ML20213A702
| ML20213A702 | |
| Person / Time | |
|---|---|
| Issue date: | 04/21/1987 |
| From: | Zech G NRC OFFICE OF SPECIAL PROJECTS |
| To: | White S TENNESSEE VALLEY AUTHORITY |
| References | |
| NUDOCS 8704280206 | |
| Download: ML20213A702 (81) | |
Text
. .
OfffCity
'APR 2 i W Tennessef Valley Authority ATTN: J r. S. A. White Manager of Nuclear Power 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Gentlemen:
SUBJECT:
TRANSMITTAL OF DRAFT IE INSPECTION PROCEDURE 82412, EMERGENCY RESPONSE FACILITIES APPRAISAL, REV. 3 This letter transmits a copy of the most recent draft of IE Inspection Procedure 82412 entitled, " Emergency Response Facilities Appraisal.' Revisions to this procedure will be transmitted to you as they become available prior to the appraisal at your facilities. This procedure will be used to verify that your Emergency Response Facilities (ERFs) meet the requirements of 10 CFR 50.47(b), Appendix E of 10 CFR Part 50 and orders and license conditions issued to implement Supplement 1 to NUREG-0737.
The ERF Appraisal will be conducted using the team approach and will follow the methodology outlined in Appendix 1 of the procedure. Each major inspection area states the acceptance requirements followed by a series of questions used in determining if the requirements have been met. The questions are intended as an aid to provide an overview of the areas of interest and guidance to prevent oversights in critical areas. The questions are not intended to be all inclusive.
The ERF Appraisal is planned to be conducted in conjunction with the next scheduled annual emergency preparedness exercise inspection at your facility.
Final determination and scheduling of the appraisal is dependent on resource availability. The ERF Appraisal team leader will contact the appropriate cognizant emergency preparedness personnel in your organization through the NRC Office of Special Projects inspection staff in order to establish the schedule and coordinate the logistics.
Should you, have ,any questions regarding the ERF Appraisal procedure, please contact this office.
Sincerely, Gary G. Zech, Assistant Director Inspection Programs Division of TVA Projects Office of Special Projects
Enclosure:
(See page 2) 8704280206 870421 DR TOPRP EUTT D
~
-e.
}
L.
Tennessee Valley Authority 2
Enclosure:
IE Inspection Procedure 82412 pc w/ encl:
vH. P. Pomrehn, Site Director M.BrownsFerryNuclear' L. Gridley, Director Plant Nuclear Safety and Licensing l
R. W. Cantrell, Acting Director
/ Nuclear Engineering l W . J. May, Site Licensing Manager di. L. Abercrombie, Site Director
/ Sequoyah Nuclear Plant VM. R. Harding, Site Licensing j Manager vG. Toto, Site Director A. Mcdonald, Site Licensing
/.WattsBarNuclearPlant Manager bec w/ encl:
J. N. Grace,.RII
. G. Keppler, OSP
. D. Ebneter, OSP
. A. Zwolinski, OSP
. D. Liaw, OSP
/S: D. Richardson, OSP
/S.' R. Connelly, OIA
' P. Barr, RII RC Resident Inspectors I NRC Document control Desk i State of Alabama i State of Tennessee l
RII RI RII RII RI RII. -
ETesta
' J8 TDecker
, a W C L, Yr Algnat+onis FMc y frod Dolin%PStohr rr 4/ 3/87 4/3/87 4/ ' /87 4/f/87 4/j/87 4//f/87 4/70/87 4/y /87
___a
> a-4 Rev. III (2/18/87)
Page 1 of 9 DRAFT i
IE INSPECTION PROCEDURE 82412 EMERGENCY RESPONSE FACILITIES APPRAISAL PROGRAM APPLICABILITY: 2515 82412-01 INSDECTION OBJECTIVE To determine that the Emergency Response . Facilities (ERF) at licensed nuclear
' power plants and at plants applying for an operating license are capable of supporting those functions needed to assist licensees in taking adequate protective measures in the event of a radiological emergency.
82412-02 INSPECTION REQUIREMENTS Verify that the ERFs meet the requirements of 10 CFR 50.47(b), Appendix E of 10 CFR Part 50 and orders and license conditions issued to implement Supple-ment 1 to NUREG-0737.
82412-03 INSPECTION GUIDANCE 03.01 Conduct appraisa's and inspectinns of comcleted facilities usirg the gu':a-ce centaine: r t" s c-::etu-e to review i .c'eva',3:e t*e
- . h . D; 03.02 This guidance is aimed at the post-implementation review (discussec in Section 8.4.2 " Documentation and NRC Review" of Supplement I to NUREG-0737).
23.C3 The EDr appraisal is to be conducted using the eethodology ir Appendix 1 of this procedure,
- a. This methodology provides a series of questions that address the criteria for the Technical Support Center (TSC),
Operational Support Center (OSC), and the Emergency Operations Facility (EOF) in performing emergency response functions. For each major inspection area the acceptance requirements are stated and then followed by a series of questions to provide a scope in determining if the requirements have been met. The l
questions provided on worksheets are intended to provide the reviewer pertinent guidance to perform a comprehensive evaluation of the ERFs and to verify that when fully implemented they will meet the requirements of 10 CFR 50.47(b),
Appendix E of 10 CFR Part 50 and Suppl.ement I to NUREG-0737.
The " bullets" or " thought joggers" which are written under some of the questions are to assist the reviewer in evaluating the licensee's design by suggesting possible methods of meeting the requirement or problems that should be considered. These I
l
-- - , - + - - -w-*,--o- -,-w-y- ,- w - +- w- -,r - - -
-p--- -
Rev. III (2/18/87)
Page 2 of 9
" bullets" are not requirements and need not be addressed to demonstrate compliance with either Supplement 1 to NUREG-0737 or the regulations,
- b. The questions and " bullets" in Appendix 1 are not all inclusive. i They are intended as an aid to a knowledgeable individual, to l provide an overview of the areas of interest and guidance to I prevent oversights in critical areas. Thus each question or
" bullet" may not be answered, but all major areas must be explored and a sufficient number of questions answered to detennine whether or not that area can be accomplished. The procedure will be used during all post-implementation appraisals of final ERFs; however, flexibility will be pennitted in the application.
03.04 Each inspection area in Appendix 1 has some items with a certain degree of interchangeability. Some items are repeated in two or more areas. These repeated items need only be evaluated once.
03.05 The program design anticipates that the appraisal of the final ERFs will be conducted using a team approach. The appraisal will be conducted at each operating power reactor site and any site for which the licensee has made an application for an operating license, contingent upon completion and approval of final ERFs. The appraisal should be conducted in conjunction with the licensee's annual emergency preparedness exercise inspection and to the extent possible the appraisal team will also be utilized for the exercise inspection (IE Procedure 82301). If a licensee has a SALP 3 rating in EP or the annual exercise cannot be scheduled between Monday afternoon and Tuesday evening, an ERF appraisal separate from the annual exercise should be considered. A separate appraisal should also be considered, if in the opinion of the region, the inspection logistics or other conditions make it impractical to complete the ERF appraisal / annual exercise inspection within one week. In the event a separate appraisal is selected, if I possible, it should be scheduled in conjunction with the licensees '
" dry run" prior to the annual exercise or during operational drills involving each ERF. The schedule for completion of these inspections will depend on the completion dates for the final ERFs negotiated b Regulation (y the Project NRR), when theManager in the Office annual exercise of Nuclear is scheduled and otherReactor considerations discussed above.
03.06 Appraisal team personnel: ,
- a. The appraisal team will consist of a regional team leader and personnel with expertise that encompasses the following disciplines: '
- Regional Team Leader
- Regional Inspector - OSC(,) i 1
(a) To be provided only during combined ERF appraisal / annual exercise to l assist regional team leader and perform assigned exercise observation and appraisal items.
1
Rev. III ,
(2/18/87) l Page 3 of 9 4
e Reactor Operations - Control Room, TSC e Source Tenn Assessment - TSC e Oose Assessment - EOF Operations e
l HumanFactorsEngeering
!
- Computer Science
- e Meteorology - EOF Dose Assessment ;
! b. Some of the team members comprising the above disciplines also
- possess the expertise necessary to cover all areas of interest
- during a concurrent exercise inspection. The recommended exercise. assignments for these members are indic6ted after their discipline above. The team leader, human factors
- . engineer and computer scientist who nonnally will not be assigned an exercise inspection responsibility will be free to move from one ERF to another to observe appropriate areas of
, interest. Use of the same team to observe the exercise allows team members the opportunity to observe the licensees emergency organization and facilities in full activation. Information
-gathered during exercise observations will be valuable for the following appraisal.
- 03.07 The team leader, provided by the NRC regions, will manage both the appraisal and the exercise inspection. Other team members i will be professionals provided by the Office of Nuclear Reactor l Regulation, the Office of Inspection and Enforcement, NRC con-l tractors, their subcontractors and .the NRC regional staff. Con-l tractors and their subcontractors will be under direction and control of a contractor team leader. Each NRC team member and the contractor team leader will be under the direction and control of the regional team leader. Team members will be provided specific areas to appraise on an independent basis. Although appraisal outlines and guidance are provided, reasonable . flexibility will be allowed each member in the conduct of these appraisals to account for the plant specific character of the ERF design. At the dis-cretion of the team leader, an indepth review greater than defined by the scope of the questions in Appendix 1 may be pursued for areas where weaknesses are suspected. Each onsite appraisal will be preceded by dedicated advanced preparation and familiarization of site-specifics such as plant design and layout, final ERF conceptual design, and emergency preparedness appraisal findings. During this period a major portion of the review of the structures, equipment, models and hardware design, and emergency procedures for the ERFs should be perfonned.
03.08 The appraisal of the final ERFs will be performed at the licensee's nuclear power plant site and will involve the evaluation of 1) faci-lities, 2) equipment, 3) procedures, and 4) interviews. Upon completion of the onsite appraisal, a fonnel appraisal report will be written by the NRC Region. Discussions and coordination of report .
findings may necessitate conferring with team members. ]
(b) Team member will be provided only where the ERF information management !
system uses ADP techniques.
Rev. III (2/18/87)
Page 4 of 9 4
- 03.09_ Appraisal team assignments
- a. A matrix recomending item assignments and 'a blank matrix to i note the assignments made for each team member is given in Appendix D. It should be noted that for a number of items more i than one team member is assigned and that one of these team members has been assigned the responsibility to write the i
evaluation of each of these items for the appraisal report.
4 The remaining designated team members will provice supporting i inputs on an as requested or as observed basis, depending upon the team leader's judgment and the needs of the team member responsible for writing the iteft for the report. Should any supporting or other team member observe any potential problem area (s) warranting further evaluation by the team member i responsible for preparing the applicable portion of the report, i
the item should be brought up for discussion with the j responsible team member. Should team members responsible for
! preparing specific sections of the report find they may not be l able to complete all assigned sections, the team leader shall j be alerted.
i b. Team members will coordinate their activities to minimize the ,
need for the licensee to demonstrate the same equipment or
- process more than once (e.g., demonstrations of Data Acquisi-i tion Systems, and dose asmsment systems, and discussions of j complex programs or docume..ts.)
! 03.10 The appraisal findings must specify if there is reasonable assurance that the final ERFs can and will perform their functions in the event
, of a radiological emergency. Violations and deficiencies must be
! identified and related directly to Supplement 1 to NUREG-0737, to
- the standards of 10 CFR 50.47(b) and 10 CFR Part 50, Appendix E, or the ability to perfom intended functions. Deviations must be
! referenced to specific connitments by the licensee in the FSAR or i
other documentation. Open items shall include incomplete systems or areas where the licensee agrees to make changes prior to the submission of the appraisal report. Although the body of the ERF i Appraisal Report may reconnend improvements in the ERFs to enhance
! their operational capabilities, only violations, deficiencies, deviations, and open items shall be included in the report findings.
Deficiencies, deviations, and open items will be handled in i accordance with regional policy and violations will be handled under j normal inspection and enforcement procedures.
l 03.11 The following schedule should be adhered to when initiating an ERF i appraisal:
- a. In preparation for the ERF Appraisal the team leader should
- begin working with licensee emergency fireparedness personnel to establish the schedule. If the appraisal is to be combined
, with the annual exercise, the exercise must be scheduled between Monday afternoon and Tuesday evening to allow for training, badging, pre-exercise meetings, and site access; i Wednesday, Thursday, and Friday morning for appraisal
i i
- Rev. III !
