ML20195G603
| ML20195G603 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/15/1987 |
| From: | Dan Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | James O'Reilly GEORGIA POWER CO. |
| References | |
| TAC-45950, TAC-45951, TAC-45952, TAC-46021, TAC-46022, TAC-46093, TAC-46094, TAC-46309, TAC-46310, NUDOCS 8710260205 | |
| Download: ML20195G603 (48) | |
Text
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nr.; p; Georgia Power Company ATTN: Mr. James P. O'Reilly E
Senior Vice President-
$c Nuclear Operations O c3if P. O. Box 4545 e
Atlanta, GA 30302
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Gentlemen:
,?
SUBJECT:
EMERGENCY RESPONSE FACILITY (ERF) EVALUATION y
Enclosed for your information and future discussion is a copy of the latest draft of the NRC inspection procedure for performing evaluations of licensee's upgraded ERFs.
As Mr. T. De:ker has recently discussed with your staff, we would like to schedule such an evaluation at your Hatch facility in the near future.
You should note that this evaluation would be in lieu of an observation of an emergency exercise.
In addition, the 5, cope of the inspection j
has been refined to focus on assessment of releases, data availability, facility function and habitability.
We will be discussing these issues with you in the very near future.
Sincerely, Douglas M. Collins, Chief Emergency Preparedness and Radiological Protection Branch Division of Radiat b-tafatv and Safeguards
Enclosure:
Emergency Response Facility Evaluation cc w/ encl
- J. T. Beckham, Vice President Plant Hatch H. C. Nix, Site Operations General r
N nager A. Fraser, Acting Site QA Supervisor L. Gucwa, Manager, Nuclear Safety and Licensing bec w/ encl:
NRC Resident Inspector Hugh S. Jordan Executive Secretary Document Centrol Desk State of Georgia l
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DRAFT REVISION A INSPECTION PROCEDURE 82412 EMERGENCY RESPONSE FACILITIES EVALUATION PROGR8M APPLICABILITY 2515 This inspection procedure is applicable to evaluation of emergency response r
facilities required for licensed nuclear power plants.
82412-01 INSPECTION OBJECT!YE To determine if the Emergency Response Facilities (ERFs) at licensed nuclear power plants meet the requirements of 10 CFR 50.47(b), Appendix E. Paragraph IV.E. 8 of 10 CFR Part 50 and the orders and license conditions issued to implement Supplement 1 to NUREG-0737 Requirements: 6.1 b (2nd paragraph), 6.1
- c. 6.1 d, 8.2.1 a, 8.2.1 f, 8.2.1 h, 8.4.1 a, 8.4.1 b and 8.4.1 g.
82412-02 INSPECTION REQUIREMENTS 02.01 General. Perfonn an onsite inspection of the licensee's ERFs, including the data and infor: nation systems and equipment in the Technical Support Center (TSC) and the Emergency Operations Facility (EOF) to determine if these facilities provide adequate and reliable support to the prie ipal emergency managers during radiological accidents. This inspection shall be conducted during the licensee's annual emergency preparedness exercise by a special team of NRC and contractor personnot.
02.02 Assessment of Radioactive Releases.
Evaluate whether the ERFs are adequately equipped to detennine the magnitude of and for continuously assessing the impact of a release of radioactive material to the environment.
5 02.03 Meteorology. Detennine if the meteorological measurements provide a reliable indication of the meteorological variables ( wind direction, wind i
speed and atmospheric stability) specified in RG 1.97 (Rev. 2) for site i
1
!ssue Date:
10/01/87
DRAFT REVISION A meteorology.
Evaluate whether the data system and any appropriate modeling provide a reliable indication of these variables that are representalive of meteorological conditions in the vicinity (up to about 10 miles) of the plant site. Determine if information on meteorological conditions for the region in which the site is located are available via comunications with the National Weather Service or equivalent meteorological service organization.
02.04 TSC Data Availability. Detennine if the RG 1.97 (Rev. 2 or 3) Type A, B, C, D and E variables that are essential for the TSC managers to perfonn their functions are available in the TSC.
Principally those data must be available that would enable he TSC managers to evaluate incident sequence, detemine mitigating actions, evaluate damage, estimate actual and potential radioactive releases and detemine plant status as well as the meteorological data and systems as described in 02.03 above.
02.05 TSC Functions. During periods of activation, determine if the TSC will operate uninterrupted to provide TSC and plant managers with the capability to technically support plant operations personnel and relieve them of peripheral duties and comunications not directly related to reactor systems manipulations.
Determine whether the TSC is equipped to provide the TSC managers with the capability to perfom the EOF functions during Alert, Site Area Emergency and General Emergency classifications until the EOF becomes functional.
02.06 TSC Habitability. Detemine whether the TSC is equipped to assure that the radiation exposure to any person working in it would not exceed 5 rem whole-body dose, or its equivalent to any part of the body, for the duration of an accident.
02.07 TSC Data Systems. Detemine whether the data systems in the TSC will provide the TSC managers with reliable data collection, storage, analysis, display and comunications sufficient to detemine plant site and regional status and forecast status to take appropriate actions.
02.08 EOF Functions. When the EOF is activated, detemine if it is equipped to provide N EOF managers with the capability for management of the overall 2
1ssue Date:
10/01/87
DRAFT REY!$ ION A 0
liceasee emergency response, coordination of radiological and environmental assessments, dtvelopmnt of recospendations for public protective actions and coordination.f emergency resoonse activities with Federal, Stete and local agencies.
02 3 EOF Data Availability. Determine if the primary indicators needed for the EOF managers to monitor containment conditions and releases of radioactiv-ity from the plant are available in the EOF, Acquisition, display and evalua-tion of the radiological data, meteorological infomation (including the data and systems described in 02.03 above) and containment condition parameters must be adequate to evaluate the magnitude and effects of actual or potential radioactive releases from the plant and to detri. wine projected dose ensite and offsite. Detemine if these data are adequate for the EOF managers to make proper protective action deterzinations and recomendations.
02.10 EOF Locations And Habitability. De. ermine if the EOF location and habitability meet the requirements of Table 1 of Supplement 1 to NUREG-0737.
02.11 EOF Data Systems. Determine whether the data systei.s in the EOF will provide the EOF managers with reliable data collection, storage, analysis, display and comunications sdficient to determine plant site and regional status and forecast status to take appropriate actions.
82412-03 INSPECTION GUIDANCE 03.01 Inspection procedure. The inspection shall be conducted at the licensee's plant' site after the final physical facilities, data acquisition and other equipment systems, software programs, and procedures for the ERFs have been developed and installed. The inspection procedures and techniques to be used are aa follows:
a.
The inspection shall be conducted using a team consisting of the following individuals:
l 3
Issue Date:
10/01/87
{
i onAFT REVISION A i
1 1.
Regional Team Leader 2.
Reactor Systems Engineer 3.
Neteorologist 4.
Dose Assessment Specialist 5.
Computer Systems Specialist (only when a computerized data acquisitioni=usedintheERFS).
b.
The inspection shall be conducted during the licensee's annual exercise using this procedure rather than Inspection Procedere 82301. The usual observation of the licensee's activities will not be perfonned under this procedure. The NRC Regions slay determine that the licensee's exercise must be observed using Inspection Procedure 82301 if special circumscances justify its observation (e.g., significant deficiencies or open items from previous exercise).
In this case the ERF evaluation may M deferred until the riext annual exercise or perfonned separately uring scheduled drills involving the ERFs.
If the exercise is a full participation exercise to be conducted in conjunction with offsite authorities, the Federal Emergency Management Agency should be advised that an NRC critique of its exerc.ise observations will not be provided.
The exercise must be scheduled to take place between Monday afternoon and Wednesday evening to ensure that the inspection team has adequate time to gain entrance to the site, observe the annual exercise and evaluate the capability of the licensee's ERFs to support the emergency managers and prepare a sumary of findings to be used by the Regional l
team leader during his exit meeting with the licensee.
It is anticipated that this inspection can be conducted during a four day j
period onsite.
