths above .the NRC inspectors concluded that transfer of' g _ ; responsibilities ~ appeared to be adequate. ' ' 3. 3. 2 = TSC-Support a .3.3.2.1' Technical-Support i , _ i ~The NRC.inspectorsLreviewed EOF. technical support against the ' jrequirements of; Supplement 1 to NUREG-0737, item 8.4.1.a. * ~The NRC inspectors observed EOF walkthroughs and noted that ~ '~ coordination with vendors and outside consultants was demonstrated , adequately'by contacts made with Combustion Engineering, INPO, and - - several. logistic sources. . , c . -Based on.the above, the.NRC inspectors concluded that EOF ( -capabilities for providing technical support to the TSC under , 0 - , accident conditions appeared to be adequate. . ' 3. 3. 2. 2. : Logistic Support . LThe NRC inspectors observd 'the performance of support functions by ', _ -the-Administrative Logistics Manager during an EOF-walkthrough. The . inspectors determined that the' EOF was capable of providing extensive ~ c < _ logistic support to the TSC. The NRC inspectors noted that the - [. 'AdministrativeLogisticsLManagercontrolledandcoordinatedthese "' - efforts >through an assigned logistics. support group consisting of: . communications support, administrative support, a finance ~
1 , f_ ~ coordinator,:an accommodations coordinator, commissary support, a
human resources- coordinator, a materials management coordinator, a transportation coordinator, an accounting coordinator, a security coordinator, and a computer specialist. In addition, the inspectors ~ '__ noted that. adequate reference material was available, including ~ emergency phone numbers and the INP0 Emergency Resources Manual. ' ' Based on the above, the NRC inspectors concluded that the TSC
.
- logistic support provided by EOF appeared to be adequate.
'
~3.3.2.3 Implementation of Mitigating Actions
"
'The NRC inspectors observed walkthroughs.in both the TSC and the E0F, - ' and held discussions with licensee personnel associated with the . }w Y. .f:q; - K _ ma _ ~ ___ . - _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ . _ _ _
, - .. -.
7
b9: '
.(?o ~
i ' ,3
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' s * < + ,. 54 . s- > y ;de0elopmentandimplementationofmitigatingactions. These- . , f discussions ir.cluded considerations of offsite impacts associated with' mitigating actions, the' liaison and coordination maintained
P
* v between the two facilities during formulation of mitigating actions and alternative actions, and their. potential offsite impact. The NRC - ; inspectors noted that the: capability to develop dose assessments of : alternative mitigative actions was available and could be done on the i computer in"between real time assessment computations. . . . , " ; Additionally, 'the NRC inspectors' determined that an Advisory Support Group in the EOF proposed and< formulated alternative protective . actions for: consideration by the. Recovery Manager. _ Mitigative - ' ' actions were coordinated ~with the state of. Nebraska ~ Rad Health c ~ 1 Director and with the state of Nebraska. Civil' Defense Director, both located in'the EOF.' Coordination'with the state of Iowa was ' ,
(i -
accomplished with a representative of the state of Iowa Office of Disaster Services, who maintained an open line from the EOF to his * -superiors.in Des.Moines. ' ,, I The NRC. inspectors noted that-responsibility for implementing , _ mitigative actions under various conditions appeared.to be clearly understood by those . interviewed, that EOF notification procedures . _ were' observed to provide for adequate notification in-the event of .R containment venting or other planned releases, and that adequate ' decision aids were available to the EOF staff for planning ~a venting ' ~ of the containment. - Based on;the above, the NRC inspectors concluded that the area- inspected. appeared to be adequate. E " 3.3.3 EOF Functions -- . 3.3.3.1 Notification / Communication .The NRC inspectors reviewed EOF-Notification / Communication capabilities against the requirements of 10 CFR 50.47(b)(6), and : Supplement 1.to NUREG-0737, item 8.4.1.f. The inspectors reviewed the emergency plan and EPIPs and observed a walkthrough. 7 ThelNRC inspectors determined that the licensee's notification , procedures contained in OSC-2 were consistent with the emergency classification' scheme and contained a provision for verifying messages when necessary. Procedures for alerting, notifying, and , activating personnel-and organizations were complete, and the content ' , * , of emergency messages to offsite authorities was adequate. Based'on the above, the NRC inspectors concluded that the EOF notification and communications functions appeared to be adequate. t 4 ' 1 . - - - , - . - . , - - - - - - , ~ - - - , - - - - - , , - - - - - - -