(2/18/87)
Page 5 of 9 activities and Friday afternoon for an exit interview. In the i case of a combined inspection effort it is suggested that the i
exercise critiques be scheduled separately from the ERF appraisal exit meeting. Because of the heavy adninistrative i burden placed on the team leader by a combined inspection the regional inspector provided with the team could be used to j assist with or perfom some of these functions. ,,
- b. At least 15 days prior to the projected onsite appraisal, the team leader will contact plant management and the Resident Inspector to arrange for team access and workspace. This will be confinned by the region, including detailing the schedule for the appraisal, team composition (by name and affiliation) and other appropriate logistical details,
- c. Two weeks prior to team arrival at the site the team leader
, will provide to plant management a proposed schedule for
. perfonning various appraisal and exercise activities and a list of personnel, including individuals who may not be part of the
- plant staff but who may be requested to provide infonnation in particular areas. These personnel should be requested to be at
{ the plant at the times their area of expertise is to be
- evaluated.
, d. Appendix 4 of this document provides a suggested form to be completed by the licensee that will provide the team with the
- names, organization and telephone numbers of persons to be
. contacted and reference documentation for each area. In order I
to aid the licensee in scheduling his resources this form
, should be provided 4-6 weeks prior to the appraisal and the
! team leader should request its return prior-to his. arrival on site, j e. Appendix 5 of this document provides a list of various 4
documents and other information needed by team members to conduct the appraisal. Appendix 5 should be supplied at the same time as Appendix 4 to provide licensee contact personnel adequate time to locate and collect the indicated documenta-tion.
- f. A meeting of the ERF Appraisal Team will be held to prepare the ,
team to conduct the appraisal and to familiarize them with site '
specific conditions. This meeting should be scheduled in the :
geographical location of the plant site before the appraisal is I initiated. The specific time and place of the meeting should be set at the discretion of the team leader. The information '
) covered during this meeting should include the following:
o discussion of the licensee's management and emergency
{ organization l 1 l
1 i
.~ _ ._ . ._ _ _. _ _ . . _ _ - - . - . .
4 Rev. III i (2/18/87)
Page 6 of 9
)l e coordination of the team member assignments for both the
- exercise and the appraisal and the assignment of specific
! team members to prepare the various portions of the l t appraisal report ;
e relationship of the emergency functions among the various i ERFs for the specific site ,
e site specific aspects of the licensee's Emergency Plan and EPIPs e time phasing of the accident scenario for the exercise o work space and arrangements provided for the team by the licensee e review past or existing problems identified during i emergency preparedness exercises and inspections e cover exercise items contained in " Discussion" section of j IE Procedure 82301 exercise.
I
- g. A meeting between team members and the licensee's expert and key personnel should,be scheduled for Monday. This meeting will offer team members an opportunity to meet their primary i licensee contacts, schedule interviews, and identify additional personnel or resources necessary to provide appraisal information. If a combined ERF appraisal / annual exercise inspection is being conducted this meeting might be scheduled following an exercise controller or player briefing where many of the licensee's key personnel would be in attendance.
03.12 An entrance meeting with licensee management personnel should i be scheduled as soon as practical following team arrival at the site. During the meeting team personnel should be introduced, the inspection effort explained, the schedule discussed. and i needs for implementing the inspection explained. In addition j this meeting will allow licensee management the opportunity for -
i requesting additional infonnation or clarification concerning j the conduct of the concurrent appraisal / exercise.
l 03.13 Evaluation of emergency data acquisition systems:
- a. As a part of evaluating the information management and data acquisition system for the TSC and the EOF, the availability of i
the report on the implementation of Regulatory Guide 1.97 should
- be determined. This report is required for each site by l Supplement I of NUREG-0737 and must be submitted by the licensee describing how the requirements are to be met. Deviations from the guidance are explicitly shown and a supporting justification or alternatives are presented in this. report. The NRC Head-quarters Technical Coordinator in the Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement will determine the availability of this report or
! any other SER or NRC evaluations of the licensee's submittal, i
Copies will be provided to the individuals performing reactor oper.-tions, source term assessment, human factors, engineering, j meteorology evaluations, and regional team leader to assist j them in evaluating the adequacy of the TSC and E0F database.
i i
,,e,,,..-.- , ~,--.,--,..n,.,, ,, ,,,,,..n, .n- _- nn.,,c__,,,-,- - , . _.. -_,,,,.,,..n,n.,,-,,,.,.n<
Rev. IU (2/18/87)
Page 7 of 9 ,
1
- b. In addition, if the licensee states that the Safety Parameter Display System (SPDS) is a part of the emergency data acquisition
- system for the TSC and/or the E0F, then the SPDS must be installed in these facilities and be operational. The evalua-tion of the SPDS as a part of this ERF appraisal will be 3
performed only for its adequacy as a part of the emergency data acquisition system for the use of TSC andfo.r the EOF and not as an operator aid in the Control Room. The adequacy of the SPOS as a part of the emergency data acquisition system will have no j bearing on its acceptability as an operator aid (reference Supplement I to NUREG-0737 item 4).
03.14 Preparation of the ERF appraisal report:
- a. Prior to leaving the site a written sununary of findings must be provided to the team leader. No later than two weeks after leaving the site, all non-NRC appraisal team members will prepare a final report to the Regional Team Leader and the NRC Headquarters Technical Coordinator evaluating the areas assigned, and giving their findings during the appraisal and should include a list of licensee personnel with whom they had contact by name and title. The findings must provide the facts to justify any significant deficiency. These reports shall be used by the Team Leader to prepare the draft ERF appraisal report.
- b. No later than three weeks after leaving the site, all nonregional 4 NRC appraisal team members will submit their final written 4 reports to the Regional Team Leader and the NRC Feadquarters Technical Coordinator through their Branch Chief or responsible supervisor. The Team Leader will prepare a draft ERF Appraisal
) Report within one week after receiving the last team members l- input and will submit copies to the NRC Headquarters Technical 4
Coordinator and each appraisal tear member for review, coments i and corrections. The NRC Headquarters Technical Coordinator
- and appraisal team members will provide their coments and i corrections to the Team Leader no later than two weeks after j the receipt of the draft appraisal report. -
- c. Although the time frame for the issuance of this report is not in accordance with the criteria set forth in IE Manual Chapter 3 0610-07 it is acceptable for this program. No later than 60 days after leaving the site or within 35 days of receiving the ;
last team members input, the NRC Region will provide a final j ERF Appraisal Report to the licensee or applicant signed by the I l
appropriate regional management and the Team Leader. Also this report will have the concurrence of the, Chief. Emergency Preparedness Branch, Division of Emergency Preparedness and Engineering Response, Office of Inspection and Enforcement. If the report contains identified deficiencies or deviations that require the licensee to remove or rip out ERFs or equipment
- that had been installed in good faith to meet previous guidance t
t 1
Rev. III (2/18/87)
Page 8 of 9 in order to meet the requirements of Supplement I to NUREG-0737, the concurrence of either the Director, Office of Nuclear Reactor Regulation or the Director, Office of Inspection and Enforcement will be obtained prior to the issuance of the report.
- d. The appraisal report will follow standards and guidelines given in Manual C.hapter No. 0610. " Inspection Reports - Format and -
Content." The report will clearly identify all violations, deficiencies, deviations, and open items observed during the ERF Appraisal in the findings. These items and any other items which the licensee has agreed to correct anytime prior to the issuance of the final ERF Appraisal Report will be tracked for correction within a schedule to be negotiated between the licensee or applicant, regional management and the Project Manager, Office of Nuclear Reactor Regulation. Prior to the NRC exit meeting with the licensee corrective action completion
, dates should be negotiated and may be included in the report.
Renegotiation of these dates after the report has been issued should be accomplished through nonnal channels. When corrections cannot be agreed to, reconsnendations for possible further regulatory action will be forwarded to the Director of the appropriate licensing division of Office of Nuclear Reactor Regulation. If the correction of any deficiency or deviation requires the licensee to remove or rip out ERFs or equipment that were installed in good faith to meet previous guidance in order to meet the requirements of Supplement 1 to NUREG-0737, the approval of any such orders or agreements will be obtaired from either the Director, Office of Nuclear Reactor Regulation or the Director, Office of Inspection and Enforcement.
- e. The exercise report should be prepared and issued in accordance with current inspection guidance and regional policy.
02412-04 REFERENCES 04.01 Appropriate sections of the following documents may be used to prepare team members for the appraisal:
U.S. Code of Federal Regulations (CFR). 1982. Title 10, Part 50, " Licensing of Production and Utilization Facilities," Appendix E " Emergency Plans for Production and Utilization Facilities."
U.S. Code of Federal Regulations (CFR). 1982. Title.10, Part 50.47,
" Emergency Plans."
U.S. Nuclear Regulatory Commission (NRC). 1980a. Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Sup) ort of Nuclear Power Plants. NUREG-0654, FEMA-REP-1, Rev. 1, Wassington, D.C.
Rev. III (2/18/87)
Page 9 of 9 U.S. Nuclear Regulatory Comission (NRC). 1983. Clarification of TMI Action Plan Requirements, NUREG-0737, Supplement No.1. Washington, D.C.
U.S. Nuclear Regulatory Comission (NRC). 1981. Functional Criteria for Emergency Response Facilities. NUREG-0696, Washington, D.C.
U.S. Nuclear Regulatory Comission (NRC). Inspection and Enforcement Manual, IE Inspection Procedure 82301, " Evaluation of Exercise for Power Reactors."
U.S. Nuclear Regulatory Comission (NRC). Inspection and Enforcement Manual, Chapter 0610.
04.02 Other useful information can be found in the following documents (ifavailable):
- Licensee's ERF Conceptual Design Plan e The NRC review of the Conceptual Design Plan
- Documentation of source term and dose assessment model e The Emergency Preparedness Appraisal Report e Available Hardware Specifications
- Emergency Preparedness Implementing Procedures associated with ERF Operations
- Emergency Preparedness Exercise Reports e Complete case file relative to emergency preparedness (e.g.,
inspection reports and resulting outstanding items list; corre-spondence on exceptions to NRC policy) e Licensee's plan for implementation of Regulatory Guide 1.97 and the NRC evaluation w mer .--
Rev. 11 Page 1 of 42 APPENDlX 1 APPRAISAL CRITERI A AND REQUIREFENTS 1.0 TECHNICAL SUPPORT CENTER (TSC)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
?j 1.1 PHYSICAL FACILITIES 1.1.1 Destan 10 CrR 50.47(b)(8)(11), Supplement 1 to NUREC-0737 (TSC requi rements a,b.c,d,e.f,h, and k) 1.1.1.1 Stre and Layout
- a. Is the size of the TSC adequate to acconnodate and support NRC and predesignated ifcenseepersonnel,equipmentanddocumen{ation to perform the intended functions? e.g., t
- work space for individual and working group functions
- work space for document and draafng review space for equipment and Instru-mentation use, maintenance end repair,
- b. Is the layout desigaec to accomodate and suppeat NRC a*d licensee predesignated personnel, equipment, and documentation and provide unimpeded personnel traffic and information flow?
j
- c. How were the TSC size and layout verified as adequate? ,
1.1.1.2 Location is the TSC located within the plant site protected area and situated to facilitate the necessary interaction ofth the Control Room, OSC, E0r, and other personnel involved in the essergency ?
Isote 1: The " bullets" following the questf ons are $ requirements and need not be addressed to demonstrate compliance with either Supplement 1 to NUREC-0737 or the regulations (see inspection Caldence items 03.03a and b).
Rev. Il Page 2 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIROMENTS/ COMMENTS 1.1.1.3 Structure
- a. Was the TSC built in scenedence with the Uniform Butiding Code? ,
- b. Is it capable of operating uninter-ruotefduringperiodsofactivation?
e.g., t adverse weather conditions
- loss of offsite power systems reliability 1.1.1.4 Habitability / Environment
- a. Does the TSC provide redlological protection and to assure that any person merking in the TSC would not receive rediation exposure in excess of 5 rem shole body or its equivalent (25 rem to t$ethyroid)fo{thedurationofthe accidenti e.g.
gemma radiation shielding isolated ventilation ventilation filteafag for alabo*ne todine and particu'ates reserve air supply positive pressure atmosphere
- b. What are the bases for determining theegequacyofTSChabitability, e.g., t design basis for outside environ-ment (10 CrR 50, Appendia A, NUREC 0578, etc.)
- documentation of calculations and measurements Ente lt 1810 page 1.