During the inspection the team will evaluate the licensee's ERFs by c
.bserving the functioning of the ERFs during the exercise, by reviewing ERF systems and by interviewing key personnel.
The following areas will be reviewed during the inspection; the hardware and sof tware design of the emergency data acquisition system, the models and techniques use to detennine the source tenn, transport, and dispersion of radioactive 4
Issue Date:
10/01/87 t
i I*
DRAFT REVISION A i.
materials releases to the environmen'c. The inspectors will also interview the engineering and design personnel that developed the systes, procedures and techniques. During the exercise the inspectors will observe the capabilities of these facilities and their supporting data and equipment systems to meet the needs of emergency managers. The licensee should be requested ;o operate data acquisition systems, run computer models, demonstrate software designs and operate emergency ventilation and lighting equipment for the inspection team to verify compliance with the requirements and conunitments for these systems.
d.
A matrix reconsnending the inspection assignments of each team member is provided in Appendix 1 of this procedure. A blank assignment matrix is also provided for use by the Regional team leader.
Assignments for the various team members provide specific areas to be inspected on an inde-pendent basis. Although the procedure provides guidance for conducting the evaluation, reasonable flexibility will be allowed each inember to account for the plant specific character of the ERF design. At the discretion of the team leader, an indepth review greater than defined by the scope of the guidance may be pursued for areas whe re weaknesses are suspected.
Each onsite inspection 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, hardware design, tnd emergency pro-cedures for the ERFs should be perfortned.
Upon completion of the inspection of the final ERFs, a fonnal inspection e.
report will be written by the NRC Region. This report will be developed from the individual written inputs from the team members assigned to the various areas to be evaluated. Discussions and coordination of the report findings may necessitate the team leader conferring with the team.Tembers.
The team members are responsible for providing a written evaluation and findings for each inspection item assigned.
It should be noted that some of th>Se items or inspection areas are assigned to more than one 5
!ssue Date:
10/01/87
DRAFT REVISION A team member. However, the team men 6er responsible for preparing the evaluation is designated in the assignment matrix with the other team members providing supporting inputs on an observed or requested basis depending on the team leader's judgement and the needs of the team member responsible for evaluating the item for the report. Should any support-ing or other team member observe any potential problem area (s) warrant-ing further evaluation by the team member responsible for preparing the applicable pcrtion of the report, the item should be discussed with the responsible team member. Should team members responsible for preparing specific sections of the report find they may not be able to complete all assigned sections, the team leader will be alerted, f.
Team members will coordinate their activities to minimize the need for the licensee to operate or demonstrate the same equipment or process more than once (e.g., demonstrations of data acquisition systems, and dose assessment systems, and discussions of complex programs or documents.) See item 03.01.d. above.
9 The inspection findings shail specify if there is reasonable assurance that the licensee's ERFs, including the data and infonnation systems and i
equipment provide adequate and reliable support to the principal emergency managers during radiological accidents.
Identified violations of requirements must be 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 perform intended functions.
Deviations must be referenced to specific ccanitments by the licensee in the FSAR or other de:umentation. Open items shall include incomplete systems or areas where the licensee agrees to make $=nges prior to the issuance of the inspection report. Although the ERF Evaluation Report may recomend improvements in the ERFs to enhance their operational capabilities, only violations, deviations, and open items shall be included in the report i
findings.
Deviations, and open items will be handled in accordance with regional policy and violations will be handled under nonnal inspection and enforcement procedures.
6
!ssue Date:
10/01/87
DRAFT REVIS!0N A h.
The following schedule should be adhered to in initiating the ERF Evaluation Inspection:
1.
The team leader should provide the following to the licensee
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approximately six weeks prior to the scheduled inspection:
(a) Appendix 2 of this procedure which provides a 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 to be evaluated. The licensee should assure the availai,ility of these individuals during the last three days of the ch:ite inspection in orde-for the team to complete its evaluatie within the allotted inspection period.
(b) Appendix 3 of this procedure which provides a list of various documents and other information that are needed to conduct the inspection and should be provided to the team when it arrives onsite.
2.
At least 15 working days prior to the projected onsite inspection, the team leader will contact plant management and the Resident Inspector to arrange for team access and workspace. This will te confirmed in writing by the Region, including detailing the schedule for inspection activities, team composition (by name, affiliation and assignment) and other appropriate logistical details. A fonn is provided in Appendix IC to assist in transmitting the names and assignments of team members.
3.
A meeting of the inspection team should be scheduled prior to tho' inspection to familiarize them with the site specific conditions.
This meeting should be scheduled in the geographical location of 4
the plant site. 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:
7 Issue Date:
10/01/87 i
4
DRAFT REVISION A (a) discussion of the licensee's management and emergency organizstion.
(b) coordination of the team inspection assignmerits including the preparation of the written evaluations and findings for the various portions of the inspection report.
(c) relationship of the emergency functions among the various ERFs for the specific site.
(d) site specific aspects of the licensee's Emergency Plan and EPIPs.
(e) time phasing of the accident scenario for the exercise.
(f) work space and arrangements provided for the team by the licensee.
(g) review post or existing facility related problem areas.
4.
A meeting between the team members and the personnel listed by the licensee in Appendix 2 should be scheduled at the earliest time available after the team arrives onsite. This meeting vill offer the team members an early opportunity to meet their primary licensee contacts, schedule interviews and identify additional personnel or resources needed for infonnation.
1.
Preparation of the Inspection Report i
1.
The inspection team will provide the Regional team leader with a l
sumary of their findings before the exit meeting scheduled prior to the team leaving the site. No later th:n ten working days after leaving the site all team members will provide a final evaluation l
8 Istue Date:
10/01/87 i
DRAFT REVISION A and findings report to the Regional team leader and_ the NRC Headquarters technical coordinator evaluating the areas assigned
- 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 violation, deviation or open item. These reports shall be used by the team Isader to prepare the final ERF evaluation report.
2.
No later than ten working days after receiving the last report from the individual team members, the Region will provide the final ERF evaluation report to the Chief. Emergency Preparedness Branch, Division of Radiological Protection and Emergency Preparedness, Office of Nuclear Reactor Regulation (PEPB/NRR) for review and concu rrence. The Chief, PEPB/NRR will pmvfde a concurrence by telephone to the Region within five working days after receipt of the ERF evaluation report.
3.
No later than 30 working days after leaving the sit) or within 20 working days after receiving the last report frem the team members, the Regicn will provide the final ERF evaluation report to the licensee signed by the appropriate Regional management and the team leader.
If the report contains identified violations or deviations tett require the licensee to remove or rip out ERFs or equipment that had been installed in good faith to meet previous guidance in order to meet the requirements of Supplement 1 to NUREG-0737, the concurrence of the Director, Office of Nuclear Reactor Regulation (NRR) will be obtained prior to the issuance of the report.
4.
The inspection report will follow standards and guidelines given in Manual Chapter No. 0610, "Inspection Reports - Format and Content."
The report will clearly identify all violations, deviations, and open items observed during the ERF evaluation in the 'indings.
These items and any other items which the licensee has agreed to correct anytime prior to the issuance of the final ERF Evaluation l
I 9
Issue Date:
10/01/87
DRAFT REVISION A Report will be tracked for correction within a schedule to be negotiated between the licensee or applicant, Regional management
- and the Proje.t Manager, NRR. When corrections cannot be agreed to, reconnend4tions for possible further regulatory action will be fomarded to the Director of the appropriate project division of MRR. If the correction of &ny violation 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 require-ments of Supplement 1 to NUREG-0737, the approval of any such orders will be obtained from the Direc.or, NRR.
5.
The exercise report should be prepared and issued in accordance with current inspection guidance and Regional policy.
03.02 hsessment of Radiological Releases. Evaluate whetner the ERFs are adequately equipped to detsmine the magnitude of and for continuously assesdng the irtpact of a release of radioactive material to the environnwnt using the guidance to detemine adequacy in this area as presented below.
This guidance is applicable to both the TSC and EOF unless otherwise noted.
a.
Evaluate r.ethods available for determining radioactive release rates (source term) to the envf ronment in an accident situation.
1.