sc y gg . m3 ~ [yM J . . kS f g hih; [ 55~ ~ y;Q wa x& , , , 3 . , . , M , , 13.3i3l2? tDose Assessment? , . [% ' ~ - _ LThe NRC inspectors reviewed the dose assessment capability in the EOF- "' ' :against;the requirements.of 10 CFR 50.47(b)(9), Appendix E, .. . ' . t, .! paragraph IV.E.2,' Supplement ~1 to NUREG-0737, item 8.4.1.a, and~ ' ' > Regulatory Guide 1.97.' The; inspectors interviewed licensee staff > J responsible for; development of the dose _ assessment program and its ,f _ ', , iassociated training, and' reviewed. relevant documentation. These- - included EOF-6, Section I of the emergency plan,..and the EAGLE i, ! Program User's Guide and. Technical Manual (December 1984). The Lw ~ < , ' . inspectors also checked equipmentJused to perform dose assessment and < <? .had the licensee perform sample calculations using the EAGLE computer ,n x, program. w. w : * The' inspectors. determined that if the EAGLE program was not > ' :available, a manual backup' system compatible with EAGLE was available in the control room, TSC, and EOF (see Section 1.2.4.2).- Critical . data inputs into.the manual, method included meteorological data,. ' ~ _ . stack flow rate, and process monitortreadings. , Procedure EOF-6 did ~ < t ~, Lnot provide guidance on-how to determine'if the primary source was ' not available for these inputs, or describe how to'obtain historical ' M; - ~_ , - , a meteorological. data for use during an emergency. , . - _. 'The .NRC inspectors determined that dose assessment was properly "' incorporated into the protective action decisionmaking process? The ' ' . ' dose assessment group in the control room, TSC, and EOF would report cprotective action recommendations to the' shift supervisor, ' ' f HP/ chemistry supervisor, orl emergency coordinator, respectively. : These recommendations would be based on the-EPA Protective Action h' ' -Guidelines. ' * Based on the above, the NRC inspectors concluded that dose assessment - $(4<A - in the EOF appeared to be adequate. ^ ~ ' ' . 3.3.3.3: -Protective Action Decisionmaking - ,. * , , ' p The NRC inspectors determined that the shift supervisor was assigned , Site Director duties upon declaration of an emergency, that Site Director responsibilities were transferred to the TSC upon activation , 2of the'TSC, and that responsibility for directing the emergency was .~ transferred to the Recovery Manager upon activation of the EOF. In addition,'the inspectors noted that it was mandatory for a shift . '
gi r
, , supervisor to remain onsite at all times. +' .
- Based on the above, the NRC inspectors concluded that protective ' 0 action decision-making was adequate.
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-3.3.3.4L Coordination of Radiological and Environmental Assessment The NRC inspectors held discussions with licensee representatives and . reviewed the emergency plan and procedures E0F-8 and E0F-18 to , *., - - , _ _ . _ ~ . _ _ - . , . . - _ . _ . - . _ . _ - - - . . . . - - - . _ _.- , - _ - ,-._ _ ,...-~ .-.--_ --- - , , -- _.
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-- u , 2 .. . 56' , w ^W e determine if the licensee's coordination of radiological and ' environmental assessments met the requirements of 10 CFR 50.47(b)(9), . * paragraph IV.E.2 of Appendix E to Part 50, and item 8.4.1 a. of Supplement 1 to NUREG-0737. ^ The.NRC inspectors determined that two OPPD field monitoring teams
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f based.on access roads and population centers. The field monitoring [ .
' results were displayed on a sector map board and compared with dose assessment results. The deployment of OPPD field monitoring teams was discussed with the state field team coordinator locatd in a room - adjacent to the OPPD operations area in the EOF. The NRC inspectors noted.that this coordination was ad hoc and suggested that the licensee use procedures and joint training to assure efficient and
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, well coordinated offsite monitoring actions. ' , The NRC inspectors determined that OPPD planned to use their en'vironmental program personnel to collect environmental samples ~
{" after,_the termination of the radioactivity release', using the routine
environmental sampling procedures, and forward the samples and , environmental TLDs to an offsite contract laboratory for analysis. ' The sample collection and analysis would be conducted under control .of the existing quality assurance procedures for the environmental program. Based on the above, the NRC inspector concluded that the licensee's . program in this area appeared to be adequate. 3.3.3.5 ' Walkthroughs The NRC inspectors observed E0F walkthroughs to determine whether the , > E0F appeared capable of performing its assigned functions. ' 'The NRC inspectors met with utility representatives and developed a listing of functional requirements for each ERF. The licensee l " developed scenarios which would demonstrate those capabilities, provided controllers, and ran two walkthroughs. lasting 2 hours cach with the NRC inspectors acting as observers. One walkthrough ; -involved the EOF; the other involved the TSC/0SC. I The scenario demonstrated the following functional capabilities from - each facility: dose projections, PARS, notification, classification, habitability, and technical support. In addition, the licensee den.onstrated evacuation / accountability decisionmaking and retrieval of meteorological data from non-0 PPD sources. The NRC inspectors determined that both the TSC and the E0F released and transmitted protective action recommendations without obtaining signed authorization from the site director or recovery manager. (See Section 1.3.2.2 of this report.) 1 h _ . . . .