1 i
R:v. ll Page 3 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE TEOUIREWENTS/COMwENTS
- c. Is the TSC environmentally cortrolled to provide air temperature, huridity and cleanliness within acceptable limits for personnel and ,.',
eouipment? e.g., e '*
air conditioning humidity control filtered ventilation (non-emergency)
Ifghting emergency lighting sound levels restroom facilities 1.1.1.5 Display interface I
- a. What type of data pisplays are used in the TSC7 e.g.,
analog and digital meters
- CRT chart recorder status boards
- manual techniques hard copy availability.
- b. Are the information displays adequate such that data is aporopriately labeled, legible, updated in a timely manner and organized to succort the intended function 7
- c. Are the information displays adequate in number to support TSC functions, easy to update and
- facilitate user access?
I
- d. Are data and information trending l displays adequate and appropriate to support the intended functions of the T5C?
i l
- e. Is user dncumentation readily j available and easy to usef !
MUTE 1: IBID page 1.
Rov. Il Page 4 of 42
)
1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /COM8ENTS 1
1.1.2 Radiologleal Equipment and 10 CrR 50.47(b)(8)(11) 10 CrR 50, Appendia E Supplies {lV.E.1, Supplement 1 to NUREC-0737 (TSC requirement f) 1.1.2.1 Radiation Monitoring
- a. Does the TSC have adequate equipment and instrumentation (quantity, type, range and sensitivity) to determine dose f rom direct radiation esposures and airborne concentrations underaccidentconditjons (see 1.1.1.4)? e.g., s
- rate measuring gauna instrumenta-tion (portable or flued)
- sin sampling equipment (portable or fixed)
- high range dosimeters
- TLDS or film badges
- b. Is the equipment adequately maintained, calibrated and inventoried?
- c. Are adequate radiological instruments and equipment of accrocriate quantity, type, range and sensitivity available in the TSC for use in traveling outside the TSC under accident conditions?
- d. Is a procedure available for tracking dose to individual TSC personnel throughout the course of an accident?
1.1.2.2 Protective Supplies Are there adequate and dedicated protective supplies available or readily accessibly to all TSC and augmentation personnel sufficient to suppo{ttheirassignedfunctions?
e.g., t
- respiratory protection .
- decontamination supplies and equipment
- protective clothing
- Kl leTE 1: I8ID page 1.
R:v. Il Page 5 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC1 (contd)
CRITERIA ACCEPTANCE REOUf PEMENTS/ COMMENTS 1.1.3 Non-Radiologleal Equfoment and to CrR 50.47(b)(8)(9); 10 CrR 50, Appendix E ilV.E.9, Supp1tes Supplement 1 to NUREC 0737 (TSC requirements a,g,h and ()
1.1.3.1 Records / Drawings Are appropriate records 'and drawings adequatelystored, main {ainedandupto date 17 the 75C7 e.g., :
- as built plant drawings
- FSAR
- EPIPs
- schematic diageams
- procedures (operating and emergency operatino)
- notification Ifsts
- equipment manuals.
1.1.3.2 Support Supplies Are adequate supplies and information available,stojed,andmaintainedin the TSC? e.g. :
maos (10 and 50-mile EPZ1
- plant floor diagrams
- Isopleths
- calculators
- means for data t ending
- pens, pencils, paper
- conversion tables and other references
- availability of emergency vendor asststance
- computer paper and ribbons.
1.1.3.3 Power Supp1ies Do the power supplies assure that the TSC will function without (gterruption during an emergency? e.g., s !
- diesel power supplies a battery power supplies
- alternate sources of offsite
! Power.
1 IGTE 1: 1810 page 1.
l_ - _
R0v. 11 Page 6 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.2 lNr0RMATf0N MANACEMENT 1.2.1 Variables Provided 10 CrR 50.47(b)(9); Supplement 1 to NUREC-0737 (TSC requirements h and k) and Regulatory Culde 1.97 1.2.1.1 Reg. Cuide 1.97 Variables Which of the Regulatdry Guide 1.97 variables are available in the TSCf (See Appendix 2).
1.2.1.2 Other Variables
- 4. If all the appropriate Reg.
Culde 1.97 variables are not available, are there other variables that can be substituted for the Reg. Cuide 1.97 variables in the 15C to a11nn it to adequately perform its function of technical and logistical support to the Control Room and perform E0r functions untti the E0e is staffed? If not, what additional information is necessary?
- b. abat other vgriablesareavailable in the TSC7 e.g.
- offsite monitoring
- weather forecast and advisory
- regional meteorological inf ormation (e.g. , NWS) evacuation time estimates
- medical and emergency assistance i n fo rmation other plant information.
- 1. 2.1.3 Relationship to Functional Needs
- a. Are the variables provided suf ficient to a11ow the TSC to perfonn its designated functions?
Note It IBID page 1.
l l
R:v. Il Page 7 of 42 1.0 TECHNICAL SUPPORT CENTER (TSCI (contd)
CRITERIA ACCEPTANCE REOUIREMENTS/CDe4ENTS
- b. Are the available 1.97 variables adequate to determine continuous removal of heat from the core and associated cooling systems? e.g. ...
- loss of high pressure coolant f*
I injection system, such as HPCI, RCIC, or safety injection
- loss of low pressure coolant injection systems, such as RHR, LPCI or core spray
- inability to depressurire the coolant system to allow low pressure coolant injection
- loss of service water systems, such as RHR or component cooling service water or emergency service water
- loss of auxillary or emergency feedwater.
I c. Are tne available 1.97 variables adequate to determine threatened or actualdegradatjonofthefuel and fuel cladding? e.g. ;
- sustained operaticas outside clant thermal limits
- continuous rod withdrawal accfdents
- cold water accidents
- core unenverage
- subcooling margin
- Improper chemistry, such as reviewing routine sample analysts
- monitoring of parameters that reflect the condition of the fuel and fuel cladding, such as high levels of radioactivity in reactor coolant or main steam line monitors
- monitor in-core thermocouples
- high radiation levels or high l fission product activity in contalmeent.
Itete 1: 181D page 1.
l
Rev. Il Page 8 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE RE011REMENTS/ COMMENTS
- d. Are the available 1.97 variables adequatetodetermineryactnrcoolant system Integrity? e.g. :
- pressurfter level or, reactor vessel level
- 1etdown and makeup flo6e rates high radittion levels or fission product activity In contairwent high activity in the steam system (PWR)
- a physical break or crack in the coolant system piping failure of relief valves to reseat.
- e. Are the available 1.97 variables adequatetodetermineprimary containment integrity? e.g.
- a physical break or crack in a containment penetretion
- containment overstressing by high temperature and pressure
- failure of a conta8ement isolation togic or valve to operate or inability to Isolate H, concentration in containment
- high radiation or radioactivity levels in containtnent
- Interfacing systems LOCA
- f. Are the available 1.97 variablet adequate to determine the operability, capacity, and Integrity of the Ifquid, solipandgaseousradwastesystems?
e.g.
- failure of waste gas holdup tanks or their relfe' valves
- hydrogen recometner or offgas piping emplosions
- inadvertent discharge of untreated or concentrated wastes.
Itote 1: 1810 page 1.
i
Rev. Il Page 9 of 42 1.0 TECmlCAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
- g. Are the available 1.97 variables .
edequate to determine the extent of damageresultingfroyafuelingorfuel pool accidenti e.g.
- shipping cask or other heavy Iced dropped over spent fuel pool
- heavy equipment dropped over open vessel
- loss of water level in fuel pool or vessel
- high radiation or radioactivity levels in containment, the fuel hand 1tng area, fuel pool, or other auxillary plant areas.
- h. Can an evaluation be conducted of both the entsting and projected status of the core / containment and environs to .
support adequate determination of proper protective action
- recomendatIons uti1121ng
(1) the verfables provided in Section 1. 2.1.1, 1. 2.1.2, and 1. 2.1. 3 and their 'Lactional relat'oeship determine the status of plaat Systems and conditions for NUREC 0654, Appendia 1, General Emergency example initiating condition No. 4, (ii) the current radiation monitor reading in containment, process area, and effluents, and meteorological data and field monitoring results, *
(111) those offsite factors that influence the effectiveness of protective action recommendations?
- f. Are the required data representative nf the quantities to be
, seasured? e.g. t
- activity in coolant 1
- activity in containment
- coolant level in core area leste 1: 1810 page 1.
Rev. Il Page 10 of 42 1
1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTAMCE REQUIREMENTS / COMMENTS 1.2.2 Data Acquisition 10 CrR 50.47(b)(9): Supplement 1 to NUREC-0737 (TSC requirement h). General Design Criteria (CDC) 24 Regulatory Cuides 1.97, Rev. 2, and 1.75.
- 1. 2.2.1 Data Collectiom Method
- a. How are the data acquired? e.g.
- video technioues
- digital or aralog instruments
- computerized acquisition system voice communication
- b. Is the capacity of the data ecliection equipment suffletent to access all of the data to be transmitted to the 75CT 1.2.2.2 Time Resolution
- a. Is the sempling frequency of each variable adequate to ensure detection of significant changes, particularly accident conditions?
- t. Is the time reso'ution for t"e transmission of each of the available variables adequate to assure that no significant data is lost?
1.2.2.3 Isolation
- a. Is the signal isolation performed for those variables obtained from befety systems adequate to assure that the safety systems will not be degraded
]
by the data acquisition system? 1
- b. Was the isolation of the installed system verified and valfdated? If so, how?
l Mote I t 1880 Pape 1.
R:v. Il Page 11 of 42 3
1.0 TECHNICAL SUPPORT CENTER fTSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /C0*ENTS 1.2.3 Data Communications 10 CrR 50.47(b)(9): Supplemens 1 to NUREC 0737 (TSC requirements g and h)
- 1. 2. 3.1 Capacity
- a. What is the channel capacity of 4 the data systemt
- b. It it adequate to meet the needs of the data system under peak load and accident conditions?
1.2.3.2 Error Detection
- a. What techniques are used for error detection / correction?
- b. Does this technique assure error ,
detection from sensor to CPU 7 , !j 1 1.2.3.3 Transmission Between ERrg ,
- 4. What methods are used for data .
tesesmission?
I
- 3. !s data transmission adequate between the TSC, the Control Room, and the E0r?
1.2.4 Cata Analvsis 10 CrR 50.47(b)(4),(b)(9); 10 CrR 50, Appendix E ilV.B. IIV.E.2 and liv.Ds Supplement i to NUREC-0737 (TSC requirement h)
.c
- 1. 2 . 4 .1 Resetor Technical Support ,
- a. Is the data analysis adequate to support the TSC functions?
- b. Will the data analysis facilitate i determination of reactor sta{us past, present and projected? e.g. t
- Forecasting (trending)
- containment pressure vs. time
- containment temperature vs.
time Note 1: ISID Page 1.
\'
Rev. Il Page 12 of 42 TECHNICAL. $UPPORT CENTER (TSC) (contd) 1.0 _
ACCEPTANCE REQUIREMENTS /Co*ENTS CRITERIA
- containment radiation levels vs. time
- conta;nment radioectivity levels vs. time
- containment H 1evels vs. time
- primary coolant temperature vs.
time
- cffoas radioactivity vs. time
- pristrv coolant pressure vs.
time
- primary coolant r-ficactivity
' vs. time
- primary coolant Inventory vs.
time
- power level vs. time l
- primary coolant temperature vs. pressure
- rad waste storage capability vs. time
- radiation levels vs. plant or system location
- makeup water inventory vs.
t i re
- Precalculated relationships of varigbles to accident conditionst e.g.
- containment radiation levels to core conditions (with and ci*hout abnormal coolant system leakage)
- coolant radioactivity levels to
- e' core conditions
= coolant chemistry to core conditions
- H level in containment to containment f ailure and fuel clad f
conditions
- PWR subcooling margin
- containment pressure to 4'
' containment f ailure
- symptoms of potential accident .
sequences .
i 18tD Page 1.
Note 1
Rsv. 11 Page 13 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /COPPENTS
- ARM readings outside containment to containment high radiation monitor readings (with and without containment leakage)
- letdown line, main steam ilne process radiation monitor readings to coolant radioactivity levels
- affert on stack monitor readings of gamma raciation field from containment
- c. Are the parameters displayed in a eenner that makes it easy to determine devigtf ons in parameters from normalt e.g.
- parameters displayed with superimposed operating curves
~
- normal ranges of parameter values for the operating condition displayed or available
- parameters displayed in % of normal or full range values
- d. Is data aralysis per'orted in a manne* easily related to EAL criteria (classification and prctective action decision making) (i.e., do data displays contain the parameters and relationsM os recuired to allow operators to make a clear association with EAL criterialf
- e. Iscurrentgystemstatus available? e.g.