Review precalculated relationships of variables to accident conditions. Typical relationships to review include:
(a) Containment radiation exposure rates, coolant radioactivity concentrations, and coolant chemistry to core conditions (b) Hydrogen concentration in containment to containment and fuel clad failure s
(c) Area radiation monitor readings outside containment to containment high radiation monitor readings P
10 Issue Date:
10/01/87
DRAFT REVISION A (d) Letdown line and main steam line process radiation monitor readings to coolant radioactivity concentration (e) Affect on stack monitor readings of gamma radiation shine from containment.
2.
Evaluate the variables available and the calculation methods used to de'srmine source terms for all potential release pathways (e.g.,
effluent monitors, containment monitors, containment leak rate, fuel damage monitors, real time environmental monitoring, post-accident sampling results, in-plant radiological monitoring). Eval-uate methods for dealing with inoperable or offscale monitoring instruments, b.
Determine whether the dost assessment method (s) used are adequate for calculating thyroid inhalation dose commitment and whole body dose for applicable release pathways (both ground level and elevated releases) in the plume exposure pathway.
1.
Evaluate the capsbility of the primary dose assessment model to nake timely dose projections for variable release durations, variable distances ir; the plume IPZ, variable meteorological conditioris, and for variable and/or meltiple source term (s).
2.
Review the dose assessment model(s) capability for calculating current dose rates, integrated doses, and projected doses for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for specific points in the plume EPZ.
3.
Evaluate the adequacy of source term information entered into the model.
4 l
11
!ssue Date:
10/01/87 3
DRAFT REVISION A (a) Determine the adequacy of any default isotopic mixes used in the model (e.g., consideration of isotopic mix changes based on the time after shutdown, appropriate use of dose equivalentvalues).
(b)
If individual radionuclide concentrations (e.g., from effluent grab sample results or post-accident sampling results) can be input into the model, evaluate the adequacy of the radionuclides which the model will accept.
4.
Evaluate the capability for entering meteorological and tource tem data into the model.
(a) Detemine the adequacy of the primary method for obtaining meteorological and source term data to input into the dnse model in the ERFs.
(b) Assure that a backup capability exists for obtaining source term and meteorological data if this data is automatically entered into the model.
5.
Determine whether the sensitivity and uncertainty inherent in dose assessment have been established and factored into the dose projections.
6.
Review the systematic validation and verification analysis performed by the licensee er a contractor on the dose asstssment niodel.
(a)
Review any comparisons the Itcensee has mat.. with other documented dose assesstr, ant models.
i 12 1ssue Date:
10/01/87
DMFT KfVISION A (b)
If comparisons to other models are not available, compare the licensee's model to the extent practical to a straight line Gaussian plume projection model.
~
(c)
Determine the adequacy of whole body and thyroid dose conversion factors used in the licensee's model.
7.
Review how field monitoring data is used to correct or modify the dose projections (e.g., the use of the dose assessment model to backcalculate from field readings to release rate, how differences are interpreted if f41d readings and model estimates are not the same).
8.
Determine the adequacy of the backup dose calculation method which would be used if the primary method where unavailable in the TSC or EOF.
i 9.
If dose projections are used in decisionmaking (e.g., EAL detennination), evaluate the adequacy of the rapid dose projection capability on-shift, in either the Control Room or the TSC.
03.03 Meteorological Information. The inspection shall determine if the meteorological information available in the Control Room, TSC and EOF is adequate for continuously assessing the impact cf the release of radioactive material to the environment.
In making this deteimination, the inspector l
shall:
a.
Determine if recorded indications of wind direction, wind speed, and r
atmospheric stability are provided in the Control Room.
b.
Cetermine if the indications of wind direction, wind speed and atmospheric stability provided in the ERFs are representative of the meteorological conditions in the vicinity of the plant site.
In making this determination, consider facters such as:
the exposure of the 13 Issue Date:
10/01/87 l
DRAFT REVISION A sensors, their 1s 4 tion relative *o topographic features, and their location relative to potential Velease points ().g., ground-level or etai&ted).
If the site-specific indications of wind direction, wind speed, and atmospheric stability provided are not representative of the conditions in the vicinity of the plant site, detemine if other reliable meteorological information is provided that is representative of conditions in the vicinity of the plant site.
Detemine if the meteorological system provides reliable indications of wind direction, wind speed and atmospheric stability. The following steps should be followed in establishing reliability:
1 1.
Evaluate historical records o,' the availability of wind direction, wind speed, and atmospheric stability infonnation (e.g.,
approxim3tely 0.90 availability).
2.
Detemine if the instrumentation used to make the wind direction, wind speed, and atmospheric stability measurements mer the specifications set forth in RG 1.97.
3.
Detemine if instrument inspection, maintenance, and calibration procedures exist and are adequate.
4 Detemine if the meteorological instrumentation has been designed to facilitate the recognition, location, and repair, replacement or adjustment of malfunctioning modules.
5.
Detemine if *he instruments and their signal processing modules are in administrative 1y controllad areas.
6.
Other factors that hate a bearing on the rallability of the 4
indications provided should be considered (e.g., redundant sensors, backup systems and data from other locai. ions).
14 Issue Date:
10/01/87
-- a
ORAFT REVISION A d.
Determine what other site-specific meteorological information is provided that might be used in assessink the impacts of ine release of radioactive material (e.g., wind direction variability, precipitation, solar radiation, humidity).
e.
Determine if a method of voice communications has been established with the National Weather Service (or equivalent meteorological service) to obtain inforwation oa regional meteorological conditions and forecast capability.
f.
Determine if adequate facilities exist in the ERFs for the acquisition, display, and evaluation of meteorological de;a for determining protective measures, g.
Determine if the ERFs have the capability to store, analyze, display sufficient meteorological information to determine changes in status, foiecast status, and take appropriate actions.
In determining if sufficient meteorological information is available, the data requirements of all ERF functions should be considered.
For example, data on me*eorological variables, such as precipitation, that might affect I
protective action recommendations should be available, as well as all data needed by the dose assessment model.
In addition, there should be provision for obtaining meteorological data for use in dose assessment in the event that the data are unavailable from the primary data sources.
The alternate sources of information may include backup meteorological systems and default values.
If default values are provided, the basis of the values should be determined.
h.
Determine if the methods of collecting, storing, analyzing, displaying i
and communicating reteorological information in the ERFs are reliable.
03.04 TSC Variable Availability. The variables available in the TSC (by computer system display, status board or other means) are to be reviewed to 4
i 15 Issue Date:
10/01/87 i
a
DRAFT REV!$!0N A detemine their adequacy for allowing the TSC managers to perfom their function. The inspection procedures and techniques to be used are as follows:
a.
Obtain copies of documentation submitted by the Itcensee to NRC concerning commitments and progress on meeting the requirements of RG 1.97 (e.g., FSAR cosuitments and Safety Analysis Reports).
b.
Detemine which of the RG 1.97 variables are available in the TSC.
l After detemining which RG 1.97 variables are available and which are missing, detemine if the TSC variable set provided is sufficient to allow the TSC managers to perfom their designated functions. The variables provided should be sufficient to allow detemination of the following plant and environmental status:
1.
The continuous removal of heat from the core and associated cooling systems (e.g.,RHR,componentcoolingwater,cargencyservice water and auxiliary feedwater system status).
2.
Th( threat to or actual degradation of the fuel and fuel cladding (e.g., as indicated by subcooling margin, radioactivity in reactor coolant and core exit themocouple data).
3.
The integrity of the reactor coolant system (e.g., as indicated by pressurizer level, reactor vessel level, relief valve position and PWR steamline radioactivity).
4.
The integrity of the containment structure (e.g., as indicated by isolation valve status or by threats to containment such as increased hydrogen concentrations, temperature and pressure).
5.
The status and integrity of the liquid, solid and gaseous rad waste i
systems (e.g., radiation monitors and alarms associated with waste gas holdup tanks, liquid effluent lines, etc.).
16 issue Date:
10/01/87
DRAFT REVISION A 6.