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4 - #'i , - 57 W - m . , , . ~ , The*NRC) inspectors noted that_the EOF human resources coordinator- 4 -s .provided the administrative logistics manager with a manpower- ' - - schedule;for long tern EOF staffing which listed two positions as - ,", - > staffed by single individuals-on 24-hour ' rotation, the administrative supervisor!and the. licensing' administrator. The licensee explained -that the latter-position was "on call"; butithe alternate had ' ' - , ' terminated employment three weeks earlier and'no_ replacement was in- - - . . place. :The NRC. inspectors were informed that usually emergency v - positions in the OPPD/FCS emergency organization had.two individuals , assigned. The: inspectors concluded'that in some cases this scheme - would be insufficient to support long. term staffing of the ERFs. ' " The-following~ deficiency _was identified: ' ' ~ * An. insufficient depth of trained personnel was provided in their ~ - emergency' organization to ensure that ERF staffing will be - - adequately; staffed during long term emergencies. (285/8620-21) @ . . . . .. 14;0 Unresolved Items a- " ' An unresolved item-is a matter about which more information is required _ , - in' order. to a'scertain whether it is an acceptable item, a deviation, or a ' violation. -Unresolved-items are identified in paragraphs 1.2.1.1 and - ?1.2.2.3. -5.0 Exit-Interview ~ The exit' interview was conducted on July 11,.1986, with licensee < representatives. Mr. Donald E. Sells, Project Manager, NRC-HQs was in attendance.- , Mr.LNemen'M.-Terc,1the NRC Team Leader, summarized the team comments and observations in the subject areas of the Emergency Response Facility I : Appraisal. The NRC inspectors discussed the nature and specifics of each 1 of the 14 deficiencies, the 7 unresolved items, and 22 improvement items. . Er jf Further, the NRC Team Leader stated that the 22_ improvement items would "" not appear in the written report and did not require a response, but that I the NRC team believed the licensee should consider them for enhancing their emergency preparedness program. 1 > , % > -[
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hw x ' 4 ACRONYMS AND INITIALISMS
: ~ (& . * > }_y 1 ,- 1 ASCII: lAmerican' Standard Code of Information Interchange- ' , LABT- ~ Automatic' Bus _-Transfer.(switch) ' COP- Conference OperationsiNetwork ' CPU' . -Central Processor Uniti CSTWLL Condensate Storage Tank Water Level :CR ;v/ :. Control _ Room
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, ) CRT : :DAS _ Cathode Ray Tube. Data Acquisition Subsystem- 'DCRDR Detailed Control Room Design Review DCC Document Control Center: Emergency Action Level ' :EAL' . ; w ;-EAGLE - Emergency Assessment of Gaseous and Liquid Effluents - EI' Energy Incorporated M ENS. Emergency Notification System EOC: Emergency Operations Center E0P ' Emergency Operating Procedure ' E0F - Emergency Operations Facility EPA . Environmental Protection Agency A LEPIP . ; Emergency Preparedness Implementation Procedure , ERFCS- ~ Emergency Response Facilities Computer-System , ' ERE' . Emergency Response Facility- o ESF; Emergency Safety Features . - - * . . FCS- : Fort Calhoun-Station ;e ' /FLIC. ~FLIC Computer Software Package ~ GDC~ General Design Criteria " GSE .(0 PPD) Generating Station Engineering ' s >- HOST- Supervisory Computer System < , -HVAC Heating,- Ventilation, and Air Conditioning - ' , ~ICCI -Inadequate _ Core Cooling. Instrumentation- INPO ' Institute ~for Nuclear Power Operations
in -ISC' -Intelligent Systems Corporation C KV' - Kilovolts f.
- LOCA- Loss of Coolant' Accident ' - MAX IV . Computer Software Package MAXNET Computer Software Package " , MPP Miscellaneous Power Panel- / ' .NAWAS- National Weather Advisory Service .NRC Nuclear Regulatory Commission Omaha Public Power District ' ' . . 0 PPD - OAD. Operational. Acceptance Demonstration .01 Operating-Instruction OSC: Operations Support Center PABX -Private Automatic. Branch Exchange- PARS Protective Action Recommendations PASS . Post' Accident Sampling System
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, -PSIG' Pounds Per Square Inch Gauge QSPDS Qualified Safety Parameter Display System RERP Radiological Emergency Response Plan RCS- Reactor Coolant System - - u_
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' ' 2 . ' ?.l;. . _ , ' ' ' . SPDS ' Safety Paramethr Display System STA Shift Technical Advisor TSC- Technical Support Center E - UPS Uninterruptable Power Supply - , USAR -Updated Safety Analysis Report
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