- valve position equipment operation
- pump status 1.2.4.2 Dose Assessment
- a. Dves the data analysis capability provide adequate dose assessment?
Note l IBID pago 1.
R;v. If Page 14 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUiWEMENTS/ COMMENTS
- b. Does the IIcensee have the capability to provide dose projections at the site boundary within 15 minutes of an emergency declarationf
- c. Can the Itcensee make timely plume exposure dose projections and formulate adequate protective action recommendations for the following conditions variable release durations
- variable distances up to 10 mi.
variable meterological conditions
- variable and/or multiple source term (s)
- d. Based on the variables available and calculational methods used, is there adequate information to determine Sourcetermsfo{allpotentialrelease ,
pathways? e.g. t
- effluent monitors
- containment monitors
=
containment leak, rate
'uel storage live time eavir:neental monitor oost-accideat sampling results in-piant radiological monitoring Inoperable or of f scale monitoring instruments.
- e. Do source term methods provide for contribution by radionuclides 7 If sn, what radionuclides and what contributions and how are they determined? e.g.
- real time measurements def ault values
- simplifying assumptions
- laboratory analysis
- f. Are the meteorological varf ables and calculational methods adequate to characterite the meteorological conditions to about 10 elles from the ,
site for release pathways (ground level and elevated releases)?
Note 1: 18tD Page 1
Rev. 11 Page 15 of 42 1.0 TECHNICAL SUPPORT CENTER (TSCl (contd)
ACCEPTANCEREQUIREMENTS/COMM{NTS CRITERIA g.
What meteorological information is Ovailable (nind speed, direction, e.g. :
stability and forecastingi?
- primary measurements system and/or supplemental measurement system (where appropetate)
- histori cal / climatological rel ationships
- precipitation
- teansport and dif fusion modeling
- topography
- supplementary data (e.g., NWS)
- def ault values
- simplifying assumptions
- mixing heights and boun dary layers
- h. Have the onsite
, meteorological monitoring systems historically provided a reliable indication of meteorological variables?
i.
C3n ingestion pathway dose projectiers te made to about 50 i
eiles f ror t"e s te to dete**iae the necessity to deploy radiological monitoring systems in the ingestion pathway EPZ.
(see 1.2.1.2)?
J, Do the dosimetry model calculational methods adequately -
determine thyroid inhalation dose cemitment and whole body dose for appilcable release pathways (both ground level and elevated releases) to about 10 miles from the plant?
k.
Has the sensitivity and uncertainty inherent in the dose assessment been established e.g.and r
factored into the projections?
- source ters
- diffusion and transport
- dosif.etry i 1
I Uote 1: IBID page 1. l l
R:v. Il Page 16 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
- 1. Is the field monitoring data used to correct or modify the dose projections? How? If not, how is the
. data used? e.g. :
- number and availability of measurements adequate deployment plan
- adequate nueber of prepositioned monitoring points
- instrumentation capabilities sampling technique capabilities in-situ measurement capabilities laboratory analysis of samples under accident conditions
- use of data from offsite organizations (DOE, EPA, State, local, etc.)
live time environmental monitors -
- e. Have the computer program (s) used for dose assessment been systematically verified by the licensee or his contractor?
- 1. 2. 4.3 Central processor Capability
- a. Is the g rocessing system capacity edequate to support the data acquisition analysis, display, and storage?
- b. Are there other computational requirements on this processing system?
- c. Do these computational requirements adversely affect the data '
ecquisition system?
- d. In multiuser or multitask systems, how are priorities for competing tasks
]
resolved? '
1 l
Note 1: IBID page 1 i
l
Rav. Il Page 17 of 62 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.2.5 Data Storage Supplement 1 to NUREC-0737 (TSC requirement h) 1.2.5.1 Storage Capabilities
- a. Is data storage adequate to suppyrt the necessary data hand 11ng7 e.g. 3 trending analytical requirements
- b. Is data storage adequate to allow analytical review of plant response to tranglentsforaugmentationpersonnel?
e.g. t
- snort- and long-term data. storage capacity
- data storage methods and accessibility anticipated maximum quantity of data to be stored under accident conditions 1.2.6 v e dels and Systems Reliabilftv aad Validity
- 1. 2 . 6.1 Verification
- a. How was the verification done?
e.g. :
- design docurrentation
- implementation test records or exercise results correlation of readings with Control Room indicators.
- b. Was an irdependent verification performed?
- 1. 2. 6. 2 Computer Based Systems
- a. How has the reliability of the compytersystembeendetermined?
e.g.1 unavailability records !
maintenance logs comparison with similar systems 1 end-to-end test.
100te 1: IBID page 1.
l
R:v. ll Page 18 of 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
- b. Does the data acquisition and storage system have an uninterrupted backup poner supply to assure continuity of data acquisition and storage?
1.2.6.3 Manual Systems
- a. What methods are employed to assure that any data that are manually gathered, processed and/or pisplaf ed in the TSC are reitable? e.g. :
independent sources of information crosschecks confirmation between source and destination use of formal procedures or checkl1sts , 1.2.7 On-Shift Dose Assessment 10 CrR 50.47(b)(9)
Inspection Culdance
- 1. 2. 7.1 Dese Assessmeet Proficiency 1
- a. Dces the licensee have the capability to provide rapid dose projections at the site boundary within 15 minutes to determine emergency classification.
- b. Is the rapid dose projection capability on-shif t, in either the Control Room or Technical Support .
Centers, adequate to scope the magnitude of the potential impacts?
(See 1.2.4.2)
- c. Determine whether the model used by the licensee is consistent with models used by of fsite authorities.
., -,-r -* r
R;v. ll Page 19 of s2 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 1.2.7.2 Dose Assessment Technical Adeouacy
- a. Can the on-shift staff perform dose assessment calculations without interfering with the immediate response to an accident?
- b. Are the assessment procedures designed to allow the operators to inherently arrive at the proper Emergency Action Levels to classify the energencyT 1.3 SUNCTIONAL CAPABILITIES
- 1. 3.1 Control Room Support Supplement 1 to NUREC-0737 (TSC requirement a)
Are the TSC operations capable of providing technical support to the Control Room under accident conditions by performing the following tasks based on the eueecise observation?
- a. Evaluating the status of cruciel p* ant systems,
- b. Determining proper protective actions associated with core / containment conditions.
- c. Forecasting and trending.
- d. Determining the relationship of variables to accident conditions,
- e. Determining corrective actions to maintain crucial plant functions to include consideration of offsite cor. sequences,
- f. Requesting and directing corrective actions implementation by -
the OSC teams.
l
- g. Determining if the facilities, equipment and organizational interf aces support personnel with key roles to
! perform their functions adequately.
t L
__ , - . , ~
R;v. Il Page 20 cf 42 1.0 TECHNICAL SUPPORT CENTER (TSC) (contd)
CRITERIA ACCEPTANCE REQUtREMENTS/C0pgsENTS 1.3.2 Initial E0r fu nctions Supplement 1 to NUREC-0737 (TSC requirement a) is the TSC staff able to perform the EOF functions during an Alert or higher emergency classification and until the EOF is functional including the foll ossing?
dose assessment
- prctective action decisionmaking
- coordination of radiological and environmental assessment offsite.
4 e
+
1
l I
Rev. It' l Page 21 of 42 2.0 OPERATIONAL SUPPORT CENTER (OSC)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 2.1 PHYSICAL FACILITIES 2.1.1 Design 10 CrR 50.47(b)(8); Supplement 1 to NUREC-0737 (OSC functions and OSC requirement b) 2.1.1.1 OSC Location is there an OSC located onsite, separate f rom the Control Room and TSC?
(Note Each assembly staging or other location used for support of the OSC oust be evaluated.)
2.1.1.2 Alternate OSC Location (s)
Ars provisions made to perform CSC functions elsewhere if the primary OSC becomes uninhabitable?
2.1.1.3 Size Layout and Environment
- a. Are the size and layout of the OSC and alternate OSC adequate to provide ar assembly area for all assigned sueeert personne1 and to facilitate performance of supocrt functions and tasks?
- b. How was the OSC layout verified as adequate?
- c. Are environmental conditions (temperature, lighting, noise level) acceptable for operations? .
2.1.1.4 Display Interface
- a. Is informJ ion adequately displayed or made available for use in the OSC fnr planning prior to gispatch of teams into the plant? e.g. :
situational status
- site environmental data
- radiological conditions other.
Note 1: IBID page 1.
Rev. II Page 22 of 42 2.0 OPERATIONAL SUPPORT CENTER (OSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /COPHENTS 2.1.2 Radiological Equipment and 10 CrR S0.47(b)(11); 10 CrR 50, Ap;:endix E IIV.E.1 Sucplies 2 .1. 2 .1 Radiation Monitoring
- a. Are direct radiation and airborne radioactivity monitored in the OSC to assure habitability?
- b. Are radiological instrumentation and equipment to be used by teams entering the plant available or readily accessible to the OSC or alternat{ OSC under accident conditions? e.g., :
radiological monitoring instru-ments
- portable air samplers and filters survey data sheets
- c. Is a procedure available for inventory of OSC radiological instrumentation and equipment?
- d. Is the instrumentation adequately calibrated anc *aintained?
2.1.2.2 Personnel Dosimeters
- a. Are dosimeters with adequate ooerating ranges available in the OSC for each individual?
- b. Is a procedure available for tracking dose to individual OSC -
personnel exposures throughout the course of an accident?
2.1.2.3 Protective Supplies
- a. Are there adequate and dedicated protective supplies such as Kl, respiratory protection equipment, and protective clothing available or readily accessible to all OSC personnel sufficient to support their assigned functions? )
I Ilote 1: 1810 page 1.
R;v. Il Page 23 of 42 2.0 OPERAf f 0NAL SUPPOAT CENTER (OSC) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
- b. How are protective supplies made available at the alternate OSC location?
2.1.3 Mon-Radiologleal Equipment and 10 CrR 50.47(b)(11); 10 CrR 50, Appendia E llV.E.3 Supplies and 4 2.1.3.1 Support Sueplies
- a. Are the following supplies available, adequately stored and maintained and available or accessible to the OSC and the alternate OSC?
e.g.
data sheets and other forms
- plant floor diagrams calculators
- pens, pencils, paper
- re'erences
- b. Are supplies to be used by teams eetering the plant readily accessiele te tge OSC under accident conditions?
e.g.
- 'irst aid equipment
- maintenance equipment and tools damage control equipment
- reference manuals Note 1: 181D page 1.
Rsv. Il Pagi 24 ef 42 3.0 EMERCENCY OPERAft0NS FACILITY (E0r)
CRtTERIA ACCEPTANCE REQUIREMENTS / COMMENTS 3.1 PHYSICAL rActLITIES 3.1.1 Design 10 CrR 50.47(b)(8); Supplement 1 to NUREC-0737 (E0r reautrements t,c,d,e and k and Table 1) 3.1.1.1 Size and Layout
- a. Is the size of the E0r adequate to accowNdate and support re deral, State, local, and licensee predesignated personnel, equipment, and documentation to rform the intended functions? e.g.,
- work space for individual and working group functions
- work space for document and drawing review
- space for equipment and instru-mentation use, maintenance and repair,
- b. Is the layout adequately designed to scovide unimpeded personnel traffic and infomation flow?
- c. ken aere the EOF size 49c layest verified as adequate?
3.1.1.2 Location
- a. is the EOF located as described in Table 1 in the Supplement 1 to NUREC-0737?
9
- b. Which option (1 or 2) was chosen and does the E0r meet all the criteria in the specified option?
- c. Have adequate provisions been made for the NRC site team near the plant if the E0r is beyond {0 miles from the plant site? e.g.,
Note It IBID page 1.
K:v. 11 Page 25 of 62 3.0 ENERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS size of facility
- communications
- radiological monitoring data availability regional approval access during adverse weather conditions
- d. Does the Ecr and the backup Ecr (if required) have adequate physical security to keep out unauthorized personnel and to maintain the EOF in a readiness state?
3.1.1.3 Structure 10 CrR 50.47(b)(11); Supplement 1 to NUREC-0737 (EOF requirements b and e and Table 1) was the E0r and the backup EOF (if required) built in accordance with the Uniform Suilding Code and capable of functioning during adverse conditions such as, loss of electrical power, high winds, heavy rains, floods, ete?