Indications of damage resulting from a fueling or fuel pool accident (e.g., alarum and monitors associated with fuel pool water 1evel, and fuel handling area radiation levels).
~
c.
If a computer based data acquisition system is used to transmit and display variables in the TSC, a complete computer poin? list should be obtained from the lic.asee and used to verify the avail, tility of RG 1.97 variables, d.
If telephones (or radios) and status boards are used as the primary means of obtaining any RG 1.97 variables, the adequacy of the status boards as well as the qualification, numbers and assignment of communicators, and quality of the communication link to the Control Room must be verified. Where a video data transmission system is used, the system capability to accurately obtain Control Room RG 1.97 instrument data must be verified.
e.
The data determined to be available in item b above, is reviewed for its adequacy to evaluate the existing and projected status of the core, coolant system and containment to support adequate determination of proper protective action recoewendations (e.g., as in NUREG-0654, l
Appendix 1, General Emergency example initiating condition No. 4).
1 03.05 TSC Functional Capabilities. Determine if during periods of activation l
the TSC will operate uninterrupted and whether the TSC is equipped to provide the TSC managers with the capability to perform EOF functions until the EOF becomes functional. In order to make this determination the inspector should evaluate the following areas:
a.
Determine whether the power supplies will assure that the TSC will function without interruption during an emergency (i.e., normal power, UPS systems, emergency diesel, emergency battery supplies, and alternate S
sources of offsite power).
Individual systems and components for which reliable power is important include telephones, radios, data acquisition 17 Issue Date:
10/01/87
DRAFT REYI510N A systems, data display systems, computerized dose assessment systems, facility lighting, ventilation systems, microfiech card readers, and radiation monitoring systems.
b.
Determine if data analysis is adequate to support the TSC functions by evaluating the following areas:
4 1.
Determine whether current system status is available (e.g., valve position, equipment operation, pump status).
2.
Detemine whether data analysis will facilitate detemination of reactor status past, present, and future.
For example, evaluate trending capability to detemine if trends of the following parameters are maintained versus time:
containment pressure, containment temperature, containment radiation exposure rates, contairanent hydrogen concentrations, primary coolant temperature, offgas radioactivity, primary coolant pressure, primary coolant I
inventory, power level, plant radiation exposure rates and concentrations, and makeup water inventory.
3.
Determine whether precalculated relationships of variables to accident conditions have been established (e.g., Containment radiation levels vs fuel damage, containment pressure to containment failure).
l t
l 4
Detemine whether data analysis is performed in a manner easily related to EAL criteria (i.e., data displays should contain the parameters and relationships required to allow a clear association
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with EAL criteria).
03.06 TSC Habitability. Evaluate the habitability of the TSC to determine if 4
the radiological protection provided is adequate to ensure that any person working in the TSC would not receive a radiation exposure in excess of 5 rem
]
whole body or its equivalent (e.g., 25 rem to the thyroid) for the duration of i
18 Issue Date:
10/01/87 i
l
DRAFT REVISION A an accident.
Severe accident conditions where the control room would not be habitable, should not be used to evaluate TSC habitability (see GDC 19). The evaluation *should include the TSC game radiation shielding and the emergency ventilation system (i.e., ventilation filtration, posi,tive pressure isolation, acceptance / surveillance / maintenance records).
In addition the bases for detennining the adequacy of TSC habitability should also be examined (i.e.,
design basis, documentation of calculations and measurements).
03.07 TSC Data Collection, Storage, Analysis and Display. Careful reviews of licensee documentation and corresponding system hardware are required to estab-lish whether TSC data systems will provide the TSC managers with reliable data collection, storage, display and comunications such that correct plant site and regional status can be detennined in time to take appropriate actions.
a.
The inspector should perfonn a review of the TSC systems:
i 1.
Methods for data collection will need to be established. Data
(
acquisition may be done using, digital / analog instrumentation, voice comunication, etc. Once it is established how data are gathered, evaluate if the methods used will provide timely plant status infonnation.
l 1
2.
Identify and characterize the use of data displays in the TSC.
I
]
Typical displays would include:
analog and digital reters; j
catahode ray tubes (crt's); hard copy devices; chart recorders;
{
status boards; and other manual displays. After identifying i
display devices determine: 1) if displayed data are appropriately l
labeled, leg ble, updated in a timely manner, and properly i
j organized; 2) if TSC displays are adequate in number, easily updated, and facilitate user access; 3) if trending displays support the intended functions of the TSC; and 4) if user 5
documentation is readily available to explain the use of displays.
l l
J j
19
!ssue Date:
10/01/87 4
DRAFT REVIS! ] A 3.
Ascertain the time resolution of the data to determine if plant parameter changes can be detected and reoorted without the loss of
'significant information (e.g., pressure spike in containment due to ahydrogenburn).
4.
Review signal isolation effects of installed systems.
Specifically, review the system interface design and any system isolation verification / validation documentation to assure that significant signal degradation of installed systems is r.ot occurring and that interference, degradation or damage to any element of the safety system is prevented (see GDC 24).
5.
Established whether:
- 1) the data comunications capacity of the data acquisition system (s) is sufficient to access all data to be transmitted to the TSC; 2) the time resolution for data transmission of each of the variables is adequate to assure that no significant data are lost; 3) the data transmission is accurate; and 4) the means of transmission are technically adequate.
6.
Determine if the processing system capacity of the central processor is adequate to support data acquisition analysis, display, and storage requirements for the TSC. Other computational requirernents will need to be identified along with total projected processor resources utilized at peak loads to identify probable system degradation during an accident situation, if the central processor is using multitasking, it will need to be established whether essential TSC tasks would be degraded by concurrent tasks supporting other non-TSC functions.
7.
Data storage capacity will require review. This will include:
1) determining if data storage is adequate to support necessary data handling such as trending and analytical requirements; and 2) 4 l
determining if data storage is adequate to allow analytical review of the plant response to transitnts for TSC management.
20 Issue Date:
10/01/87 N
~ _ _____-
a k
REVISION A i
8.
Model and system mitability and validity will need to be reviewed i
to find out how the verification was done and whether the l
verification was an independent effort.
{
9.
The reliability of computer systems supporting T5C functions should be estabitshed by reviewing unavailability records, i
maintenance legs, vendor technical specifications, stallar system
{
I comparison, or end-to-end tests. The system should exceed an overall avt.ilability of 0.95 to be considered reliable, 4
- 10. Manual ostems need to be identified and reviewed to assure that i
I any data gathered, processed, or displayed in the TSC are reliable.
l Checks to support this review may include:
independent sources of j
information; crosschecks; confinnation between Source and i
t I
destination; and use of femal procedures or checklists.
i
}
1 f
- 11. Specifications of environmental control systems (i.e., air j
conditioning and humidity control systems) need to be reviewed to detemine if they meet the requirements of vendor supplied computers and peripherals used in the T:C.
i b.
As a part of evaluating the info. nation manager. ant and data acquisition f
system for the TSC and the E0F, the availability of the report on the j
l implementation of RG 1.g7 should be detemined. This report is required l
for each site by Supplement 1 of MUREG-0737 and must t>e submitted by the j
f licensee describing how the requirements are to be met. Deviations from the guidance are explicitly sh:nsn and a supporting justification or alternatives are presented in this report. The NRC Headquarters l
Technical Coordinator will detemine the availability of this report or' I
any other SER or NRC evaluations of the licensee's submittal. Copies will be provided to the individuals performing reactor operations.
l I
dose assessment, meteorology evaluations, and regior,a1 team leader to l
l assist them in evaluating the adeque.cy of the TSC and E0F database, f
il i
l
\\
\\
21 Issue Date:
10/01/87 l
i l
t ORAFT REVISION A 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 rnd/or the EOF, an evaluation will be performed of the SPDS as a part of this ERF inspection. This SPD? evaluation will be performed only for its adequacy as a par <. of the eargency data acquisition system for the use of TSC and/or the EOF and not '.s en operator aid in the Control Room. The adequacy of the SPDS as a part of the emergency data acquisition system will have no bearing on its acceptability at an oper-ator aid (reference Supplement I to NUREG-0737, item 4).
03.08 EOF Location and Habitabilitg. Detennine if the EOF location and habitability meet the requirements of Table 1 of Supplement 1 to NUREG-0737.
l To make this determination the inspector should evaluate the following areas:
a.
Determine if the EOF is located as described in Table 1 in Supplement I to NUREG-0737, b.
Identify which option as specified in the Supplement was chosen and determine if the EOF meets all the critoria for that option.
1 c.
If the EOF is located within the 10 sille EPZ determine if the appropriate nabitability requirments have been met. Evaluate the gama protection factor (PF) for areas used for comunications, dose assessrent, and decisionmeking to ensure that it is at least a PF of 5 for 0.7 MeV gama. The ventilation system HEPA filtration, facility isolation and the acc ptance/ surveillance / maintenance records should also be evaluated.
l l
03.09 EOF Functional Capabilities. Determine if the EOF is equipped to pro-vide the EOF ma.1 agers with the capability for management of the overall licen-see emergency response, coordination of radiological and environmental assess-ments, development of protective action recommendations and coordination of S
j 1
22 Issue Date:
10/01/87 t
DRAFT REVISI^] A emergency response activities with Federal State, and local agencies.
In order to make this detemination the inspector should evaluate the following areas:
~
a.
Determine if data analysis is adequate to support the EOF functions by evaluating the following areas:
1.
Detemine whether data analysis will facilitate determination of reactor status past, present, and future. For example, evaluate trending capability to detemine if trends of the following parameters are maintained versus time: containment pressure, containment radiation exposure rates, containment temperature, containment hydrogen concentrations, offgas radioactivity, and plant radiation exposure rates.
2.
Detemine whether precalculated relationships of variables to accident conditions have been established (e.g., containment radiation levels vs fuel damage, containment pressure vs containment leakage).
3.
Detemine whether data analysis is perfermed in a manner easily related to EAL criteria (i.e., data displays should contain the parameters and relationships required to allow a clear association with classification and protective action decisionmaking criteria).
4 Detemine if parameters are displayed in a mnner that makes it easy to determine deviations in parameters from nomal (e.g.,
superimposed curves, nomal ranges also displayed, displayed in percent of nomal),
b.
If a backup EOF is provided determine if it is adequate to accept the transfer of the dose assessment, comunications, and decisionmaking functions of the EOF if the primary EOF must be evacuated (e.g.,
comunications capability, data availability).
23 Issue Date:
10/01/87
DRAFT REV!$10N A c.
If the licensee has a primary E0F within 10 miles of the plant site and a backup EOF outside of the 10 mile radius, a degree of reliability is provi*ded by the redundant locations.
EOF power supplies ned only be evaluated if one of the following two situations is encountered:
1) there is a single EOF outside the 10 mile plant ~ radius or 2) the primary ano backup EOFs are on a comon power grid which has a high probability of causing a power failure affecting both EOFs.
If either situation exists, detemine whether the power supplies will assure that the EOF will function reliability during an emergency using the same procedure described in item 03.054. for the TSC.
03.10 E0F Variable Availability. ThevariablesavailableintheEOF(by computer system display, status board or other means) are to be reviewed to determine their adequacy for allowing EOF managers to perform their function.
In contrast to the more all inclusive set of variables expected to be available in the TSC, the set of variables required in the EOF are limited to those necessary to monitor actual or potential containment conditions and releases of radioactivity from the plant. The inspection procedures and techniques to be used are as follows:
a.
Obtain copies of documentation submitted by the licensee the NRC concerning comitments and progress on meeting the requirements of f
RG 1.97 (e.g., FSAR commitments).
I b.
Detennine which of the RG 1.97 variables I;re available in the EOF and which are missing. Detennine if the EOF variable set provided is sufficient to allow the EOF managers to perfonn their designated i
functions. The variables provided should be sufficient to allow t
detennination of the following containment, radiological and a
environmental status:
l l
4 1
1 24 Issue Date:
10/01/87
DRAFT REV!$!0N A 1.
The integrity of the containment structure (e.g., as indicated by isolation valve status or by threats to containment such as
' increased hydrogen concentrations, temperature and pressure).
~
2.
The release of radioactivity from the plant (e.g., as indicated by process radiation monitors on ratease points, building area and containment radiation monitors, and ventilation system flowrates).
3.
Meteorological variables.
(Note: meteorological variables are l
covered in Section 03.03 of this procedure).
1 If a computer based data acquisition system is used to transmit and l
c.
display variables in the EOF, a complete computer point list should be j
obtained from the licensee and used to verify the availability of RG j
1.97 variables.
d.
If telephones (or radiot) and status boards are used as the primary means of obtaining any RG 1.97 variables, the adequacy of the status I
boards as well as the qualification, numbers and assignment of i
comunicators, and quality of the comunication link to the EOF must be l
I verified. Where a video data transmission system is used, the system i
capability to accurately obtain Control Room RG 1.97 instrument data i
mutt be verified.
i The data determined to be available in item b above is reviewed for its i
e.
adequacy to evaleats the existing and projected status of the l
I containment and the actual or potential releases of radioactive material from the plant.
i 03.11 EOF Data Collection, Storage, Analysis and Display.
In-depth reviews f
of licensee documentation and corresponding system hardware are required to establish whether data systems supporting the EOF will provide the EOF I
managers with reliable data collection, storage, display and comunications l
such that correct plant site and regional status can be determined in time to 25 Issue Date:
10/01/d7 l
l
_. _a
DRAFT REVISION A take appropriate actions. The review methods described in Section 03.07 for the TSC data acquisition system should be repeated in the evaluation of the EOF data ac'quisition system. The EOF evaluation should be considered from the perspective of the needs of the EOF managers. The necessity to complete a
~
separate review or to repeat all the steps in the evalaution is dependent on whether these data systems use the same acquisition hardware, finnvare and software as well as whether they use a conmon data base, a
82412 INSPECTION RESOURCES
(
04.01 The estimated resources need to coeplete a typical ERF evaluation at a nuclear power plant site are:
i a.
The esti 9tted manhours for each specialist team member:
l 1.
Preparation time for the inspection = 16 2.
Travel time to and from the site
= 16 3.
Conducting the inspection onsite
= 32 l
4.
Writing a report of results and
= 40 findings 5.
Total 104 Manheurs j
b.
The estimated manhours for the Regional team leader:
i 1.
Preparation time for the inspection = 20 8
2.
Travel time to and from the site
=
i l
3.
Conducting the inspection onsite
= 32 f
4 Writing and staffing the inspection = 48 report 5.
Total 108 Manhours The total estimated resources needed for an ERF avaluation are (104 c.
x 4 + 108 = 524) 524 Mai. hours.
i l
26 Issue Date:
10/01/87 l
DRAFT REVISION A 82412-05 REFERENCES U.S. Code of Federal Regulations (CFR). Title 10. Part 50, ' Licensing of Production and Utilization Facilities," Appendix A, ' General Design Criteria for Nuclear Power Plants " Criteria 19 and 24.
U.S. Code of Federal Regulations (CFR). Title 10, Part 50, "Licensing of Production and Utilization Facilities," Appendix E, "Emergency Plans for Production and Utilization Facilities."
U.S. Codo of Federal Regulations (CFR). Title 10, Part 50.47, "Emergency Plans."
U.S. Nuclear Regulatory Comission (NRC).
1980. Criteria for Preparation and Evaluation of Radio'.tological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants. NUREG-0654, FEMA-REP-1, Rev. 1. Washington D.C.
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 Fact 11 ties. NUREG-0696, Washington, D.C.
l U.S.NuclearRegulatoryCc;sission(NRC).
Inspection and Enforcement Manual, IE Inspection Procedur9 82301, "Evaluation of Exercise for Power Reactors."
U.S. Nuclear Regulatory Comission (NRC).
Inspection and Enforcement Manual, 4
Chapter 0601, i
1 l
l 27 Issue Date:
10/01/87
ORAFT REVIS!1] A APPENDIX 1A REC 0* ENDED INSPECTION ASSIGNMENT MATRIX Inspection Items Technical Area Assigned t.