3.1.1.4 Habi ta bi l i ty /Env i ro nment (Habitability requiremeats 'or E0rs located at 10 miles or less f rom plant site. See Supplement 1 to NUREC-0737).
- a. What is the protection f actor (PC) from 3.7 MeV gamma for the E0r areas used for dose assessment, communteations, and decisionmaking?
How was the Pr calculated? .
Rule of Thumb for Pr of 0.7 MeV Canna Inches of Concrete PC (Approximate) 5 6 10 8 15 9 20 10 25 11 30 11.5 40 12 50 13
Rsv. Il Page 26 of 42 3.0 ENERCENCY OPERAft0NS FACILITY (E0rl (contd)
CRITERIA ACCEPTANCE REQUIREMENTS /C0>e8ENTS
- b. Are the EOF areas for dose assessroent, comunications, and decisiontraking protected by HEPA filtration and isolation capability in the ventilation system?
- c. Is the E0r environmentally controlled to provide air teeperature, hur-itity and cleanliness within acceptaD!e lirrits for personnel and equipment? e.g., :
air conditioning hurnidity control filtered ventilation (non-emergency) lighting
- emergency lighting sound levels food, water and restroom facili-ties.
- d. Is the backup E0r adequate to accept the transfer of the dose assessment, comunications, and dec' sic 9 making fu :ticas of the E0r if they-imaryE0r ms: te evacuated?
a g. -
size of the facility work station layout activation tim
- comunications capability
- data availability availability of support supplies
- e. Has the backup EOF been activated during a drill exercise?
- f. Can the licensee demonstrate that dose assessment, comunications and decisionmaking functions can be transferred from the E0r without loss I
of continuity?
- g. Ithere are the other E0r functions performed that are not transferred to the backup EOF if the primary EOF must be evacuated? Are these provisions adequate?
Note 1: 181D page 1.
R:v. 11 Page 27 of 42 3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 3.1.1.5 Display interface
- a. What type of data displays are used in the E0r? e.g.
- analog and digital meters
- CRT
- chart recorders status boards manual teche.icues
- hard copy availability
- b. Are the displays adequate such that data is appropriately labeled, legible, updated in a timely manner, and organized to support the intended function?
- c. Are the information displays adequate in number to support EOF functions, easy to update and facilitate user access?
- d. Are data and information t ending displays adequate and accropr' ate to support the intended fuection of the E0r!
- e. Is user documentation readily available and easily understood?
3.1.2 Radiological Ecuipment and 10 CrR 50.47(b)(11); 10 CrR 50, Appendix E Supplies llV.E.1, ilV.E.3, Supplement 1 to NUREC-0737 (E0r requirement b) 3 .1. 2 .1 Radiation Monitoring
- a. Does the E0r have adequate radiological monitoring capabilities to measure radiation levels that may result in doses of 5 rom to the whole body or its equivalent to any part of the body during the course of the ,
accident? I Note 1: 188D page 1.
R;v. ll Page 28 of 42 3.0 EMERCENCY OPERATIONS rACILITY (E0r) (contd)
CRITER'A ACCEPTANCE REQUIREMENTS / COMMENTS
- b. Does the EOF have adequate radiological and instrumentation equipment available to readily determine dose from direction radiation and airborne radioactivity concentrations under accident conditions (see 3.1.1.4)? e.g., :
- rate measuring gamma instrumenta-tion (portable or fixed)
- air sampling equipment (portable or fixed)
- high range dosimeters
- TLDs or film badges,
- c. Is the equipment adequately maintained, calibrated and inventoried?
- d. Are adequate radiological instruments and equipment of appropriate cuantity, type, range aad sensitivity available in the E0r for use in traveling outside the E0r under eccident conditions?
- e. 15 a procedure available 'or tracking dose to individual Egr personnel throughout the course of an eccident?
3.1.2.2 Protective Supplies Are there adequate and dedicated protective supplies evallable or ,
readily accessible to all E0r and augmentation personnel sufficient to suppo{t their assigned functions?
e.g., 1 respiratory protection
- decontamination supplies and equipment
- protective clothing Kl.
Note 1: 1810 Page 1.
R v. Il Page 29 of 42 3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 3.1.3 Non-Radiological Equipment and to CrR 50.47(b)(5),(11); 10 CrR 50, Appendix E, Supplies flV.E.1, ilV.E.93 Supplement 1 to NLREC-0737 (E0r requirements b,g h and 1) 3.1.3.1 Records / Drawings
- a. Are appropriate records and drawingsadequatelystored,maintgined, and up to date in the E0r; ,,g,, ;
as-built plant drawings FSAR emergency plan EPIPs
- schematic drawings operating procedures notification lists
- equipment manuals.
3.1.3.2 Support Supplies
- a. Areadequatesuppliess{oredand maintained in the E0r? e.g. :
- maps plant fleor diagrams
- 'soplet's
- calculators means for data trending pens, pencils, paper conversion tables and other references
- computer paper and ribbons.
3.2 INr0RMATION MANACEMENT 3.2.1 Variables Provided 10CrR50.47(b)(9); Supple $ent1toNUREC-0737 (EOF requirements g and k) 3.2.1.1 Regulatory Cuide 1.97 Variables Which of the Regulatory Guide 1.97 variables are available in the E0r?
(see Appendix 2)
Note 1: 191D page 1.
Rsv. Il Pag) 30 of 42 3.0 EMERCENCY OPERATIONS FACillTY (E0r) (contd)
CRITERIA ACCEPTA9CE REQUIREMENTS / COMMENTS 3.2.1.2 Other variables a.
If all the appropriate Regulatory Guide 1.97 variables are not available, are there other variables that can be substituted for the Regulatory Cuide 1.97 variables in tne EOF to allow it to adequately perform its functions of protective action decisionmakfeg? If not, what additional information is necessary?
b.
in theWhatotherva{fablesareavailable E0r7 e.g., :
offsite monitoring data from resources other than Ifeensee's weather forecast and advisory regional meteorological information (e.c., NWS) evacuation time estimates availability of emergency vendor assista ce medical and emergency assistance informatinn
- other plant in'armation.
3 . 2.1. 3 Relationship to r unctional D
- a. A re t'e variables provided suf ficient to al'on the E0r to perform its designated functions?
b.
Are the available 1.97 variables '
adequate to determine regetor coolant system integrity? e.g., :
LOCA high pressure in containment high radiation levels in coolant
- high radiation levels in containment high radiation levels in mainstream Ifne high levels of airborne radioactivity in containment inste It 1810 page 1
R r,v . Il Paga 31 of 42 3.0 ENERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE RECulRENENTS/C P ENTS
- c. Are the available 1.97 var =bles adequatetodetermineprimary contain.e9t integrity? e.g. :
- containment overstressing by high temperature and pressure hydrogen concentration in containment
- pressure changes in contairwent
- high radioactivity levels in c" gas syste s
- high radiation or radioactivity levels in aumiliary buildings, other plant systems, ano site facilities high offsite radiation or radioactivity levels.
- d. Are the available 1.97 variables adequate to determine the operability, capacity, and integrity of the liquid, sclip,andgaseousradwastesystems?
e.g. :
- hydrogen recombiner or offgas piping emplosions
- accidental discharge of untreated aeste 9;r. aciation or radioactivity levels in ausiliary buildings and other site facilities high offsite radiation or radioactivity levels.
- e. Are the available 1.97 variables adequate to determine the extent of damage resulting from a ryfueling or ,
fuel pool accident? e.g. :
loss of water level in fuel pool or vessel
- high rad (st(on of redioscthity 1evels in the fuel handling area, fuel pool, containment, or other auxiliary plant areas, l
- f. Can en evaluation be conducted of !
both the existing and projected status <
of the core / containment and environs to I support determination of proper loote 1: IBID page 1.
l l
l R:v. 11 Pag) 32 of 42 I
3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd) 1 CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS protect *<e action reconnendations utilizing:
(1) the variables provided in Sections 3.2.1.1 and 3.2.1.2 and their functional relationship to determine the status of plant systems (ii) the current containment, process, and area radiation monitor reading (iii) meteorological data and field monitori9g results (iv) weather condit)ons and forecasting (v) deployment of 0.7 site emergency per.unnel (vi) those offaite factors that influence the effectiveness of protective action recorrendations?
3.2.2 Cata 4ccuisitice 10 CrA 50.*7(b)(9); Supplement 1 to NUREC-0737 (E0r recuirement g); GDC 24 and Regulatory Cuides 1.97, Rev. 2 and 1.75.
3 . 2 . 2.1 Data Co11ection Method 1
- a. How are the data acquired? e.g. :
- video technique of gital or analog instruments .
computerized acquisition system voice connunication.
- b. Is the capacity of the data i collection equipment sufficient to access all of the data transmitted to the E0r? l Note 1: 181D page 1 ,
R;v. Il Page 33 of 42 3.0 EMERCENCY OPERATIONS FACILITY (EOF) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 3.2.2.2 Time Resolution -
- a. Is the sampling f requency of each variable adequate to ensure detection of significant changes, particularly eccident conditions?
- b. Is the time resolution for the transmission 08 each of the avai?ab?e variables adequate to assure that no significant data is lost?
3.2.2.3 Isolatien
- a. How is the signal isolation performed for those variables obtained from safety systems adequate to assure that the safety systems will not be degraded by the data acquisition system?
- b. Was the isolation of the installed system verified and validated? If so, how?
3.2.3 Data Conranicatie s 10 CrR 50.47(D )' ?); S.:alement 1 to NUREC-0737 (EOF requirements f and g) 3.2.3.1 Capacity
- a. What is the enannel capacity of the data system?
- b. Is it adequate to meet the needs .
of the data system under peak load and accident conditions?
3.2.3.2 Error Detection
- a. What techniques are used for error detection / correction? .
- b. Does the technique assure error detection from sensor to CPU?
l
- e ~ _ . _ n
R;v. ll Paga 34 of 42 3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS 3.2.3.3 Transmission Between ERes
- a. What methods are used for data transmission?
- b. Is data transmission adequate between the TSC, the Control Room, and the E0r?
3.2.4 Data Analysis 10 CrR 50.47(b)(4),(9); 10 CrR 50, Appendix E ilV.B, ilV.E.2 and ilV.D; Supplement 1 to NUREC-C'37 (E0r requirement g) 3.2.4.1 Reactor Technical Support
- a. Is the data analysis adequate to support the E0r functions?
- b. Will the data analysis 'acilitate Setermination of reactor stayus, past, present and projected? e.g. -
- Forecasting (trending)
- containment pressure vs. time
- co n t ai neer.: temperature vs.
t4me containme9t radiatfor or radicactivity level s vs. time containment H concentration vs. time reacter coolant radioactivity vs. time
- offgas radioactivity levels vs.
ti me radiation or radioactivity ,
levels in various plant systems locations or systems vs.
time.
Precalculated relationships of varigblestoaccidentconditions?
e.g. :
containment radiation levels to core conditions (with and without ab' normal coolant system leakage)
- coolant radioactivity levels to core conditions H 1evel in containment to containment failure Note 1: IBID page 1.
R:v. Il Page 35 of 62 3.0 EMERCENCY OPERATIONS FACILITY (ECr) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS affect on stack monitor readings of gamma radiation field from containment.
- Logistics and Engineering Support?
- availability of equipment and supplies
- availability of personnel
- vendor data a'd specifications offsite data and information availability of offsite support and emergency services (fire fighting, etc.).
- Protective action decisionmaking?
- weathee forecasts and conditfors plant systems status
- dose projections
- evacuation plan.
- c. Are parameters displayed in a manner that makes it easy to determine desigtionsincarametersfromnormal?
e.g.
- ;arameters displayed with superimposed operating curve
- normal ranges of parameter values for operatinc conditions displayed or available paremeters displayed in % of normal or full range values.
- d. Is data analysis performed in a .
manner easily related to EAL criteria (classification and protective action decisionmaking)?
3.2.4.2 Dose Assessment
- a. Does the data analysis capability provide adequate dose assessment?
Itote 1: IBID page 1
Ray, il Page 36 of 42 3.0 EMERCENCY OPERATIONS FACILITY (EOF) (contd)
CRITERIA ACCEPTANCE REQUIRENENTS/CO* ENTS
- b. Can the licensee make timely plume esposure dose projections and formulate adequate protective action recomendations for the following conditions .,.
- variable release durations
- variable distances up to 10 mi.
- variable meterological conditions
- variable and/or multiple source term (s)
- c. Is the licensee capable of continuing dose assessment by evaluation of radiological and meteorological data to determine protective measures as accident and meteorological conditions change?