4 l
I "3
5 1
i e
i s
e s
e 7
S 3
Il i
cI E
a ac r
03.0Z Assessment of Radiological Releases (a) Source Term X
e I
K (D) Dose Assessment 03.03 Meteorological Infonnation (a) Control Room Information e
~
(b) Representive Data e_
e ~
(c) Data Reliability
=
e (d) Other Data Availability
~
~~
RW5 DaTi~ Availability e
4 (e) e (f) Data Adequacy
~(g) Data 5torage, Ulsplay, e
X Analysis T
I (h) Data Handling Reliability
~
03.04 T5C Variable Availability (a) Documentation for Reg Guide 1.97 Yariables lb) Reg Guf de 1.97 Vartable j~
e X
X i
Availability & Sufficiency T
K
~~
(c ) Coriputer Data
~
Ldj Manual Deta
~
e K
te) Data adequacy
~
03.05 T5C Functional Capabilites 4
(a) TSC Power Supplies o
I (b) T5C Data Analysts e - Responsible for write-up of item
't X - Should provide input fee item IA.1
!ssue Date:
10/01/87 i
ItEY!$!ZJ A l
Inspection Items Technical Area Assigned ef nl
.3 i
- l 5
5 e
a e
e e
m t
I I
S a
E a$
E W
03.06 TSC Habitability e
x
~
D3.07 T5C Data Collection Storage, Analyses and 1
Display (i. ) Review TSC Ssstees X
e
~~
(b) Data Acquisi3 ion Systees I
I I
e 03.05 EOF Variable Availability l
(a) Documentation for Reg Guide 1.97 Variables e
(b) Reg. Guide 1.97 Variable a
X X
Availability and sufficiency (c) computer Ta ta e
I (d) Ranual Data o
(e) Data Adequacy I
03.09 EOF Location and Habitability e
(a) location e
(b) Meets'Criterla of Supp. I
~
(c) Habitability I
03.10 EOF Functional Capab111ttes e
_X_
(ap Data Analysis Adequacy
__ _(b i Backup EOF I
I
,I e
(c j EOF RelIabilily e
~
UEll EOF Data Collection.
A
~~
Storay, Analysis and 1
Display s
a - Responsible for write-up of item X - Should provide input for itee i
l 1A.2 Issue Date:
10/01/87
DRAFT REV!51 ] A APPDG!X 18 RECOM4 ENDED INSP'.CTION AS$!9 MENT WTRIX Inspection Items Technical Area Assigned 6
4 na 5
l 5
e 1 yl I
I 8
E 3
g a
E a$ a E
L 5
~
i 03.0Z rssessment of Radlological Releases (e) Source Ters (b) Dose AssessRent 03.03 Meteorological Information (e) Control Room Information (b) Represent 1ve Data (c) Data Rellat111ty (d) Other Data Availability (e) M ita Availability (f) Data Adequacy
~ '
(g) Data 5torage, Display, Analysis (h) Data Mand 11ng Reliability 03.04 T5G Varf able Availability (a) Documentation for Reg Guide 1.97 i
Variables (b) Reg Evide 1.97 Variable Availability & Sufficiency (c) Caerputer Data (di Manual Data (e) Data Adequacy 03.05 15c Functional capabilites
~
v (a) TSC Power Supplies (b) T5C Da ta Analysis 03.06 T5C Habitability l
i 1
J 18.1 Issue Date:
10/01/87
DRAFT REV!5!] A Inspection items Technical Area Assigned 5=
n l
a 1
e s
e 5
j~u i
I s
b 1
03.07 TSC Data Collection Storage, Analyses and Display (a) Review TSC Systems (D) Data Acquisition Systems 03.05 EOF Yariable Availability (a) Documentation for Reg Guide 1.97 Variables (b) Reg. Guide 1.97 Variable Availability and Sufficiency (c) Computer Data (d) Ranual Data (e) Data Adecuacy 03.09 Ear Location and Habitability (a) location (b) Meets criteria of Supp. I (c) HabItabil1ty 03.10 EUF Functional Capabilities (a) Data Analysis Adequacy (b) Backup 10F (c)
EOF Reliability
~
U3.13 EOF Data collection,
.Storaga, Analysis and
, Display 4
18.2 Issue Date:
10/01/87
y IttV!510 A APPEW !X IC Team Assignments Team M r ERF Assignment *
~
Exercise Assignment 4
Specific appraisal assignments are as specified in Appendix 1A, Issue Date:
10/01/87
DRAFT A APPEWIX 2 1maseetien Itese Lieeneee Conteet Peree N I Re ferenCe/0GMmente Itas henttetten Individuellej, Powme 80s.
01.02 Assessmeat of Redf elenteel Releasel (a) Souree fers (b) Dese Assisement 03.03 Meteorele-test laformatten (a) Centrol Room Information (b) Representive Data (c) Data Reltability (d) Other Data Aval10111ty (e) Inr$ Data Availability (f) Data Adequacy (g) Data Storage, Otsplay, Analysts (h) E0F Data Mandling Reliant 11ty 03.04 TSC Verf able Availabtitty (e) Documentatien for Reg CuIde 1.97 Variable:
(b) Reg Guide f,97 Vartable Availability and Suffletency (c) Computer Data (d) Manual Data (e) Data Adequacy 03.05 TSC Funettonal CaeabtItties (a) TSC Power Suppites (b) TSC Dets Analyst s c
03.04 TSC Habitablit ty 03.07 TSC Data Ce11ectica Storsee, Analyses and 01soley (a) Review TSC Systens (b) Cata Acquiettien $; stems 03.04 EOF Vertade Avatleblitty (a) Deewmentatten for Reg Cvide 1.97 Variables (b) Reg. Cutde 1.97 Variable Availablitty and Suf fittency 2.1 IS$ue Date:
10/01/87 l
IN N T REV!5!O A APPDDIX 2 Bef eta 04/C4540att
.Ineseetten items (f eensee Centact Perstaael t
ites Greeat set tea ladtviduelle)
M (e) Computer Dete (d) Menuel Date (e) Dete.'f:; a 03.09 (1llr tweettea sad Mobitehtitty (e) Leestten (b) Meete Criterte of Supp.1 (e) Habitehtitty 03.10 for Funetienal Ceseti11tioo (4) Date A64yet e Adequacy (b) Sechup EOF (c) eor Rettability 03.11 ter Dete Co11ection. Stoceee.
Aa ireis eas eteeier 2.2
!ssue Date:
10/01/87
n l l DRAFY REVISION A APPEN0!X 3 Documentation needed to conduct the ERF appraisal.
Docuamitation for all team men 6ers:
o Emergency Plan e EPIPs e FSAR e Description and location of alternate ERFs Flant Systems Description Manuals e
Listing of types and quantities of equipment maintained in ERFs e
protective clothing dosimeters survey instruments SCBAs procedures referen:e material Dose Assessment Documentation:
Implementing procedures for both computerized and manual dose assessment.
e e User's guide for computertred dose assessment model.
Technical basis document for dose assessment model, e
Documentation of any coeparative studies done between the licensee's e
model and the state model(s).
Documentation of any verification studies done on the licensee's DA
- program, Maps of the area (10 and 50 mi radius).
o Computer Systems Doe mentation,:
Computer conf'.guration specification for Emergency Data Acquisition Syf tem, Plant Computer, and SPDS Description of data system operation (i.e., "user's guides')
e Records of system availability e
e Documentation of computer code verification 4
Examples of hard copy output for routine reports and graphical displays e
Block diagram of computer systems showing interf6ces.
4 1
3.1
!ssue Date:
10/01/87
ORAFT REV!$!0N A
_ Reactor Operations Documentation:
ElectPical one line diagrams from off-site to the TSC, nonnel power, emergency power, Itghting, phones, communication systems, station PgI, afero-wave, plant process compter, data acouisition systems. Same for E0F if near-site; if *ar-site, power feeds to the building.
EPIPs covering clas',1fication, core-damage assessment. TSC Manager responsibility 6nd EOF Manager responsibilities.
Integrated, living schedule for all ERF related items R.G.1.97 items e
R. C.1.97 submittal. EG4G review, final SER SAR by Itcensee on its Data Acquisition System and SPOS e
Plant Infonnation Manual on Plant Process Computer, SPDS, Radiation e
Monitoring System Electrical Distribution Inventory of TSC and EOF documents and references.
e i
Meteorological Documentation:
A block diag *am of the meteorological system showin; the ;;ath data takes from sensor to storage and display, identifying the main components in the system e.g., sensors, signal conditioning, dat-cquisition systems.
data processing, data storage, and data displays a.., Aeir locaticns.
l Technical specifications for system sensors and other system components, e
and a list of their special features, such as heaters for wind l
instruments.
A detailed descri tion of the tower and sensor mounts, and a plan view e
drawing, preferab y to scale.
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.
4 Documentation on meteorological data availability.
e A copy of the most recent joint frequency distribution of wind direction, wind speed and atm3 spheric stability.
l 1
i i
l l
3.2 Issue Date:
10/01/87 i
1
j DRAFT REVISION A A list of the locations where onsite meteorological data would be available during an emergency.
A list of sources of regional ueteorological data and forecasts noting e
formal agreements end contracts.
Written procedures reisted to dose assessant and activation of the ERFs.
A generic description of the methods used to evaluste transport, a
diffusion, deposition and other atmospheric processes in all ERFs.
Listings of computer codes used in dose assessment.
Suppor*bq documentation for atmospheric models including those in the dose asiesament codes, e.g., theoretical bases, code verifications, user's guiets, Maps of the area (10 avid 50 mi radius).
e
, Source Term Documentation:
One line drawings of plant's ventilation system showing the following:
e
- vent flow rates
- points monitored and description of monitors
- fan and damper line-ups for nonnal and accident modes Any studies / evaluation ude of potential unNeiitored release paths, Effluent monitor calibration procedures and calibration data.
o Description of methods used to verify manufacturers primary calibration.
Core damage estimate procedures, Description of ARMS, and CAMS) plant radiation monitoring systems (process monitors.
e One line drawing fo* these systems and a list of manitors powered from vital power, Description of the plants pest accident monitoring system and its e
capabilities, Listing of and rational for nuclide library used by dose assessment e
procedures or computer programs.
- A description of the 'asic source ters assum tions used for accident 3
scenarios treated by manual and computerized iose assessment methods, and the rationale behind each.
3.3 Issue Date:
10/01/87
Eairref A/ve r Q-0 D7
.S u,yle,nn f /
the,dtr d{ /W b[M L.50cw t
8.4 Emergency Oper1tions Facility (EOF) 8.4.1 Requirerents a.
The EOF is a Itt.ensee controlled and operated facility.
Tne EOF provides for management of overall licensee emergency response, coordination of radiological and environmental assessment, rievelopment of recomendations for public protectiva actions, and coordination of emer-gency respnse activities with Federal State and local agencies.
Wh n the EOF is activated, it will be staffed by pre-designated emergency personnel identified in the emergency plan. A-designated senior licensee official will mnage licensee activities in the EOF.
Facilities shall be provided in the EOF for the acquisition, display and evaluation of radiological and meteorological data and containment conditions necessary to determine protective measures. These facilities will be used to evaluate the magnitude and effects of actual or potential radio-active releases from the plant and to determine cose projections.
The EOF wil' be:
b.
Located and provided with radiation protection features as described in Table 1 (previous guidance approved by the Cotmission) and with appropriate radiological monitor-ing systems, c.
Sufficient to accomodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the EOF.
d.
Structurally built in accordance with the Uniform Building
- Code, e.
Environmentally controlled to provide room air temperature, t.umid't) and cleanliness 6ppr'Wia'.e for p(nonnel and equipment.
I f.
Provided with reliable voice and data comunications l
facilities to the TSC and control room, and reliable voice comunication facilities to OSC and to NRC, State and local emergency operations centers.
t t
~.
i
- 23 g.
Capable of reliable collection, storage, analysis, display and communication of information on containment conditions, 7
radiological releases and meteorology sufficient to deter-mine site and regional status, determine changes in status, forecast status and take appropriate actions. Variables from the following categories that are essential to EOF functions shall be availsble in the EOF (1) variables from the appropriate Table 1 or 2 of Regulatory Guide 1.97 (Rev. 2), and j
(ii) the meteorological varisoles in Regulatory Guide 1.97 (Rev. 2) for site vicinity and regional data available via comunication from the National Weather Service.
h.
Provided wii.h up to date plant r
.cs (crawings, I
schematic diagrams, etc.), procedures, emergency plans l
and environmental information (such as geophysical data) i needed to perform EOF functions.
i 1.
Staffed using Table 2 (previous guidance approved by the Comission) as a goal.
Reasonable exceptions to goals i
for the number of additional staff personnel and response times for their arrival should be justified and will i
be considered by NPO staff.
I j.
Provided with indust tal security when it is activated to exclude unauthorf atd personnel and when it is idle to maintain its readit.ess.
k.
Designed taking into account good human factors en@-eering i
principles.
8.4.2 Documentation and NRC Review The conceptual design for emergency response facilities (TSC, OSC, and EOF) have been submitted to NRO for review. In many cases, the lack of detail in these submittals has precluded an NRC decision of acceotability, Se,=c de: ige.: here beer.
?
- hsapproved because they clearly did not meet the intent of i
the applicable regulations. NRC does sect intend to approve each design prior to implementation, but rather has provided
[
' n this document those requirenents which should be satisfied.
These requirements provided a degree of flexibility within which licensees can exercise management prerogatives in designing and bitt1 ding emergency response facilities (ERF) that satisfy spt :ific needs of each licensee. The foremost consideration regarding ERFs is that they provide adequate i
24 -
adequate capabilities of licensees to respond to emergencies.
NUREG guidance on ERFs has been intended to address specific issues which the Corriission believes should be considered in i
achieving improved capabilities.
Licensees should assure that the t-esign of IRFs satisfies these requirements. Exemptions from or alternative rethods of implenenting these requirements should be discussed with NRC staff and in some cases could require Conmission approval.
Licensees should continue work on ERFs to complete them accord-ing to scliedules that till be negotiated on a plant-specific basis.
NRC will conduct appraisals of completed facilities to verify that these requirements have been satisfied and that ERFs are e.apable of performing their intended functions.
Licensees reed not document their actions on each specific item contained in NUREG-0696 or 0814.
4
[
\\-
m A
.}*1m /
- s. C
- 9 C O.8
- =m*
M N
k' o '
8.4.3 Reference Documents (Emergency Response Facilities) 10 CrR 50.47(b) -- Requirements for emergency facilities and i
equipment for OLs.
10 CFR 50.54(q) and Appendix E. Paragraph 1*I.E -- Requirement.
for emergency facilities and equipment for ors.
NUREG-0660 -- Description of and implementation schedule for TSC, 05C and E0F.
l Eisenhut letter to power reactor licensees 9/13/79 -- Request for comitment to meet requirements Denton letter to power reactor licensees 10/30/79 -- Clartftca-ston of requirements.
)
WUREG-0654 -- Radiological Emergency Response Plans f
i NUREG-0696 -- Functional criteria for emergency response facilities.
!!UREG-0737 -- Guidance on meteorological monitoring and dose 1
assessmr.-t.
Eisenhut letter to pcwer reactor license 2/18/81 -- Comission approved guidance on location, habiubility and staff for j
emergency facilities. Request and deadline for submittal of conceptual design of facilities.
i MUREG-0814 (Draf t Report for Cocment) -- Methodology far evalu-
. ation of emergency response f acilities.
NUREG-0818 (Draf t Report for Corseent) -- Emergency Action Levels
]
Reg. Suide 1.97 (Rev. 2) -- Guidance for variables to be used j
it selected emergency response facilities.
COVcA 90t'7, January 21, 1981 -- Comission approval guidance en f 7 location and habitability.
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Secres, memorandum 581-1'), February 19, 1981 -- Comission j
approval of WUREG-0696 as pencral guidance only.