- d. Based on the variables available and calculational methods used, is trere adequate information evillatle to determine source terms for all potential release pathways? e.g. :
- effluent monitors containment monitors containmeat leak rate
- %e' storage
- post-accident samolln; results
- in-plant radiological monitoring
- offsite radiological monitoring
- inoperable or offseale monitoring instrurents,
- e. Do the source term methods provide for a spectri.pn of radionuclides? I f*
so, what radionuclides and what '
contributions and how are they determined? e.g. :
default values
- simplifying assumptions
- laboratory analysis
- real time measurements,
- f. Are the meteorological variables and calculational methods adequate to characterize the meteorological conditions to about 10 miles f rom the site for release pathways (ground level and elevated releases)?
Isote I n 1810 page 1.
R:v. II Paga 37 of 62 1
3.0 ENERCENCY OPERATIONS FACll.lTY (E0r) (contd)
CRITERIA ACCEPTANCE REQUIRENENTS/CO*ENTS
- g. Whatmeteorylogicalinformationis available? e.g. :
- primary measurement system and/or supplemental measurement system (where appropriate) 'J.
- hi storical/ climatological relationships
- precipitation
- transport and diffusion moceling
- topography
- supplementary data (e.g., NWS)
- default values simplifying assumptions mixing heights and boundary layers.
- h. Have the onsite meteorological monitoring systems historically pro-vided a reliable indication of meteorological variables?
- 1. Can ingestion pathway dose projections be made to about 50 mi from the site to determine the necessity to deploy radiological monitoring systems ia tne injection EPZ? (see 1.2.6. ).
J. Do the dosimetry model calculational meth:ds adequately oetermine thyroid inhalation dose ecruitment and whole body dose for applicable release pathways (ground level and elevated releases) to about 10 miles f rom the plant?
- k. Has the sensitivity and uncertainty inherent in the dose assessment been established and factored into the projections? e.g.
source term
- diffusion and transport
- dosimetry.
Ilote 1: IBID page 1.
RIv. Il Pag) 38 of 42 3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REDUIROMENTS/ COMMENTS 1 Is the field monitor'ng data from all sources incorp3 rated or used to correct and modify the dose projections or assessments? How? If not, how is this data used? e. g. :
number and availability of measurements adequate deployment plan instrumentation cacacilities sampling technique capabilities
- in-situ measurements capabilities Iaboratory analysis of samples under accident conettions use of data from of' site organizations (DOE, EPA, State, local, etc.)
- m. Are provisions for meteorological forecasting adequate te assure that the ef f ects, timing and location of changes arejncludedinthedoseassessments?
e.g. :
site forecasts (approximately 10 miles)
- regio 9al forecests (5:Dre 3,ately 50 miles) short-time (2-12 hr) and long-time (1-3 days) conditions precipitation information maps (aerial coverage) from National Weather Service sources available for forecasting.
O. Determine whether the model used .
by the licensee is consistent with models used by offsite authorities.
3.2.4.3 Central Processor Capability
- a. Is the processing system capacity adequate to support the data acquisition analysis, display, and storage?
leste f r 181D page 1.
- 4 v _ , - e - - e
R;v. 11; Pag 2 *9 d 42 3.0 ENERCENCY OKRAT_ IONS FACILITY (E08) (contd)
CRITERIA ACCEPT M E REQUIREMENTS / COMMENTS e
s 4
- b. Are there other computational requirements on this processing system? -
-o
- c. Do these computational I ;
requirements adversely affect the data i I
acquisition system? .
.4 '
- d. In multiuser or multitask systees ,
hour are priorities for competing tasks
resolved? -
3.2.5 Data Storage Supplement 1 to NUREC-0737)(E0r requirementh) e
- 3. 2.5.1 Storage capabilities
- a. 15 data storage adequate to suppyrtthenecessarydatahandling?
e.g. : ,
I trending
- analytical requirements. 5-
- b. Is the data storage adequate to allow analyti:al review of plant response to t*angierts for augientation persor el ? e. g. -
snort- anc long-ter* cata storage Capaci*y data storage methods and accessibility -
anticipated maximum Quantity of [ ,.
data to be stored under acc* dent ,
8) conditions. ,
j\
3.2.6 Models and Systems Reliability '
and Validity 3 . 2. 6.1 Verification ,
- a. How was the verification done?
e.g. :
- design documentation ';4 implementation tes; records or ;
exercise results correlation of readings with Control Room indicators. 4
, r Note 1: IBID page 1.
l l
4
Rev. Il Page 40 of 42 3.0 EMERCENCY OPERATIONS FACILITY (E0r) (contd)
CRITERIA ACCEPTA4cE REQUIREMENTS / COMMENTS
- b. Was an independent verification performed?~
3.2.6.2 Computer-Based Systems
'f a. How has the reliability of the compyter system been determined?
e.g. 2 u9 availability records
- maintenance logs
= comparisor, with similar systems end-toaerd test.
g b. Does the data acquisition and storage system have an uninterrupt*t?e backup power supply to assure continuity of data acquisition and storage?
3.2.6.3 Panual Systems
- a. What methods are employed to assure that any data that are P aually ,
gathered,crocesseseac/orfisclayedia the Ecr are etiacle? e.g. :
- independent sources of information
crosscheck s -
confirmation between source and destination
- use of formal procedures or checklists.
3.3 rVNCTIONAL CAPABILITIES I
3.3.1 TSC Support Supptement 1 to NUREC-0737 (Ecr recairement a)
Are the E0r operations capable of providing support to the TSC under accident conditions by performing the following tasks based on exercise observations?
i llote 1: IBID page 1.
l l
Roy. 11 Page 41 of 42 3.0 ENERCENCY OPERATIONS rACILITY (E0r) (contd)
CRITERIA ACCEPTANCE REQUtREMENTS/CDMMENTS
- a. Coordinate engineering support such as vendor, plant support personnel, outside cons.itants and
- emperts,
- b. Provide adequate logistic support.
- c. Assist t*e TSC in 3etermining t*e advisat:lity o' mitigative actions that may have offsite impacts? What does this assistance include?
- d. Coordinate radi: logical and other environmental assessments of implemented mitigative actions? Does this include coordination with offsite ageacies?
3.3.2 E0r runctions 10 CrR 50.47(b)(9); 10 CrR 50, Appendix E ilV.E.2, Supplement 1 to NUREC-0737 (E0r requirement a) 3.3.2.1 Dose Assessmeat
- a. Are there adequata cr:cecures o perform c:se assess eat in the Control Room, TSC anc E0r?
- b. Are the procedures, systets and techni:ues used f or dose assass. meat in each facility appropriate?
- c. If different systems are used in each facility, are the results consistent?
- d. Does the dose assessment provide for determining dcse to the whole body and critical crgans for both the plume esposure ar.d ingestion pathways?
- e. Have adequate backup systems and procedures such as manual technique, alternate methods of obtaining data and ~
backup power sources been developed if ADP systems or critical instrumentation becomes unavailable?
Dev. Il Page 42 of 42 3.0 EMERCENCY OPERATIONS FACILITY (EOF) (contd)
CRITERIA ACCEPTANCE REQUIREMENTS / COMMENTS
- f. Is dose assessment properly incorporated into p.'otective action decisionmaking?
- g. Is the methodology used consistent with the techniques and models used by offsite authorities?
- h. Does the Ear's notificatien procedure specify adequate notification in the event of containment venting cr other planned releases?
1 Does the E0r staff have adequate decision aids for planning a ventir; of the containment?
l I
- -w e y -
y +---- w-,y-- r 'y + r' -
- APPENDIX 2 REGULATORY GUIDE 1.97 VARIABLES (Rev. 3)
BWR VARIABLES Tyce Monitored Not Mon *tc*ed Vari abl es A B C D E TSC E0,r But Provided by Neutron riux x Control Rod Position x RCS Soluble Beron x Concentration Coolant Level in Reactor x BWR Core Thermocouples x x RCS Dressure a x Ory e'l *essure x x x Drywell Sump level a Primary Contai* rent Pressure x x Primary Contai'eent x isolation Valve Positon Radioactivity Concentration x or Raciation Level in Cir-culating Primary Coolant Analysis of Primary Coolant x (Comma Spectrum)
Primary Containment Area x x Radiation Drywell Drain Sumps Level n (Identified and Unidentified Leakage) 2.1
~
i I
BWR VARIABLES (Contd)
Ty pe Monitored Not Monitored Variables A B C D E TSC EOF But Provided by Suppression Pool Water Level a x Containment and Drywell x ,
Hydrogen Concentration .*
Containment and Drywell x 0xygen Concentration (ror inerted Contaimme't Plants)
Containment Effluent Radio- x activity--Noble Cases (From identified Release Points)
Radiation Exposure Rate x x (Inside Buildings or Areas)
Main Feedwater flow
- Condensate Storage Tank n Lesel Suppression C5a-ter Spray a clow Suppression Pool Water a Temperature Drywell Atmosphere x Temperature Drywell Spray Flow a .
Main Steamline Isolation x Valves Leakage Control System pressure Primary System Safety Relief x Valve Positions isolation Condenser System x Shell-Side Water Level isolation Condenser System x Valve Position 2.2
BWR VARIABLES (Contd) i Type Monitored Not Monitored Variables A B C D E TSC E0r But Penvided by RCIC Flow x SPCI rion x
.q Core Spray System rion x LPCI System rion x SLCS rion x SLCS Storage Tank Level x RHR System Flow x RHR Heat Exchanger Outlet x Temperature Cooling Water Temperature x to E$r System Components Cooling Water riow to E$r ,
System Components w p Redicactivity Liquid x aa Level Einergency Ventilation Careper x Position Status of Standby P--er and x Other Energy Sources impor-tant to Safety Reactor Building or Second- x ary Containment Area Radia-tion Noble Cases and Vent Flow x Rate Drywell Pur,r, Standby a Cas Treatment System Purge (for Mark I and il Plants) and Second-ary Containment Purge (for Mark lli Plants) 2.3
BWR VARIABLES (Contd)
Type Monitored Not Monitored Variables A B C D E TSC E0r But Provided by
- Secondary Containment x Purge (for Mark I, il ill Plants)
- Secondary Containment x Auxiliary Building x
- Conenon Plant Vent or a Multi Purpose Vent Discharging any cf Above Releases
- All Other Icentified a Release Points Particulates and Hal . gens x
- x All Ide*tified Plant l Release Points Radiation Exposure Meters x (Continuous indication at rimed Locations)
Airborne Radiohalogens x and Particulates (Portable Sampline With Onsite Analy-sis capability)
Plant and Environs Radiation n (Portable instrumentation)
Plant and Environs Radio- x activity (Portable Instru-mentation) wind Direction x Wind Speed x Estimation of Atmospheric x Stab (11ty 2.4 T
y,, , _ . , _ . ___r- , ,m_... , , , _ . , . , _ . _ _ _ 7. . _, - _ _ _ _ _ , , __
i e
l I
i BWR VARIABLES (Contd)
Type Monitored Not Monitored Variables A B C D E TSC EOF But Provided by Primary Coolant and Sump Cross Activity x
.d
- Caama Spectrum x
- Boron Content x Chloride Content x
- Dissolved Hydrogen or x Total Cas Dissolved Oxygen x
- pH x Contairwent Air
- Hydrogen Content x 0xygen Content x Cama Scectrum a O
i
)
i
. 2.5 1
._ _ _ . ~ . _ _ _ _ _ . . . _ _ _ _ _ , _ , . _ . _ . . . _ , . _ . . _ . . _ . . . . _ _ , _ _
PWR VARIABLES Type Monitored Not Monitored Variables A B C D E TSC E0r 8ut Provided by Neutron Flus x Control Rod Position a RCS Soluble Boron Concen- a tration i
l RCS Cold Leg Water T e-cer a- x ture RCS Hot Leg Water Torpera- x l ture RCS Pressure x x Core Exit. Temperature x x Coolant Level in Reactor .*
Degrees of Subcooling x Containment Sump Water Level x x x Containment Pressure a Containment isolation Valve x Position Radioactivity Concentration a or Radiation Level in Cir-culating Primary Coolant Analysis of Primary Coolant x *
(Camma Spectrum)
Containment Area Radiation a x Effluent Radioactivity- x Noble Ces Effluent from Condenser Air Removal Sys-tem Exhaust Containment Hydrogen Concen- a tration i
l 2.6
PWR VARIABLES (contd)
Type Monitored Not Monitored Variables A B C D E TSC E0r But Provided by Containment Effluent Radio- x activity-Noble Cases from Identified Release Points Radiation Exposure Rate x x (Inside Buildings or Areas)
E' fluent Radioact i vity x Noble Cases RHR System Flow x R4 Heat Exchanger Outlet x Temperature Accumulator Tank Level x and Pressure .
Accu *ulator Isolation Valve x Position Beric Acid Charging riow x fl e in NPI System a clow in LPI System x Refueling Water Storage x
- ank Level Reactor Coolant Pump a Status Primary System Safety Relief x Valve Positions (including PORV and Code Valves) or riew inrougn or Pressure in Relief Valve Lines P*essurizer Level x Pressuizer Heater Status x i
Quench Tank Level x 2.7
PWR VARIABLES (Contd)
Type Monitored Not Monitored Variables A B C D E TSC E0r But Previded tv Quench Tank Temperature x Quench Tank Pressure x Steam Cenerator Level x S te "9 Cenerator P* essure x Safety / Relief Velve Post- x tions or Main Steam Flow Main Feedwater Flow x Auxiliary or E.ergency a reedwater Flow Condensate Storage Tank x Water Level Containment Spray Fl ow x Heat Removal by the Con- x tair ent ran eat Removal O
ystem Containment Atmospr ealc n Temperature Containment Sura Water x Teeperature Makeup Flow-in a Letdown Flow-out x ,
Volume Control Tank Level x Component Cooling water a Temperature to ESF System Camponent Cooling Water a Flow to ESF System High-Level Radioactive x
- Liquid Tank Level 2.8
PWR VARIABLES (contd)
Type Monitored Not Monitored Variables A B C D E TSC EOF But Provided by Radioactive Cas Holdup x Tank Pressure Emergency Ventilation x Damper Position Status of Standby Power x and Other Energy Sour:es important to Safety Neble Cases and Vent x Flow Rate Containment or Purge a Effluent
- Reactor Shield Butid- ,
a ing Annulus Auxiliary Building x
- Condenser Air Removal x System Exhaust
- Common 8'n't Ve t er a Multipurpose Vent Dis-charging Any of Above Releases Vent from Steam "enera- a tor Safety Relief Valves or Atmosoneric Dump Valves ,
All Other identified a Release Points Particulates and Halogens x All Identified Plant Release Points Radiation Exposure Meters x (Continuous Indication at -
rimed Locations)
+
2.9 l
l
PWR VARIABLES (Contd)
Type Monitored Not Monitored Variables A B C D E TSC E0r But Provided bv Airborne Radichalogens and a Partic.* ates (Portable Sampling with Onsite Analy-sis Capability)
Plant and Environs Radia- a tion (Portable Instrumenta-tien)
Plset and Environs Radio- x activity (Portable instru-mentation)
Wind Direction a Wind Speed x Estimation of Atmospheefe a Stability Primary Coolant and Sump
- cross Acthity u Ca ma Scect v i
- Boron Content :
Chloride CcF! eat Dissolved Hydrogen or a Total Ces
- Dissolved Oxygen x
= pH Containment Air
- Hydrogen Content a
)
- Oxygen Content a 1
- Carea Spectra a A = Plant spect fic 8 = Safety function status C = Ffssion production barrier status D = Safety system operation E = Release monitoring 2.10 '
APPENDIX 3 RECOMMENDED APPRAISAL ASSIGNMENT MATRIX Appraisal Items Technical Area Assigned C
w s . 1 3 Y .5 2 E C % %
E E $
5 & % $
3 4 u 2 -
E e 4
- 3 2 e U E o E 8 S 2
m e 8 8 o
z v 8 %
z =
F
. 1.0 TSC 1.1 PEysical Facilities 1.1.1 Design 1.1.1.1 Size and Layout e 1.1.1.2 Location ! ; e 1.1.1.3 Structure l l e 1.1.1.4 Wabi tabili ty/ Envi ronment i { ei ( t X 1.1.1.5 Display Interfaces -
l + e i
. + ,
1.1.2 Radiclogical Ecuipment and Succiies 1.1.2.1 Radiation Monitoring e X l.1.2.2 Protective Supplies e 1.1.3 Non-Radiological Equipment and Supplies 1.1.3.1 Records / Drawings e X ,
1.1.3.2 Support Supplies X X X X . X e l 1.1.3.3 Power Supply e X X 1.2 Infonnation Management 1.2.1 Variables Provided 1.2.1.1 Regulatory Guide 1.97 Rev. 2 or 3 Variables X e 1.2.1.2 Other Variables X e 1.2.1.3 Relationship to Functional Needs e X X X e - responsible for item write-up, where more than one is indicated a separate write-up is expected from each.
3.1
Appraisal Items Technical Area Assigned E
m 2 2 $
e a e : s E E 3 & D A 3 4 c 3 -
% e & "
3 e a U E o E 8 S
=
2 8 m S r
8 u
r =
F 1.2.2 Data Acquisition 1.2.2.1 Data Collection Method e X 1.2.2.2 Time Resolution e X 1.2.2.3 1 solation e X 1.2.3 Data Communications 1.2.3.1 Capacity e 1.2.3.2 Error Detection X e 1.2.3.3 Transmission Between ERFs e 1.2.4 Data Ana. lysis 1.2.4.1 Reactor Technical Support e X X X Dose Asse?; ment e
. 2.4.2 e ! t t e
. 2.4.3 Central F ;cessor Cacability '
l e f i 1.2.5 Data Storage 1.2.5.1 Storage capabilities X X X X e X 1.2.6 System Reliability 1.2.6.1 Verification X X X X e X 1.2.6.2 Computer Based Systems X e 1.2.6.3 Manual Systems X X e . X 1.2.7 On-Shift Dose Assessment 1.2.7.1 Dose Assessment Proficiency e o X X e 1.2.7.2 Dose Assessment Technical Adequacy e e X X e 1.3 Functional Capabilities 1.3.1 Control Room Support 1.3.1.1 Technical Support e X 1.3.2 Initial EOF Functions (See S'ection 3 ,3)
- - responsible for item write-up, where more than one is indica ted a separate write-up is expected from each.
3.2
Appraisal Items Technical Area Assigned C
m E e k=
s ; e 2 3 0 0 5 E
E 3 # D
. t 2 .-
E . e =
3 2 ;
t E . a a 3 3
= m 8
o 8 l
=
5 v
E =F 2.0 OSC 2.1 Physical Facilities 2.1.1 Design 2.1.1.1 Location e 2.1.1.2 Alternate 05C Location (s) e 2.1.1.3 Size, Layout and Environment e X 2.1.1.4 Display Interface o X 2.1.2 Radiological Equipment and Supplies 2.1.2.1 Raciation Monitoring e X E l.2.2 Personnel Cosimeters e x 2.1.2.3 Frotective Supplies !
- 2.1.3 Mn-Radiological Eaufpment and Supplies 2.1.3.1 Support Supplies e X X 3.0 EOF 3.1 Physical Facilities .
i 3.1.1 Design l 3.1.1.1 Size and Layout e !
3.1.1.2 Location e 3.1.1.3 5tructure e
=
3.1.1.4 Habitability / Environment e X X 3.1.1.5 Display Interface e 3.1.2 Radiological Equipment and Supplies 3.1.2.1 Radiation Monitoring 'e X 3.1.2.2 Protective Supplies e l e - responsible for item write-up, where morit than one is indica Led a separate write-up is expected from each.
3.3
Appraisal Items Technical Area Assigned r
m E b a $ b e e : s a E 3 & s *
% c S -
8 . e' "
=
3 EI:
t E . a 8 I
S 2
=
8 m
8 ci l
z v 8
z
% m F
3.1.3 Non-Radiological Equipment and Supplies 3.1.3.1 Records / Drawings e X 3.1.3.2 Support Supplies X X X X X e 3.2 Infonnation Management Systems 3.2.1 Variables Provided 3.2.1.1 Regulatory Guide 1.97 Rev. 2 or 3 Variables X e 3.2.1.2 Other Variables X e 3.2.1.3 Relationship to Functional Needs e X X X 3.2.2 Data Acquisition 3.2.2.1 Data Collection Methods e X 3.2.2.2 Time Resolution e X 3.2.2.3 Isalation e X 3.2.3 Data Communications i 3.2.3.1 Capacity e 3.2.3.2 Error Detection X e 3.2.3.3 Transmission Between ERFs e 3.2.4 Data Analysis 3.2.4.1 Reactor Technical Support e X X X X 3.2.4.2 Dose Assessment e o X e 3.2.4.3 Central Processor Capability e 3.2.5 Data Storage ,
3.2.5.1 Storage Capabilities X X X X e X ,
l 3.2.6 System Reliability {
3.2.6.1 Verification X X X X e X 3.2.6.2 Computer Based Systems X e 3.2.6.3 Manual Systems X X e X e - responsible for item write-up, where more thiin one is indica ;ed it separate write-up is expected from each.
3.4
Appraisal Items Technical Area Assigned l
U ;
j 8 .
- Y E 2 3 0 t 5 E E 3 A &
, M r 2 -
5 . 4 =
3 2 2 t E . a 8 3 ac m a 8 o
z v 8 %
x F
ec 3.3 Functional Capabilities 3.3.1 TSC Support e X X X X 3.3.2 EOF Functions 3.3.3.1 Dose Assessment e e X e e - responsible for item write-up, where more than one is indicated a separate write-up is expected from each.
9 9
i J .
I i 3.5
APPENDIX 3a APPRAISAL ASSIGNMENT MATRIX ~
Appraisal Items Technical Area Assigned E
E m k-
- ; .e 2 E C O %
i E :
S A & 3 3 W u -
8 . # '
3 2 :
U E e E E S
=
8 m
E e
I z
5 v
z =
?
1.0 TSC 1.1 T6ysical Facilities 1.1.1 Desian
~
1.1.1.1 Size and layout 1.1.1.2 Location 1.1.1.3 Structure 1.1.1.4 Habitability / Environment 1.1.1.5 Display Interfaces 1.1.2 Padiological Equipment and Supplies .
1.1.2.1 Radiation Monitoring 1.1.2.2 Protective Supplies 1.1.3 Non-Radiological Equipment and Supplies 1 1.1.3.1 Records / Drawings 1.1.3.2 Support Supplies -
1.1.3.3 Power Supply l l
1.2 Information Management 1.2.1 Variables Provided 1.2.1.1 Regulatory Guide 1.97 Rev. 2 Variables 1.2.1.2 Other Variables 1.2.1.3 Relationship to Functional Needs l 3a.1
Appraisal Items Technical Area Assigned C
- 2 e E' E e a r -
s 1 E 3 & D 3 s S -
e o 4 "
3 E 2 t E o a 8 3
=
a m ci 8 I z o 8
- E l =?
1.2.2 Data Acquisition 1.2.2.1 Data Collection Method 1.2.2.2 Time Resolution 1.2.2.3 Isolation 1.2.3 Data Communications 1.2.3.1 Capacity 1.2.3.2 Error Detection 1.2.3.3 Transmission Between ERFs 1.2.4 Data Analysis 1.2.4.1 Reactor Technical Suppoi t l 1.2.4.2 Dose Assessment i i i 1.2.4.3 Central Processor Capability t i
' l 1.2.5 Data Storage 1.2.5.1 Storage capabilities 1.2.6 System Reliability
- 1.2.6.1 Verification 1.2.6.2 Computer Based Systems 1.2.6.3 Manual Systems -
1.2.7 On-Shift Dose Assessment 1.2.7.1 Dose Assessment Proficiency 1.2.7.2 Dose Assessment Technical Adequacy 1.3 Functional Capabilities 1.3.1 Control Room Support 1.3.1.1 Technical Support 1.3.2 Initial EOF Functions (See Section3 3)
~
k 3a.2
l
)
I Appraisal Items Technical Area Assigned E
- 2 5 : i I e e : s*
& E A ?