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. TABLE 1.
. EMERGENCY CPERATIONS FACILITY
.-l
- Option 2 Option 1
^ - One Facility Two Facilities o'At or Beyond 10 miles.
~
C1:s'e-in Primary:
Reduce liabitability" o within 10 illes o.No sp,ecial protection factor.
~
o If beyond 20 miles, specific.
..:, j jf o protection-factor = 5
~
with HEPA (no charcoal) 4 Commission, and some provi-
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o ventilation isolation Aproval required by the
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sion for NRC site team closer
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- to_ site.
,,. N o Strongly recommended location ~
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- J be coordinated with offsite -
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. authorities.
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Backup EOF 2N
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- o b tween 10-20 miles
.. 3 ' ~-
-' o.30 separate, dedicated
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fccility
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, o arrangements for portable
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bxkup equipment
~
o strongly recommended location
' 'e
/ be' coordinate <1 with offsite f
tuthorities o ccntinuity of dose projection End decistor, making capability For both Opt. ions:
- located outside security boundcry
- space for about 10 HRC employees y.'.
- none des!0nated for severe phenomena, e.g., earthquakes
.. ~.-
abitzbility requirements are only for the part of the EOF in which dose assessments communications and ecision making take place.
a utility has begun construction of a new building far an EOF that is located with 5 miles, that new cility is acceptable (with less than protection factor of 5 and ventilation isolation and HEPA) provided..
ct a backup EOF similar to "B" in Option 1 is provided.-
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- 22 8.4 Emergency Operations Facility (EOF) 8.4.1 Requirements a.
The EOF is a licensee controiled and operated facility.
The EOF provides for management of overall licensee emergency remon M, coordination of radiological and environ"en' c,essment, developmant of recomendations for publu.vtective actions, and coordiriation of emer-gency response activities with Federal, State and local agencies.
When the EOF is activated, it will be staffed by pre-designated emergency personnel identified in the emergency plan. A designated senior licensee official will manage licensee activities in the EOF.
Facilities shall be provided in the ECF for the acquisition, display and evaluation of radiological and meteorological data and containment conditions necessary to determine protective measures. These facil,ities will be used to evaluate the magnitude and effects of actual or potential radio-active re? cases from the plant and to determine dose projections.
The EOF will be:
b.
Located and provided with radiation protection features at described in Table 1 (previous guidance approved by the Comission) and with appropriate radfulogical monitor-ing systems.
c.
Sufficient to accomodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the EOF.
d.
Structurally built in accerdance wita the Uniform Building l
Code.
e.
Envi.3nmentally controlled to provide room air tenperature, humidi y and cleanliness appropriate for personnel and equipment.
f.
Provided with reliable voice and data comunications i
facilities to the TSC and control room, and reliable l
voice communication facilities to OSC and to MC, State and local emergency operations centers.
I
9
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g.
Capable of reliable collection, storage, analysis, display and communication of information on containment conditions, radiological releases and meteorology sufficient to deter-mine site and regional status, determine changes in status, forecast status and take appropriate actions. Variables from the following categories that are essential to EOF functions shall be available in the EOF:
(1) variables from the appropriate Table 1 or 2 of Regulatory Guide 1.97 (Rev. 2), and (ii)
- the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for site vicinity and regional data available via comunication from the National Weather Service.
h.
Provided with up to date plant records (drawings, schematic diagrams, etc.), procedures, emergency plans and environmental information (such as geophysical data) needed to perform EOF functions.
i.
Staffed using Table 2 (previous guidance approved by the Comissien) as a goal. Reasonable exceptions to goals for the number of additional staff persennel and response times foJ their arrival should be justified and will be considered by NRC staff.
j.
Provided with industrial security when it is activated to exclude unauthorized personnel and when it is idle to maintain its readiness, k.
Designed taking into account good human factors engineering principles.
8.4.2 Documentation and NRC Review The conceptual design for emergency response facilities (TSC, OSC, and EOF) have been submitted to NRC for review.
In many cases, the lack of detail in these submittals has precluded an NRC decision of acceptabliity. Some designs have been disapproved because they clearly did not meet the intent of the applicable regulations. NRC does not intend to approve sach design prior to implementation, but rather has provided in this document those requirements which should be satisfied.
These requirements provided a degree of flexibility within which licensees can exercise management prerogatives in designing and building emergency response facilities (ERF) that satisfy specific needs of each licensee. The foremost consideration regarding ERFs is that they provide adequate t
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adequate capabilities of licensees to respond to emergencies.
NUREG guidance on ERFs has been intended to address specific issues which the Commission believes should be considered in achieving improved capabilities.
Licensees should assure that the design of ERFs sat'sfies these requirements. Exemptions from or alternative Aethods of implementing these requirements should be discussed with NRC staff and in some cases could require Commission approval.
Licensees should continue work on ERFs to complete them accord-ing to schedules that will be negotiated on a plant-specific basis. NRC will conduct appraisals o/ completed facilities to verify that these requirements have been satisfied and that ERFs are capable of performing their intended functions.
Licensees need not document their actions on each specific item contained in NUREG-0696 or 0814.
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e 8.4.3 Reference Documents (Emergency Response Facilities) 10 CFR 50.47(b) -- Requirements for emergency facilities and equipment for OLs.
10 CFR 50.54(q) and Appendix E, Paragraph IV.E -- Requirements for emergency facilities and equipment for ors.
NUREG-0660 -- Description of and implementation schedule for TSC, OSC and EOF.
Eisenhut letter to power reactor licensees 9/13/79 -- Request for comitment to meet requirements Denton letter to power reactor licensees 10/30/79 -- Clarifica-tion of requirements.
NUREG-0654 -- Radiological Emergency Response Plans NUREG-0696 -- Functional criteria for emergency response facilities.
NUREG-0737 -- Guidance on meteoro',ogical monitoring and dese assessment.
Eisenhut letter to power reactor license 2/18/81 -- Comission approved guidance en location, habitability and staff for emergency facilities. Request and deadline for submittal of conceptual design of facilities.
NUREG-0814 (Draf t Report for Coment) -- Nethodology for evalu-ation of emergency response facilities.
MUREG-0818 (Draf t Report for Coment) -- Emergency Action Levels Reg. Guide 1.97 (Rev. 2) -- Guidance for variables to be used in selected emergency response facilities.
CON A-80-37, J anuary 21, 1951 -- Comission approval guidance en EOF location and habitability.
Secretary memorandum 581-13, February 19, 1981 -- Comission approval of NUREG-0696 as general guidance only.
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TA8tE 1
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. EMERGENCY OP' ERAT 10N5 FACILITY
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- Option 2 Option 1
~ One Fad 11ty
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Two Facl11 ties
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(
C1csh-in Primary:
Reduce liabitability*
- o At or Beyond 10 miles.
o
- a 10 alles o No special protection factor.
- o If beyond 20 miles, specific. -
.. :.'.q h o
act.fon-factor = 5 approval required by the
. :- 1. ',.
...iation isolation c
[~.o with IIEPA (no charcoal) 4 L
' Commission, and some provi-
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~
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sion for NRC site team closer
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..+'.
'to site.
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,,..' o Strongly recommended location-
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~3 be coordinated with offsite l
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.:.,authortties.
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Backup EOF i i_-
- o between 10-20 m11es l
',o no separate, dedicated
- ., !., b;
-. : V, *. -
facility
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, o arrangements for portable
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backup equipment
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o strongly recommended location he~ coordinated with offsite I
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cuthorities o continuity of dose projection
^
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cod decision making capability
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Fnr both Options:
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- located outside security boundary
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- space for about.10 NRC employees
~
- none des 1Dnated for severe phenomena, e.g., earthquakes
.:,.-i abitability requirements are only for the part of the EOF in which dose assessments communications and
~
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ecision making take place.
~
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a' utility has begun cor.struction of a new building for an EOF that is located with 5 miles, that new cility is acceptable (with 1ess than protection factor of 5 and ventilation isolation and llEPA) provided at a backup EOF slal1ar to "B" in Option 1 is provided.-
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