. 8 u -
5 o E
- 2 2 U E o E E .2 2
a .o 8 e o x 8
v z a F
2.0 OSC 2.1 Physical Facilities 2.1.1 Design 2.1.1.1 Location 2.1.1.2 Alternate 05C Location (s) 2.1.1.3 Size Layout and Environment 2.1.1.4 Display Interface 2.1.2 Radiological Equipment and Supplies 2.1.2.1 Radiation Monitorirg l D .2.2 ;ersonnel Cosimeters ! i 2.1.2.3 Protective Sucolies i , i i
2.1.3 Non-Radiological Equipment and Supplies 2.1.3.1 Support Sucolies 3.0 EOF 3.1 Physical Facilities
- 3.1.1 Desig 3.1.1.1 3 Tie and Layout 3.1.1.2 Location 3.1.1.3 Structure 3.1.1.4 Habitability / Environment i 3.1.1.5 Display Interface 3.1.2 Radiological Equipment and Supplies 3.1.2.1 Radiation Monitoring 3.1.2.2 Protective Supplies 3a.3
l l
Appraisal Items Technical Area Assigned l
1 E
, . 2 E b
~
e a e s
& E S & a 3
, % .c -
8
. e 3 Eia E . E E E
~
.5 3
a m a e c
5 z
8 v f a
?
3.1.3 Non-Radiological Eauipment and Supplies 3.1.3.1 Records / Drawings 3.1.3.2 Support Supplies 3.2 Information Management Systems 3.2.1 Variables Provided 3.2.1.1 Regulatory Guide 1.97 Rev. 2 Variables J 3.2.1.2 Other Variables i i Relationship to Functional i 3.2.1.3 l
. Needs ! ! 6 f
3.2.2 Data Acquisition 3.2.2.1 Data Collection Methods 3.2.2.2 Time Resolution l 3.2.2.3 Isolation l ;
i ,
l 3.2.3 Data Communications 3.2.3.1 Capacity 3.2.3.2 Error Detection ,
3.2.3.3 Transmission Between ERFs 3.2.4 Data Analysis 3.2.4.1 Reactor Technical Support 3.2.4.2 Dose Assessment 3.2.4.3 Central Processor Capability 3.2.5 Data Storace 3.2.5.1 Storage capabilities '
3.2.6 System Reliability 3.2.6.1 Verification 3.2.6.2 Computer Based Systems 3.2.6.3 Manual Systems 3a.4
)
Appraisal Items Technical Area Assigned 1
0 m 2 e $
e a e s
& E 3 & D A 3 5 %
u 2 -
e v & "
B 8 2 t d . 5 3
a m 8 E o
I z
k v
~2 E =
E 3.3 Functional Capabilities 3.3.1 Technical Support t 3.3.2 EOF Functions 3.3.3.1 Dose Assessment a
4 9
f I
i 3a.5
APPENDIX 4 Appraisal items Personal Contacts Reference /Conments i t em Organization Individual (s) Phone No.
10 ist 1.1 Physical Facilities 1.1.1 Design 1.1.1.1 Site and Layout 1.1.1.2 Location 1.1.1.3 Stru:ture 1.1.1.4 Habitability / Environment 1.1.1.5 Display interfaces 1.1.2 Radiological Equipment and Supplies 1.1. 2 .1 Radiation Monitoring 1.1.2.2 Protective Supplies 1.1.3 Non-Radiological Eculpment and Supplies 1.1.3.1 Records /Crsaings 1.1.3.2 Support Supp11es 1.1.3.3 Power Supply
'2. Information w= a;eaeat
- 1. 2.1 Variables o re. i dec
- 1. 2 .1.1 Regulatory Guide 1.97 Rev. 2 Variables
- 1. 2 .1. 2 Other Variables 1.2.1.3 Relationship to functional Needs 1.2.2 Data Acquisition
- 1. 2 . 2.1 Data Collection Method
- 1.2.2.2 Time Resolution 1.2.2.3 Isolation 1.2.3 Data Communications
- 1. 2. 3 .1 Capacity 1.2.3.2 Error Detection 1.2.3.3 Transmission Between Errs 1
1
- 1. 2.4 Data Analysis
- 1. 2. 4.1 Reactor Technical Support
- 1. 2. 4. 2 Dose Assessment
- 1. 2.4.3 Central Processor Capability 1
4.1 1 1
I
]
Appraisal items Personal Contacts Re f erence/ Comments Item Organfration Individuel(s) Phone No.
1.2.5 Data Storage
- 1. 2 .5.1 Storage Capabilities 1.2.6 System Re'iability
- 1. 2 . 6.1 Verification 1.2.6.2 Computer Based Systems 1.2.6.3 Manual Systems 1.2.7 Cn-Shift Dese Assessment
- 1. 2.7.1 Dose Assessment Proficiency 1.2.7.2 Dose Assessment Technical Adequacy 1.3 runctional Capabilities 1.3.1 Control Room Support
- 1. 3 .1.1 Technical Support 1.3.2 Initial E0r runctions 2.0 25$
2.1 Physical facilities
- 1.1
. Desiga 2.1.1.5 Locat'en 2.1.1.2 Alternate OSC Location (s) 2.1.1.3 Size, Layout and Environment 2.1.1.6 Display Interface 2.1.2 Radiological Equipment and Supplies 2.1.2.1 Radiation Monitoring 2.1.2.2 Personnel Dosimeters .
2.1.2.3 Protective Supplies 2.1.3 Non Radiological Equipment and Supplies 2.1.3.1 Support Suoplies 3.0 _E_g
. r ,
l 3.1 Physical racilities I 1
3.1.1 Design 3.1.1.1 Site and Layout 3.1.1.2 Location 3.1.1.3 Structure 4.2 l \
1
1 Appraisal items Personal Contacts Re ference/Commeats item Organfration Individualfs) Phone No.
3.1.1.4 Habitabf11ty/ Environment 3.1.1.5 Display Interface
] 3.1.2 Radioloateal Equipment and Supplies 3 .1. 2 .1 Radiation Monitoring j 3.1.2.2 Protective Supplies 3.1.3 Non-4adiologient Equipment and Saoplies 3.1.3.1 Recorcs/ Drawings 3.1.3.2 Support Supplies 3.2 Information Management 3.2.1 Varisoies Provided 3.2.1.1 Regulatory Cuide 1.97 Rev. 2 Variables 3.2.1.2 Other Variables 3.2.1.3 Pelationship te runctional Needs 3.2.2 Data sequisition 1.2.2.1 Data Collection Methods 3.2.2.2 Time Resolution 3.2.2.3 I sol a ti on 3.2.3 Data Communications 3 . 2 . 3 .1 Capacity 3.2.3.2 Error Detection 3.2.3.3 Transmission Between ERFs
) 3.2.4 Data Analysis 3 . 2 . 4 .1 Reactor Technical Support .
3.2.4.2 Dose Assessment I 3.2.4.3 Central Processor Capability 3.2.5 Data Storage 3 . 2 . 5.1 Storage Capabilities 3.2.6 System Reliability a
3 . 2.6.1 Verification 3.2.6.2 Computer Based Systems
! 3.2.6.3 Manual Systems 3.3 runctional capabf14 ties 3.3.1 TSC Support 3.3.2 E0F runettons
- 3. 3. 2.1 Dose Assessment 4.3 l l
6 4
i t APPENDIX 5 Documentation needed to conduct the'ERF appraisal, e,
Documentation.for all team members:
j e EmergencyPINn'
- EPIPs 4
e FSAR e . Description and location of alternate ERFs t e Listing of types and quantities of equipment maintained .in ERFs
! - protective clothing 1 - dosimeters J -
survey instruments SCBAs '
i - procedures reference material <
) Dose Assessment Documentation: -
- Implementing procedures for both computerized and manual dose assessment, e User's guide for computerized dose assessment mcdel.
j e Technical basis document for dose assessment model.
}-
- Documentation of any comparative studies done between the licensee's model and the state model(s).
- Documentation of any verification studies done on the licensee's DA 4 program, i
! e Maps of the area (10 and 50 mi radius).
i l Computer Systems Documentation:
! e Computer configuration specification for Emergency Data Acquisition
! System, Plant Computer, and SPDS l e Description of data system operation (i.e[i " user's guides")
) e Records of system availability ,
e Documentation of computer tode verificatiof
- e Examples of hard copy output for routine reports,and graphical displays
- e Block diagram of computer systems showing interfaces.
4 5.1 1 ,
-_,._.-,,..._,._,,_m_- - --
- c. ...,_-,,,,._..__..,_..,,.,.,-._,,,-4,.v._ ..,m_-,% n,.._.-,%,,,__ , , , , , ,,y, _ , , , , , , _.m__,,...eur'
Peactor Operations Documentation:
e Electrical one line diagramt from off-site to the TSC, normal power, emergency power, lighting, phones, comunication systems, station PBX, micro-wave, plant process computer, data acquisition systems Same for EOF if near-site; if far-site, power feeds to the building.
- EPIPs covering classification, core-damage assessment, TSC Manager responsibility and EOF Manager responsibilities.
- Integrated, living schedule for all ERF related items, R.G.1.97 items e R. G.1.97 submittal, EG&G review, final SER e SAR by licensee on its Data Acquisition System and SPDS e Plant Information Manual on Plant Process Computer, SPDS, Radiation Monitoring System, Electrical Distribution e Inventory of TSC and E0F documents and references.
Meteorological Documentation:
- A block diagram of the meteorological system showing the path data takes from sensor to storage and display, identifying the main components in the system e.g., sensors, signal conditioning, data acouisition systems, data processing, data storage, and data displays and their locatiors, e Technical specifications for system sensors and other system components, and a list of their special features, such as heaters for wind instruments.
- A detailed description of the tower and sensor mounts, and a plan-view drawing, preferably to scale, e Description of power sources for the sensors, signal conditioning, data acquisition systems and recorders including power conditioning, lightning protection and backup sources of power.
- Environmental controls for areas in which signal conditioning, data acquisition systems, recorders and other critical system components are located.
- All written procedures for meteorological system operations, maintenance and calibration.
- Documentation on meteorological data availability.
- A copy of the most recent joint frequency distribution of wind direction, wind speed and atmospheric stability.
5.2
e A list of the locations where onsite meteorological data would be available during an emergency.
- A list of sources of regional meteorological data and forecasts noting fomal agreements and contracts.
- Written procedures related to dose assessment and activation of the ERFs.
- A generic description of the methods used to evaluate transport, diffusion, deposition and other atmospheric processes in all ERFs.
- Listings of computer codes used in dose assessment, e Supporting documentation for atmospheric models including those in the dose assessment codes, e.g., theoretical bases, code verifications, 4 user's guides.
- Maps of the area (10 and 50 mi radius).
Human Factors Documentation:
e TSC, OSC and EOF floor plans and activation procedures e Documentation for the Emergency Data Acquisition System (EDAS) and other interactive computer systems e Sample screen fomats from EDAS and ether interactive corouter systems e Functional block diagram of EDAS, plant computer system and emergency response communication systems.
- Sample copies of any reports or foms that are circulated to emergency response decisionmakers. -
l Source Tem Documentation:
e One line drawings of plant's ventilation system showing the following: l
- vent flow rstes
- points monitored and description of monitors
- fan and damper line-ups for nomal and acci<ient modes e Any studies / evaluation made of potential unmonitored release paths.
- Effluent monitor calibration procedures and calibration data.
Description of methods used to verify manufacturers primary calibration.
- Core damage estimate procedures.
- Description of plant radiation monitoring systems (process monitors, ARMS, and CAMS). One line drawing for these systems and a list of monitors powered from vital power.
5.3
( 1 1
- Description of the plants post accident monitoring system and its capabilities, e Listing of and rational for nuclide library used by dose assessment procedures or computer programs, e A description of the basic source term assumptions used for accident scenarios treated by manual and computerized dose assessment methods, and the rationale behind each.
- R. G. 1.97 submittal, EG&G review, final SER e Written procedures related to dose assessment.
e Maps of the area (10 and 50 mi radius).
I 5.4
ENCLOSURE 3 Checklist For Determination of Completed ERFs
- 1. Physical Facilities
- a. Structures completed and operational
- b. Ventilation systems installed and cperatfor;al
- c. All furniture cnd hardware in place.
- d. All instrumentation and communications equipment installed and operational
- e. All radiation and meteorological monitoring equipment and other ,sequipment installed and operational.
- 2. Data Acquisition Systems
- a. All hardware, firmware and software designed, installed and operational,
- b. All detectors and indicators with their associated displays and read-cuts designed and installed.
- c. All infonnation and data displays and calculational models designed, installed and operational.
- d. Verification procedures for all data system displays and rrodels com-plete and docu ented.
- e. SPDS displays de;igned and installed if part of the data a.:visiti:n system.
- 3. All necessary procedures completed in final form for operation - of all facilities, instrumentation, equipment and functions.
- 4. Personnel trained to carry out ERF functions and operation of all data systems, communications, instrumentation and equipment.
- 5. All plant records, drawing and other infomation essentia'l for determining plant accident status available to ERFs.